ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Nevada Test Site – 

Occupational External Dose

Document Number:

ORAUT-TKBS-0008-6 

Effective Date: 07/21/2006

Revision No.: 00 PC-1 

Controlled Copy No.: ________

Page 1 of 57

Subject Expert: Eugene M. Rollins

Document Owner

Approval: Signature on File Date: 09/16/2004

Eugene M. Rollins, TBD Team Leader

Approval: Signature on File Date: 09/16/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 09/13/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 09/21/2004

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

4

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

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

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

6.2.1 Photon Measurement Quantities ...................................................................................... 10

6.2.2 Beta Measurement Quantities .......................................................................................... 11

6.2.3 Neutron Measurement Quantities..................................................................................... 11

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

6.3.1 Administrative Practices ................................................................................................... 14

6.3.2 Dosimetry Technology...................................................................................................... 16

6.3.2.1 Beta/Gamma Dosimeters......................................................................................... 16

6.3.2.1.1 Photographic Film Dosimeters ........................................................................ 17

6.3.2.1.2 Beta Dosimetry with Film Badges.................................................................... 19

6.3.2.1.3 Evaluating Photon Doses with Film Badge Dosimetry..................................... 20

6.3.2.1.4 Estimation of Beta Doses from Film Badge Results ........................................ 21

6.3.2.1.5 Thermoluminescent Dosimeters...................................................................... 21

6.3.2.1.5.1 NRDS Thermoluminescent Dosimeters ................................................... 21

6.3.2.1.5.2 REECo Thermoluminescent Dosimeters.................................................. 22

6.3.2.1.5.3 Energy Response .................................................................................... 25

6.3.2.1.5.3.1 Dose Measurement .......................................................................26

6.3.2.1.5.3.2 Detection Limits .............................................................................27

6.3.2.1.5.3.3 Calibration .....................................................................................28

6.3.2.1.5.3.4 Quality Assurance..........................................................................29
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6.3.2.2 Neutron Dosimeters................................................................................................. 30

6.3.2.2.1 Film Dosimeters.............................................................................................. 33

6.3.2.2.2 Thermoluminescent Albedo Dosimeters.......................................................... 35

6.3.2.2.3 Track Etch Detectors....................................................................................... 36

6.3.2.2.3.1 Dose Algorithms ...................................................................................... 37

6.3.2.2.3.2 Lower Limit of Detection .......................................................................... 38

6.3.3 Calibration........................................................................................................................39

6.3.3.1 Photons ...................................................................................................................39

6.3.3.2 Neutrons.................................................................................................................. 39

6.3.4 Workplace Radiation Fields.............................................................................................. 40

6.3.4.1 Photon Energies ...................................................................................................... 40

6.3.4.2 Beta Particle Energies ............................................................................................. 40

6.3.4.3 Neutron Energies..................................................................................................... 43

References ........................................................................................................................................47

Glossary ............................................................................................................................................51

LIST OF TABLES

Table

6-1 External dosimetry .................................................................................................................

Page

12

6-2 Panasonic UD-802 dosimeter configuration ........................................................................... 25

6-3 Photon exposure to dose correction factors for NIST reference radiations ............................. 27

6-4 Background dose equivalent rates ......................................................................................... 28

6-5 NTS Panasonic UD-802 detection limits for DOELAP categories........................................... 28

6-6 Bias values and uncertainties for NTS DOELAP accreditation categories .............................. 30

6-7 Areas and operations where neutron exposure was possible................................................. 30

6-8 RBE, quality factors, or weighting factors for neutrons ........................................................... 32

6-9 Characteristics of ISO reference radiation radionuclide sources used for 

neutron dosimeter calibration ................................................................................................. 40

6-10 Fraction of Hp(10,0) for neutrons with energies less than En................................................... 46
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 3 of 57

LIST OF FIGURES

Figure

6-1 Security badge for NRDS TLD ...............................................................................................

Page

22

6-2 REECo badge holder ............................................................................................................. 23

6-3 REECo security badge holder with Panasonic UD-802 dosimeter.......................................... 23

6-4 Typical photon energy response curves for Li2B4O7 and CaSO4 TLD 

elements ................................................................................................................................25

6-5 Panasonic UD-802 energy responses.................................................................................... 26

6-6 Ratio of fluence to dose conversion coefficients for personal dose 

equivalent Hp(10, 0°) to conversion coefficients for dose equivalent H ................................... 33

6-7 NTA film neutron energy response......................................................................................... 34

6-8 Neutron energy response of NTS albedo dosimeters ............................................................. 35

6-9 Neutron energy response of NTS TED................................................................................... 36

6-10 Electrochemically etched CR-39 dose response .................................................................... 38

6-11 Lethargy neutron spectra for 252Cf calibration source with and without 

room scatter...........................................................................................................................44

6-12 Lethargy neutron spectra for 252Cf with various thickness moderators.................................... 44

6-13 Lethargy neutron spectra for an 241Am-Be calibration source with and 

without room scatter...............................................................................................................45

6-14 Fraction of Hp(10,0) due to neutrons with energies less than En ............................................. 45
Effective Date: 07/21/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-6 Page 4 of 57

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE REV. NO. DESCRIPTION

09/21/2004 09/21/2004 00 Information contained in Revision 00 is applicable 

only to employment periods after 1962 (postatmospheric testing phase) and to workers not

1. BLANCA (October 30, 1958)

identified as involved with drillback activities prior to 

1965. In addition, Revision 00 is not applicable to 

dose reconstruction for (1) workers involved with 

weapons testing at locations other than the NTS 

(South Pacific, Alaska, etc.); (2) workers affected by 

any of the following 10 underground tests that 

resulted in unexpected release of radioactive 

material:

2. DES MOINES (June 13, 1962)

3. BANEBERRY (December 18, 1970)

4. CAMPHOR (June 29, 1971)

5.. DIAGONAL LINE (November 24, 1971)

6. RIOLA (September 25, 1980)

7. AGRINI (March 31, 1984)

8. MIDAS MYTH (February 15, 1984)

9. MISTY RAIN (April 6, 1985)

10. MIGHTY OAK (April 10, 1986) 

It is also assumed that, because they were such a 

rarity, it would be likely that if a claimant was involved 

in one of these events, he [or his survivor] would 

mention it in his CATI. All of these conditions are 

likely verifiable by careful examination of the DOL, 

DOE (incident reports and dosimetry records), and 

OCAS documents. Incorporates internal and NIOSH 

comments. Initiated by Eugene M. Rollins.

09/21/2004 07/21/2006 00 PC-1 Approved page change revision incorporates 

definitions and directions for dose reconstruction for 

non-presumptive cancers that are excluded from the 

1951 through 1962 Special Exposure Cohort. Text 

was added or modified on pages 8-9 in Section 6.1. 

Incorporates NIOSH formal review comments. No 

sections were deleted. This revision results in no 

change to the assigned dose and no PER is required. 

Training required: As determined by the Task 

Manager. Initiated by Eugene M. Rollins. Approval:

Eugene M. Rollins, TBD Team Leader

Signature on File_____________ _____07/12/2006

John M. Byrne, Task 3 Manager

Signature on File_____________ _____07/12/2006 

Edward F. Maher, Task 5 Manager

Signature on File_____________ _____07/14/2006 

Kate Kimpan, Project Director

Signature on File_____________ _____07/20/2006
Effective Date: 07/21/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-6 Page 5 of 57

James W. Neton, Associate Director for Science

Signature on File_____________ _____07/21/2006
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 6 of 57

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ALARA as low as reasonably achievable

BREN Bare Reactor Experiment Nevada

C coulomb

°C degree Celsius

cm centimeter

CR-39 Columbia Resin-39 

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

ECF element correction factor

eV electron volt

°F degree Fahrenheit

GSD geometric standard deviation

Hp(d) personal dose equivalent (d = depth in millimeters)

hr hour

HRA High Radiation Area

Hz hertz

IARC International Agency for Research on Cancer

ICRP International Commission for Radiological Protection

ICRU International Commission on Radiological Units and Measurements

IDO Idaho Operations Office

ISO International Organization for Standardization

ISO International Organization for Standardization

keV kilovolt-electron

kg kilogram

kV kilovolt

LANL Los Alamos National Laboratory 

LET linear energy transfer

LLD lower limit of detection

m meter

MDL minimum detection level

MeV megavolt-electron

mg milligram

mm millimeter

MMD maximum missed dose

mR milliroentgen

mrem millirem
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 7 of 57

NBS National Bureau of Standards

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NIST National Institute of Standards and Technology

NRC National Research Council 

NRDS Nuclear Rocket Development Station

NRTS National Reactor Testing Station

NTA neutron track emulsion, type A

NTS Nevada Test Site

OW open window

Pan Am Pan American Airways

PIC pocket ionization chamber

QA quality assurance

QC quality control

R roentgen

RadCon radiation control

radex controlled radiation exclusion areas

RBE relative biological effectiveness

RCF run calibration factor

REECo Reynolds Electrical and Engineering Company

TBD technical basis document

TED track etch detector

TEDE total effective dose equivalent

TLD thermoluminescent dosimeter

TRU transuranic

VHRA Very High Radiation Area

wk week

YMP Yucca Mountain Project

yr year

Z atomic number
Effective Date: 07/21/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-6 Page 8 of 57

6.1 INTRODUCTION

Technical Basis Documents and Site Profile Documents are not official determinations made by the 

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

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

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

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

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

reconstruction. 

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

derived:

] guidelines established under 

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

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

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

• Radiation from naturally occurring radon present in conventional structures

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Effective Date: 07/21/2006 Revision No. 00 PC-1 Document No. ORAUT-TKBS-0008-6 Page 9 of 57

Operations at the Nevada Test Site (NTS) involved atmospheric and underground weapons tests, 

experimental reactor tests intended for aircraft and rocket propulsion, and low-level transuranic (TRU) 

waste disposal. NTS workers received exposures from a range of fission and activation products from 

test programs conducted from the early 1950s. Exposure venues of particular significance include 

sample and measurement device recovery to assess weapon yields, terrestrial and airborne fallout 

plume tracking, aircraft operations (when used) for weapon assembly deployment and surveillance, 

and post-event decontamination processes.

Radiation monitoring and control programs instituted with the mission of NTS included personal 

dosimetry, area monitoring, source term characterization, and measurements of fallout 

(contamination) dispersion. As NTS test programs progressed, efforts to measure exposures and 

limit dose improved. The atmospheric nuclear test series consisted of a number of operations 

between 1951 and 1958, and again from 1961 to 1963. Each operation consisted of a number of 

individual tests. Underground nuclear tests occurred at NTS from 1961 to 1992.

NIOSH considers the available data and methods for performing external dose reconstruction to be

adequate for estimating with sufficient accuracy the external radiation doses at NTS from 1951 until 

the present time.

6.2 DOSIMETRIC BASIS OF COMPARISON

Since the start of the Manhattan Engineering District program in the early 1940s, various dosimetric 

concepts and quantities have been used to measure and record occupational exposure from external 

radiation sources. The selection of the measurement quantities to be used for radiation protection

was initially based on the radiation interaction properties of the primary radiations of interest – 

photons (X- and gamma rays), electrons (beta particles), and neutrons – and the measurement 

methods employed. Sections 6.2.1 to 6.2.3 discuss these quantities briefly. 

The problem with the use of different quantities for different radiations is that it prevented direct 

comparison of measurements of the three primary radiation qualities. In the 1950s the industry 

recognized a need for a special quantity to facilitate comparison of measurements of the various 

radiation qualities. The concepts of quality factor and dose equivalent were formally introduced by the 

International Commission for Radiological Protection (ICRP) and the International Commission on 

Radiation Units and Measurements (ICRU) in 1962 (ICRU 1962). The special unit of dose equivalent

was the rem. As early as 1961, NTS radiation exposure criteria and guidance were given in rem 

(AEC 1961).

In 1985, ICRU defined a new set of operational quantities defined as radiation quantities for 

operational radiation protection measurement purposes (ICRU 1993). These quantities have a 

common definition for the three primary radiation qualities, so they have the advantage that they 

provide a means of direct comparison of measurements of these radiations.

The operational quantity recommended for individual or personal monitoring is the personal dose 

equivalent Hp(d) where d is the depth (in millimeters) and represents the point of reference for dose in 

tissue. For weakly penetrating radiation of significance to skin dose, d is 0.07 mm and the operational 

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

is 10 mm and the operational quantity is noted as Hp(10). Both Hp(0.07) and Hp(10) are 

recommended as the operational quantities to be recorded for radiological protection proposed by the 

ICRU (1993).
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 10 of 57

These personal dose equivalents, Hp(0.07) and Hp(10), have been used in the U.S. Department of 

Energy (DOE) Laboratory Accreditation Program (DOELAP) for accreditation of the Department’s 

personnel dosimetry systems since the 1980s (DOE 1986). The International Agency for Research 

on Cancer (IARC) Three Country Combined Study (Fix et al. 1997; Fix, Wilson, and Baumgartner 

1997) and IARC Collaborative Study (Thierry-Chef et al. 2002) selected Hp(10) as the quantity to 

assess error in recorded whole-body dose for workers in IARC nuclear worker epidemiologic studies.

6.2.1

From the beginning of operations at NTS, exposure was used as the basis of photon measurement. 

When in italics, the term exposure designates a radiation measurement quantity based on the 

electrical charges created in air due to interaction of photons. It is specific to air and photon

measurement. The unit of exposure is coulomb per kilogram, and the special unit of exposure is the 

roentgen (1 R = 2.58 x 10-4 C/kg). Although the term exposure dose was used in the mid-1950s, 

exposure is not dose because it is not a measure of energy deposition in mass of material. ICRU 

recommended the use of exposure in 1962 (ICRU 1962). However, dose in soft tissue can be 

determined from exposure by the following relationship, 

Photon Measurement Quantities

D = fR

where f is a constant that relates the exposure R in roentgen for a given photon energy to dose in soft 

tissue, and was typically taken to be 0.877 rad/R. Because 1 R is numerically slightly greater than 

1 rad, it is claimant-favorable to assume a numerical equivalence of the two quantities. Therefore, if 

the energy and hence the constant f of the exposing photon radiation are unknown, it is appropriate to 

assume that 1 R equals 1 rad, which equals 10 milligray. 

Instrument and dosimeter calibrations, dose measurements, and dose records were made in terms of 

exposure in units of milliroentgen. However, as noted above, from 1961 quarterly and annual limits 

were specified in millirem (AEC 1961), where the general term dose was used without specific 

reference to dose equivalent. As a result, individual monitoring results at NTS were recorded in 

millirem, although the measurements had been made in terms of exposure. However, for photons, 

the values of exposure and dose equivalent were considered to be essentially the same (Brady 2004; 

Brady and Iverson 1968). In effect, a de facto conversion factor of 1 rem/R was used for dose 

recording purposes.

Until July 1970 individual monitoring at the Nuclear Rocket Development Station (NRDS) was 

provided by the Reynolds Electrical and Engineering Company (REECo) using the standard NTS film 

dosimeter. From July 1970 until NRDS operations ceased in January 1973, individual monitoring was 

conducted by Pan American World Airways (Pan Am) using thermoluminescent dosimeters (TLDs) 

(Boone, Bennett, and Adams 1970). However, throughout NRDS operations, dosimeters were 

calibrated in terms of exposure and doses were recorded in millirem. 

Beginning in 1987, NTS occupational exposures were recorded in terms of personal dose equivalent, 

Hp(d). From 1962 to 1986, dose reconstructors should use the recorded photon dose values in terms 

of exposure, together with the Exposure to Organ Dose coefficients in Appendix B of NIOSH (2002) to 

determine organ dose. From 1987 on, the recorded values are in terms of Hp(10), and the Deep dose 

Equivalent to Organ Dose conversion factors (of NIOSH 2002, Appendix B) should be used.
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6.2.2

As was common practice, NTS beta particle measurements were made in terms of absorbed dose, D. 

Until introduction of the ICRU defined operational quantities in 1985, beta doses were recorded as 

millirad. Because the quality factor for electrons (beta particles) was set at 1, the absorbed dose 

values are considered numerically equivalent to dose equivalent.

Beta Measurement Quantities

6.2.3

The basis for comparison for neutron radiation is complex because, historically, the calibration of 

dosimeters to measure neutron dose was based on different dosimetric quantities (such as first 

collision dose and multiple collision dose). However, the neutron dose equivalent specified by the 

National Council on Radiation Protection and Measurements (NCRP) has been used since 1971. 

Evaluation of the numerical difference in comparison with the Hp(10) dose used in DOELAP 

performance testing is used to establish relative values of the dose conversion factors for the dose 

quantities in conjunction with characteristics of the respective neutron dosimeter response 

characteristics and workplace radiation fields.

Neutron Measurement Quantities

Dose reconstructors should convert recorded neutron dose to Hp(10) using the factors in Table 6-1, 

and use the Deep Dose Equivalent to Organ Dose conversion factors from Appendix B of NIOSH

(2002) to calculate the appropriate organ doses.

6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta and photon (X- and gamma rays) radiation type, energy and geometry of 

exposure in the workplace, and characteristics of the dosimeter response are crucial to assessment of 

bias and uncertainty of the original recorded dose. Bias and uncertainty for current DOE dosimetry 

systems are well documented for Hp(0.07) and Hp(10) under DOELAP. The performance of current 

dosimeters can be compared to performance characteristics of historical dosimetry systems in the 

same, or highly similar, workplaces. In addition, current performance testing techniques can be 

applied to earlier dosimetry systems to achieve a consistent evaluation of those systems. Dosimeter 

response characteristics for radiation types and energies in the workplace are crucial to the overall 

analysis of error in recorded dose.

Overall accuracy and precision of the original recorded individual worker doses and their 

comparability to be considered in using NIOSH (2002) guidelines depend on the following factors (Fix 

et al. 1997; Fix, Wilson, and Baumgartner 1997):

• Administrative practices adopted by facilities to calculate and record personnel dose based on 

technical, administrative, and statutory compliance considerations
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Table 6-1. External dosimetry.

Photon- beta dosimeters

Dosimeter Dates Operation Description Issue and exchange Measurement 

Quantity Bias GSD MDL MMD

DuPont 559 

Packet 

1961 to 

1966 All of NTS

DuPont 301-4 packet, including:

Type 508 low-range element (0.03 to 5 R)

Type 834 high-range element (5 to 800 R)

Lead filters, 0.028-in. thick, (symmetrical 

coverage on both sides)

Packet covered with 0.004-in.-thick plastic bag.

Exchanged monthly for 

general exposures and on 

exit from radiation areas 

for exposures likely to 

exceed 100 mR.

Photon: 

Exposure 1. 1.23b Photon: 40 

mR

Photon:

240 mR

DuPont Type 

556 Packet

1966 to 

1971 All of NTS

DuPont Type 556 film pack 

Type 508 (519 also referenced) low-range 

element (0.03 to 5 R)

Type 834 high-range element (10 to 1,000 R)

Four-area filter described: tantalum-cadmium, 

tantalum, Teflon, open

With fast neutron pack, dosimeter was 

sensitive to mixed fields with thermal and fast 

neutrons, X-rays, beta, and gamma

Exchanged monthly for 

general exposures and on 

exit from radiation areas 

for exposures likely to 

exceed 100 mR.

Photon: 

Exposure

Beta:

Absorbed dosea

1. 1.23b,d

Photon: 40 

mR

Beta: 

40 mreme

Photon:

240 mR

Beta:

240 mrem

Kodak Type 

III

1971 to 

1987 All of NTS Low-range element (0.03 to 10 R)

High-range element (10 to 800 R)

Exchanged monthly for 

general exposures and on 

exit from radiation areas 

for exposures likely to 

exceed 100 mR.

Photon: 

Exposure

Beta:

Absorbed dosea

1. 1.23d

Photon:

30 mRg

Beta: 

30 mremg

Photon:

240 mR

Beta:

240 mrem

Pan Am TLD 1970 to 

1972 NRDS Two-element LiF, Type 700 Quarterly

Photon: 

Exposure

Beta:

Absorbed dosea

1.25 c 1.23d

Photon:

15 mRh

Beta: 

15 mremh

Photon:

30 mR

Beta:

30 mrem

Panasonic 

802

1987 to 

2001 All of NTS

Four-element TLD – two Li2B4O7:Cu chips and 

two CaSO4:Tm chips. Filtration provided to 

determine gamma, deep dose, shallow dose, 

beta, and eye dose. NTS badge holder used 

from 1987

Quarterly

Personal dose 

equivalent, 

Hp(d)f

1.1 c 1.23d 15 mrem 30 mrem

Panasonic 

809

2001 to 

present All of NTS

Multi-element TLD containing four elements.

E1 - gamma sensitive 7Li2

11B4O7(Cu), enriched 

to 99.99% in 7

Li.

E2, E3 and E4 – neutron-sensitive 6

Li2

10B4O7(Cu) chips. Li-6 enriched to 95.33% 

and B-10 enriched to 94.64%. Elements are 

shielded with tin and cadmium on front and 

back, in various combinations.

Quarterly

Personal dose 

equivalent, 

Hp(d)f

1.1 c 1.23d 15 mremh 30 mrem
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 13 of 57

Table 6-1 (Continued). External dosimetry.

Neutron dosimeters

Dosimeter Dates Operation Description Issue and exchange Measurement 

Quantity Bias GSD MDL MMD

Kodak NTA 1966 to 

1979

Areas where 

potential for 

neutron 

exposure 

existed

Responds to neutrons with energies above 0.8 

MeV; range under near-ideal conditions 0.1 to 

few rem of neutrons; high gamma doses might 

mask neutron tracks

Exchanged monthly for 

general exposures and on 

exit from radiation areas 

for exposures likely to 

exceed 100 mR.

Dose equivalenti 5.c,j 1.52d 250 mremj,k 1,500 mrem

Albedo 

Dosimeter

1979 to 

1986

Areas where 

potential for 

neutron 

exposure 

existed

Hankins type albedo dosimeter. Consists of 

four pairs of TLD 600 and TLD 700 (6LiF and 7

LiF) in cadmium pillbox for thermal neutron 

suppression. High sensitivity to low-energy 

neutrons, with decreasing response as energy 

increases.

Monthly

Issued only to individuals 

with potential for exposure 

to neutrons Dose equivalenti 2.c,j 1.23d 20 mremj,k 120 mrem

Panasonic 

Track Edge 

Detector 

(TED)

1987 to

2001

Areas where 

potential for 

neutron 

exposure 

existed

Three pieces of CR-39 plastic used to detect 

neutrons with energies above 100 keV. 

Quarterly, except for 

limited number of workers 

(radiographers, well 

loggers, and personnel 

routinely entering HRAs).

Issued only to individuals 

with potential for exposure 

to neutrons

Dose equivalenti 2.5 c,j 1.23d 80 mremj,k 150 mrem

Panasonic 

809

2001 to 

present

Areas where 

potential for 

neutron 

exposure 

existed

Multi-element TLD containing four elements.

E1 - gamma sensitive 7

Li2

11B4O7(Cu), enriched 

to 99.99% in 7

Li.

E2, E3 and E4 – neutron-sensitive 6

Li2

10B4O7(Cu) chips. Li-6 enriched to 95.33% 

and B-10 enriched to 94.64%. Elements are 

shielded with tin and cadmium on front and 

back, in various combinations.

Quarterly, except for 

limited number of workers 

(radiographers, well 

loggers, and personnel 

routinely entering HRAs).

Issued only to individuals 

with potential for exposure 

to neutrons

Dose equivalenti 2. c,j 1.23d 40 mremj,k 80 mrem

a. Numerically equivalent to dose equivalent (Q = 1).

b. Based on uncertainty values provided by NRC (1989) and ORAU (2004a, Equation 4-1) for the 95th-percentile estimate.

c. Claimant-favorable assumption based on professional judgment.

d. Based on a claimant-favorable assumption of a sigma (standard deviation) equal to ±20% and 95th-percentile values enveloped by 2 sigma.

e. Assumed to be the same as for DuPont Type 502 and 508 films (NRC 1989).

f. Deep dose equivalent = Hp(10).

g. From NTS Fact Sheets (by M. DeMarre). 

h. Assumed to be the same as Panasonic 802 dosimeter.

i. Source: NCRP (1971).

j. See discussion in Section 6.3.4.3.

k. Based on reported values, corrected for potential energy dependent underresponse (NTA film and Panasonic TED) and ratio of the conversion coefficients for personal dose equivalent, 

Hp(10), to those for dose equivalent, H, (NCRP 1971) in the 100-2,000 keV energy range.
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• Dosimetry technology, which includes physical capabilities of the dosimetry system such as 

the response to different types and energies of radiation, in particular to mixed radiation fields

• Calibration of monitoring systems and similarity of the methods of calibration to sources of 

exposure in the workplace

• Workplace radiation fields that could include mixed types of radiation, variations in exposure 

geometries, and environmental conditions

An evaluation of the original recorded doses based on these parameters is likely to provide the best 

estimate of the actual doses received, including Hp(10) and Hp(0.07) as necessary, for individual 

workers with the least relative overall uncertainty.

6.3.1

When the testing program at NTS began in January 1951, the Los Alamos Scientific Laboratory [now 

named Los Alamos National Laboratory (LANL)] was responsible for administering the external 

dosimetry program. While contractor organizations and the military were involved in issuing and 

collecting badges for some of the early operations at NTS, LANL performed calibration, processing, 

and interpretation work. In July 1955, REECo assumed responsibility for most onsite radiological 

safety functions.

Administrative Practices

During the 7-yr period (1951 to 1958) of these testing operations, the allowable external exposure 

limits for occupational workers at NTS were generally consistent with National Council on Radiation 

Protection and Measurements (NCRP) recommendations. During the 1951 to 1952 test series, 

participants could receive up to 3 R of gamma exposure for a 13-wk period. Pilots and crew could 

receive up to 3.9 R of exposure. For the 1953 and 1955 series, workers could receive up to 3.9 R. 

Beginning in 1957 and through the 1958 series, the maximum permissible exposure for test 

participants was limited to 3 rem per 13-wk period and 5 rem per calendar year. 

After the Soviet Union ended the testing moratorium in the summer of 1961, the United States 

resumed the nuclear testing program at NTS on September 15, 1961. Most tests occurred

underground. After the signing of the Limited Test Ban Treaty on August 5, 1963, all tests were 

underground. 

The film badge dosimeter used from 1961 through 1965 at NTS was a modification of the 

premoratorium version. The film packet, a DuPont type 301-4, consisted of a type 508 low-range 

component (30 mR to 10 R) and a type 834 high-range component (10 to 1,000 R) wrapped with a 

28-mil (0.028-in.)-thick lead strip covering an area 0.5 in. by 1 in. on each side. The packet was in a 

4-mil (0.04-in.)-thick plastic bag sealed with colored tape to indicate the month of validity. The bag 

was clipped to the security badge, and all personnel entering NTS wore this dosimeter.

The film badges were exchanged monthly for all individuals and on exit from radiation areas if an 

exposure of 100 mR or more was measured (or suspected). In addition to film badges, self-reading 

pocket dosimeters were issued to persons entering controlled radiation exclusion (radex) areas, which 

were controlled locations at which an exposure was usually expected. The purpose of issuing pocket 

ionization (PIC) chambers to persons entering a radex area was to provide an action alert. A high 

reading of a PIC triggered the action of collecting and processing the personnel dosimeter being used 

at the time. PIC results were not used unless the personnel dosimeter had somehow been 

compromised. The PIC result would have been included in a special investigation of the incident. 

Although PIC results were only used as dose-of-record when film badge results were not available, 
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because of their typical overresponse characteristic, the dose reconstructor should use the 

claimant-favorable assumption that the PIC results are treated in the same manner as the film badge 

results.

The U.S. Atomic Energy Commission (AEC) Standard Operating Procedure, Nevada Test Site 

Organization (NTSO), Chapter 0524, Radiological Safety (AEC 1961), stated that the radiation 

exposure criteria for NTS personnel were 3 rem per quarter and 5 rem/yr. However, with the approval 

of the Test Manager, an NTS worker could receive as much as 12 rem/yr. 

In 1966, NTS began using a combination personnel dosimeter and security credential holder (Brady 

and Iverson 1968) to provide the increased personnel dosimetry capability necessary to meet 

radiation exposure problems associated with nuclear rocket testing and underground nuclear 

detonations. The holder was designed to accommodate a DuPont type 556 film packet, a fast neutron 

packet [containing Kodak nuclear track emulsion, type A (NTA) film], an identification plate, criticality 

accident components, the security credential, and a snap-type clip. The complete package could 

measure beta, gamma, X-ray, thermal neutron, fast neutron, high-range gamma, and high-range 

neutron exposures.

In March 1971, when the use of DuPont film ended, NTS dosimetry operations converted to Kodak 

Type Ill film packets. This two-component packet contained low-range (30 mR to 10 R) and 

high-range (10 to 800 R) films. The other components of the badge remained essentially the same. 

Between 1971 and 1979, NTA film was only issued to personnel who were identified as working in 

areas were neutron exposure was a possibility. This assessment was based on evaluation of the 

individual’s work assignment.

In 1979, NTS adopted use of the albedo neutron dosimeter, a TLD component system. This 

dosimeter was superior to NTA film because it was more sensitive and responded to a much wider 

neutron energy range. The albedo dosimeter was not part of the film dosimeter packet but, as with 

NTA film, was issued only to personnel who had a potential for exposure to a neutron source. It had 

its own holder, which had to be worn flush with the body at all times. This system was used from 

1979 through 1986.

Until 1966, there was no determination of shallow or skin dose from the film badges. With the 

introduction of a new multielement film dosimeter in 1966 until the conversion to TLDs in January 

1987, the open window (OW) was used to make a separate determination of shallow dose and deep 

dose. Shallow dose was determined by comparing the OW and closed-window readings and, 

therefore, did not include the penetrating photon component. This same type of determination was 

used for TLDs used by Pan Am at the NRDS from 1970 to 1972. Therefore, for these years and 

dosimeter types, the total skin dose should be determined by adding the reported deep (or 

whole-body) dose and the shallow (or beta) dose. On the NV-185 form used for dose records at NTS, 

the shallow dose was incorrectly called the "Skin of the Whole-Body" dose when it was, in fact, only 

the beta and low-energy photon components. Therefore, the total skin dose can be estimated based 

on the NV-185 form from 1966 to 1987 by adding the "Whole-Body Gamma" dose (column 5) to the 

"Skin of the Whole-Body" dose (column 7).

Until January 1987, film badges were exchanged routinely each month for all individuals and on exit 

from a radex area if there was suspicion that an individual had received 100 mR or more of exposure. 

In addition, personnel entering radex areas received self-reading PICs. Following the introduction of 

TLDs in January 1987, dosimeters were issued on a quarterly basis unless a particular job 

assignment indicated the need for more frequent issue and readout. Measured exposures were 
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added to the yearly and quarterly accumulated exposures. DOE (1982), Radiological Safety, set 

3 rem per quarter and 5 rem/yr as limits for exposure for occupational workers.

Beginning in 1987, with the introduction of Panasonic TLDs, the deep (or whole-body) dose and the 

shallow (or skin) dose were reported but the shallow dose included the penetrating photon component 

and can therefore be considered the total skin dose.

Table 6-1 lists the chronology of NTS external dosimetry measures beginning in 1961. The 

parameters of significance are the bias, geometric standard deviation (GSD), minimum detectable 

level (MDL), and the potential maximum missed dose (MMD).

The bias factors listed in Table 6-1, are the factors by which recorded values in the dosimetry records 

should be multiplied to provide the best estimate of the measurement quantity. The value of 

measurement quantity multiplied by the dose conversion factors in Appendix B of NIOSH (2002)

yields the organ dose. The bias values for neutrons include an estimate of potential underresponse 

due to the inherent energy response of the detector and spectral differences between the calibration 

sources and operational spectra (Section 6.3.4). Potential detector overresponse (e.g., for TLD 

albedos) maintains claimant-favorability.

The uncertainty in the best estimate of the measurement quantity is accounted for in the GSD. The 

GSD is discussed in more detail in ORAU (2004a), and defined in Equation 4-1 of that document 

presented as follows:

)()( .1 65585

1

50th percentile

95th percentile GSD =

The MDL is typically established at the point where the laboratory uncertainty of the readings at the 

95% confidence level is ±100% in normal distribution terms. The MMD is equal to one-half the MDL 

multiplied by the number of exchange or monitoring periods.

6.3.2

The dosimetry methods employed initially at NTS were adopted from techniques implemented at 

LANL from the beginning of the atmospheric weapons testing program (NRC 1989; Boone, Bennett, 

and Adams 1970). As the various atmospheric test series progressed, dosimeter configurations 

improved in relation to the radiation fields encountered, and the responsibility for dosimetry programs 

was delegated among the participating agencies and military organizations. In 1955, the site 

contractor, REECo, assumed responsibility for most onsite radiological safety functions; this included 

a site-wide service based on the use of film dosimetry for photons, betas, and neutrons. These 

methods evolved and eventually gave way to other methods including TLD and nuclear track 

detection or track etch detectors (TEDs). From the beginning, PICs were used if necessary to 

augment the passive dosimeters issued to NTS workers and visitors.

Dosimetry Technology

6.3.2.1 Beta/Gamma Dosimeters

During atmospheric test operations, beta/gamma dosimetry was accomplished with photographic film 

and subsequent densitometry measurements to quantify exposure. Individual dosimeters included 

multiple film types appropriate for the ranges of exposure considered and metallic filters to adjust film 

response to the associated photon spectra. Attempts to assess beta dose from comparisons of 
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filtered and open areas of exposed film were abandoned due to unknown relative contributions 

attributable to the photon and beta fields present.

In the late 1960s, trials were conducted with TLDs that led to implementation of a TLD-based 

personnel monitoring program at NRDS in 1970. Large-scale use of TLD dosimetry at NTS began in 

1987 with introduction of a Panasonic dosimetry system.

6.3.2.1.1 Photographic Film Dosimeters

Photographic emulsions of various types and in various holders (i.e., film badges) were used at NTS 

for personnel monitoring from the start of operations in January 1951 until 1987. The basic principles, 

theory, and practice of photographic film dosimetry for beta and photon radiation are well known and 

have been described in a number of standard texts and references (Becker 1966; Dudley 1966;

Ehrlich 1954, 1962; Kathren 1987; NRC 1989). The following paragraphs discuss factors related to 

film dosimetry that provide a general background to the interpretation and reconstruction of dosimetry 

results at NTS.

The term film badge, or film badge dosimeter, as used in this document, refers to the entire dosimeter 

issued to personnel, which typically consisted of a dental-size film packet housed in a holder of 

varying sophistication designed to improve the response characteristics and measurement 

capabilities. The film consisted of a plastic base covered on one or both sides with a layer of a 

suitable photographic emulsion. One or two pieces of film were wrapped in a light-tight paper 

package to comprise the packet, which in turn was placed in a holder containing metallic filter(s) to 

compensate for the photon energy dependence of the film. The sensitivity of the film to ionizing 

radiation was largely a function of the size of the AgBr grains in the emulsion. Because the typical 

photographic emulsion for dosimetry has an effective range of about 3 orders of magnitude, the 

packet typically contained two pieces of film with different sensitivities. The effective ranges of these 

films overlapped, which permitted an effective response range of about 5 orders of magnitude for 

penetrating photon radiations.

In principle, film dosimetry is simple; it consists of the determination of optical density or degree of 

blackening produced by exposure to radiation. The degree of blackening produced by the radiation 

incident on the film is determined in terms of the net optical density, the logarithm of which is typically 

plotted against the logarithm of exposure or dose to produce a calibration or response curve, or an 

algorithm with which the exposure registered by the film can be determined. The response of a film 

emulsion to photon radiation is not linear with dose or exposure. Near the lower (less than 50 mR)

and upper (about 1.5 to 5 R) limits of the range of the film, small increments of density represent 

relatively large changes in dose. Therefore, the uncertainty at the low and high ends of the dose 

response curve (i.e., net optical density vs. exposure), where the slope is quite shallow, is relatively 

large. However, the total uncertainty for high-energy exposures (more than about 150 keV to 3 MeV) 

for these dose ranges (1.5 to 5 R) should not exceed a factor of 2. For high-energy photon exposures

in the region of 50 to 1,500 mR (the steep portion of the dose response curve), the uncertainty is 

much less and should be within 20% to 30%. Dose reconstructors should consider these uncertainty 

values to be broad estimates, not exact values.

Although a number of different film types were used for dosimetry at NTS, they generally had similar 

characteristics and responses to beta and photon radiations. Uncertainties in dosimetry with these 

films are largely, if not exclusively, the result of external factors rather than differences in the films. 

The response of film to photon radiation is affected by a number of interrelated variables. From the 

standpoint of dose reconstruction at NTS, the most important of these are energy dependence,
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angular dependence, and effects of temperature and humidity. Because of the high relative atomic 

number (Z) of the AgBr in the film emulsion in relation to soft tissue, the response or energy absorbed 

per unit exposure of the film is a strong function of the energy of the exposing photons, rising steeply 

at energies below about 200 keV to a peak at about 30 keV, and then falling off steeply. At about 

30 keV, the energy of maximum response, the degree of blackening per unit exposure is about 30-fold 

greater than that for photons with energies of few hundred thousand to a few million electron volts. 

Because the relationship between exposure and soft tissue dose is approximately constant over a 

wide energy range, it was essential to compensate or correct for the energy dependence of the 

photographic response. This was accomplished by placing metallic filters over portions of the film to 

alter or flatten the energy dependence characteristics such that the film responded more like soft 

tissue over a wide energy range.

Although energy dependence can be a large source of error in film badge dosimetry, reasonably good 

results can be obtained by a four-element badge: one with an OW or no filter over a portion of the film 

packet, and high-Z, medium-Z, and low-Z filters over other portions of the packet. By observing the 

relationships of the degree of darkening under each filter area, and comparing these with exposures 

to known energies and doses of photon radiations, a reasonably accurate assessment of photon dose 

can be made over a wide range of energies. This is particularly true at NTS, where exposures were 

largely to photons with energies greater than a few hundred thousand electron volts at angles close to 

normal to the plane of the film. In general, errors in dose interpretation that result from energy 

dependence will result in overestimates of dose and are, therefore, conservative and 

claimant-favorable. However, given the single-element badges in use until 1966, some dose from 

lower energy photons, specifically those with energies below about 80 keV, could have been missed 

because of attenuation in the filter; therefore, the dose interpretation could be low. Given the photon 

energy spectra at NTS, any such loss of dose is likely to have been relatively low.

Photon calibration of film dosimeters was typically performed in terms of exposure in free air, with the 

plane of the film oriented normally to the direction of the incident photon beam. Initial calibrations 

were made with a high-energy photon source (e.g., 60Co or 137Cs) with an energy response by film 

similar to the energy spectra to which personnel were exposed in the field. After introduction of the 

advanced multi-element film badge in January 1966, calibration was conducted with X-ray sources as 

well as high-energy photons. This enabled interpretation of exposure to the film with a higher degree 

of accuracy over a broader spectrum of photon energies.

The response of film, whether in a bare film packet or in a holder, is a function of the angle of 

incidence of the exposing radiation on the plane of the film. Because the film is in a small flat sheet, 

an edge-on exposure (i.e., an exposure with the incident radiation parallel to the plane of the film) 

produces a different effect or optical density pattern than an exposure normal (at right angles) to the 

plane of the film. The effect is strongly dependent on both the energy of the exposing photons and 

the type of film holder or film badge, but is generally minimal at photon energies above a few hundred 

thousand electron volts and for angles of incidence ranging from about 30 to 150 degrees in relation 

to the plane of the film. However, for angles of incidence approaching parallelism (i.e., 0 degrees) 

with the plane of the film, the effect can be pronounced, and can lead to significant underestimates in 

dose. The problem, however, should be more or less minimal for exposures at NTS because these 

were (1) typically to high-energy photons and at angles close to normal with the plane of the film and 

(2) probably largely accounted for in the interpretation of the badge. 

Environmental conditions can significantly affect film badge results, and can result in large 

uncertainties and errors in dose estimation. Temperatures greater than 130o

F (50oC) can induce 

fogging, but several days at this temperature are necessary before fogging or increased density 

occurs. At higher temperatures, which could have occurred in closed vehicles or buildings at NTS, 
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fogging can result after shorter exposure times (Kathren, Zurakowski, and Covell 1966). If not 

corrected, high-temperature fogging always produces an overestimate and, therefore, a 

claimant-favorable dose interpretation. 

High relative humidity results in a fading of the latent image before development and in a decrease in 

the measured density of the film for a given dose. Latent image fading results in a low estimate of 

dose. However, studies have shown that latent image fading is not a problem until films have been 

worn for intervals exceeding 4 to 6 wk, and can be largely overcome by encasing the film packet in a 

polyethylene pouch, as was done for hot and humid conditions encountered during nuclear tests in 

the Pacific (Kathren, Zurakowski, and Covell 1966). Given the low-humidity conditions at NTS, latent 

image fading is unlikely to be of significance to film dosimetry.

Films were exchanged at varying intervals and, once collected, were developed under controlled 

conditions. Optical densities were determined with a densitometer, and doses were determined from 

calibration curves, which were log-log plots of net optical density versus dose obtained from a series 

of films exposed to known levels of photons. The use of control films at sites such as the badge 

house, where there was no expectation of a radiation field other than from background, permitted the 

inherent density of the film and any density attributable to background radiation (the so-called 

background fog) to be subtracted from the reading, thereby providing the net optical density 

attributable to occupational exposure. In practice, background films were used to zero the 

densitometer, which then gave a net optical density reading. 

6.3.2.1.2 Beta Dosimetry with Film Badges

Film response to beta radiation is accomplished with an unshielded portion of a film packet. 

Multielement film badges have an OW or unshielded portion that allows beta and photon radiation to 

reach the film. The beta dose is interpreted by removing the density attributable to a concomitant 

photon exposure as determined by the densities on those portions of the film under the various filters. 

The process is subject to large uncertainties, perhaps as much as a factor of 2.

Before 1966 film dosimetry was accomplished with a bare film packet partially covered by a lead strip. 

The exposed film was interpreted by attributing the density under the lead filter to photons. The 

density in the unshielded portion of the film was assumed to be due to both photons and betas, and 

was typically not measured. In those instances in which the unshielded portion was read, the beta 

dose was determined by simply subtracting the density under the lead strip from that in the unshielded 

portion of the film to obtain a net density. This procedure did not accurately determine beta dose 

because it attributed the entire net density of the unshielded portion to beta dose, ignoring the 

low-energy photon dose contribution (e.g., energies less than about 80 keV, of concern primarily to 

post-test drillback crews before 1965, and tunnel reentry crews in those situations when the tunnel 

was not vented before reentry), which did not penetrate the lead strip. It is estimated that as much as 

25% of the total photon dose could have been missed as a result of this attenuation (Kathren 2004; 

Coryell, and Sugarman 1951; Nelms and Cooper 1959). Therefore, the beta dose could have been 

overestimated. Given the spectra and mix of radionuclides encountered at the NTS, dose estimates 

made in this fashion are likely to be within a factor of 2 of actual beta dose and could slightly 

underestimate the overall photon dose. However, such dose estimates are likely to be unreliable and 

should be confirmed by other factors such as the beta:photon dose rate ratio obtained with monitoring 

instruments in the field, if such data are available. The value of a factor of 2 is an estimate of the 

range of uncertainty based on knowledge of the reported response characteristics of the dosimeters, 

and is presented for general information only.
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Multielement badges were introduced at NTS in 1966. A single high-Z metallic filter provided a more 

or less flat energy response for photons with energies above the uncertainty edge of the filter (about 

50 keV). The density under the filter was used to assess the dose to photons above this cutoff energy 

and to evaluate the doses to photons with energies below the cut off using densities under the other 

filters. Two filters – an OW and a low Z – were used to determine the beta dose. The response of the 

film under these filters was approximately the same for photons, but the low-Z plastic filter essentially 

removed all the betas. The reading from the low-Z filter, which was considered to be attributable to 

photons only, was subtracted from the density under the unshielded portion, which had a comparable 

response to photons and also was responsive to beta radiation. If necessary to make a beta dose 

estimate using the film badge results from shallow dose estimates from 1966 through 1986, dose 

reconstructors should double the reported value to ensure claimant-favorability and to account for 

uncertainties.

For external dose reconstruction, a positive indication of beta exposure in a dose record is considered 

to be due to betas with energies above 15 keV.

6.3.2.1.3 Evaluating Photon Doses with Film Badge Dosimetry

Different film badge dosimeters were used at NTS from its start in 1951 until 1987, when personnel 

film dosimeters were replaced with TLDs. In 1989, the National Research Council (NRC) published 

an exhaustive evaluation of film badge dosimetry in atmospheric weapons testing, effectively covering 

the period through the cessation of atmospheric testing in July 1962 (NRC 1989). The NRC 

considered only photon doses and evaluated the dose from each test series individually in terms of 

bias and uncertainty. This evaluation considered a number of specific factors, including laboratory 

factors, which was an inclusive term covering calibration, film development and processing; 

sensitometry; radiological uncertainties, including the source term photon energy spectrum; 

environmental effects; and conversion from exposure to deep dose.

On the basis of its evaluation, the NRC published tables for each test series listing bias and 

uncertainty values as well as a table listing individual film badge exposure values for approximately a 

dozen intervals less than or equal to 200 mR, the corresponding best estimate of Hp(10) in millirem, 

and the 95% confidence limits for Hp(10). For exposures above 200 mR, a simple equation was 

provided to permit easy conversion from exposure to Hp(10). The conversion from exposure to 

Hp(10) can be calculated by 

Hp(10) = 0.77E 

where Hp(10) is the personal dose equivalent, penetrating, in millirem (or rem, as appropriate), and E

refers to film badge exposure in milliroentgen (or roentgen, as appropriate).

In 1966, a multielement film badge of advanced design was introduced at NTS (Brady and Iverson 

1968). This badge was a considerable improvement over earlier badges and had a reported accuracy 

of ±20%. Certainly, this level of uncertainty at the 95% or even 99% confidence level was achievable 

under laboratory conditions. However, under field conditions, factors such as backscatter, wearing 

location, and angular dependence could result in additional uncertainty, as reflected by the GSDs 

listed in Table 6-1. 

Although the basis for determination of personal dose equivalent from exposure values reported in 

individual monitoring records has been clearly documented as indicated above, it is also important 

that one-to-one correlation is maintained with the film badge record so that the claimant can see 

evidence that original records were used in dose reconstruction and to avoid the perception that the 
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recorded doses were lowered as part of the reconstruction. Therefore, the dose reconstructors 

should use the original reported exposure values for the reconstruction, unless there is evidence that 

these values are somehow in error.

6.3.2.1.4 Estimation of Beta Doses from Film Badge Results

Early dosimeters worn at NTS could not discriminate reliably between beta and photon radiation, or 

between photons in various energy ranges. In fact, given the types of film dosimeters, reliable beta 

dosimetry was not possible until the introduction of the NTS film badge in 1966 (Brady and Iverson 

1968). Beginning in 1966, the multielement film badge permitted stripping out the density produced in 

the unshielded or OW portion of the badge by photons and interpreting the remaining density, if any, 

as beta dose. This process can be performed manually with a set of calibration curves made at 

different photon energies and a beta calibration curve or with an algorithm developed from such a set 

of calibration curves. Regardless of the technique used, the resultant beta dose estimate had a 

considerable degree of uncertainty, although it was reasonably reliable with a total uncertainty 

typically not exceeding a factor of 2 (Becker 1966, p. 102; Kathren and Larson 1969). Therefore, in 

the absence of data suggesting otherwise, a reasonable and claimant-favorable estimate of beta dose 

can be made from film badge results (i.e., from shallow dose estimates from 1966 through 1986) by 

doubling the reported value.

6.3.2.1.5 Thermoluminescent Dosimeters

6.3.2.1.5.1 NRDS Thermoluminescent Dosimeters

The first routine use of TLDs at NTS began in 1970. Starting in February 1966, Pan Am used TLDs at 

the NRDS as part of the site effluent monitoring program. These dosimeters contained a 

calcium-fluoride phosphor bound to a helically wound wire in an evacuated glass tube and were ideal 

for the intended purpose but unsuitable for personnel dosimetry.

Other installations had successfully demonstrated the applicability of LiF TLDs for personnel 

dosimetry. A brief investigation was performed and proposals submitted for institution of a LiF TLD 

program at NRDS in 1967 to complement the REECo-supplied personnel film dosimetry service.

On July 1, 1968, Pan Am initiated a TLD program at the NRDS with the cooperation of the Health 

Services Laboratory of the AEC Idaho Operations Office (IDO). This laboratory supplied LiF TLD 

chips and provided readout services and considerable consultation based on its extensive LiF TLD 

experience at the National Reactor Testing Station (NRTS). From July 1, 1968, until July 1970 NRDS 

personnel who were not likely to receive a significant exposure received TLDs. All reported 

exposures were in the background range (0 to 30 mrem). Routine use of TLDs began in July 1970

and continued until the NRDS ceased operation at the end of 1972.

Beginning in July 1970 (and repeated every quarter thereafter), a minimum of one set of dosimeters 

was exposed on the Pan Am Calibration Range to a total dose(s) of from 100 mR to 5 R. Each set of 

dosimeters consisted of 1 Pan Am R-chamber, 6 REECo film, and 12 IDO TLD chips (6 to be read out 

at Pan Am and 6 at IDO). The Pan Am Calibration Range was crosschecked against Pan Am 

R-chambers, which were calibrated by the National Bureau of Standards (NBS; one set was sent to 

NBS annually). If the difference between the range and the R-chamber exceeded ±6%, the data 

points were retaken.
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A calibration factor was calculated for each type of dosimeter as follows:

R

DoDx CF − =

where:

CF = calibration factor

Dx = mean indicated dose from the set of six exposed dosimeters

Do = mean indicated dose from the set of six unexposed dosimeters

R = R-chamber indicated dose.

As long as the average calibration factor over the dose range of interest fell in the range of 0.9 to 1.1 

(±10%), the results were documented but no further action was taken because this level of accuracy 

was considered adequate. If the average calibration factor fell outside this range, the parties 

concerned were notified and another set of data was taken. In general, the relative accuracy of both 

measurement systems, as indicated by the body of data considered, is adequate for personnel 

dosimetry purposes.

In the case of Pan Am-read TLDs, a calibration factor was always necessary due to the relative 

response of the reader. The appropriate calibration factor was employed as follows:

( ) −×= DoDiCFDc

where:

Dc = calibration factor

Do = actual dose

Di = indicated dose

Boone, Bennett, and Adams (1970) contains additional 

information on calibrations and calibration comparisons between 

Pan Am, IDO, and REECo.

Figure 6-1 shows the security badge and enclosed TLD insert. It 

was a plastic holder in which two LiF dosimeters were placed. 

One dosimeter was shielded by the tantalum-cadmium filter area 

of the present badge. The other dosimeter was unshielded. 

The two dosimeters were held in place by retainers built into the 

insert (for pliable LiF-Teflon discs) or hinged retainers (for

extruded LiF chips). The insert was punched with a binary code 

that identified security badges for issuance and collection. 

Finally, the bottom left-hand portion of the dosimeter insert was 

coded to enable security guards to determine if dosimeters were

being worn for the proper period.

6.3.2.1.5.2 REECo Thermoluminescent Dosimeters

With the advent of DOE requirements to restrict personnel 

exposures to as low as reasonably achievable (ALARA) and with 

emphasis on accurate dosimetry at low doses, REECo Figure 6-1. Security badge for 

NRDS TLD (Pan Am 1967).
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Environmental Sciences Department personnel began evaluating TLD systems and neutron 

dosimeters in the early 1980s to replace the film badge and neutron TLD.

After evaluating several dosimetry systems, the Environmental Sciences Department determined that 

the Panasonic 802 TLD and the neutron TED were the best combination for NTS exposure conditions. 

These were put into use January 1, 1987. The security credential holder was redesigned to 

accommodate both dosimeters (Figures 6-2 and 6-3).

Figure 6-2. REECo badge holder (REECo 1990).

Figure 6-3. REECo security badge holder with 

Panasonic UD-802 dosimeter. 

DOE Order 5480.11, "Radiation Protection for Occupational Workers” (DOE 1988), established the 

radiation protection standards and program for DOE, its contractor personnel, and other occupational 

personnel to protect workers from ionizing radiation. It defined policies and procedures required to 

operate DOE facilities and conduct activities to keep personnel exposures well below the limits set by 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 24 of 57

Order 5480.11 to meet ALARA goals. The site-specific Nevada Operations Office/Yucca Mountain 

Project (YMP) Radiological Control Manual (known as the RadCon Manual; DOE 1994) describes 

radiation protection standards and program requirements as they relate to the NTS and YMP 

organizations. The Nevada Test Site Radiation Protection Program, published in May 1995 (REECo 

1995a), demonstrated compliance with DOE rules for protecting individuals from ionizing radiation.

Specifically, the NTS external dosimetry program consists of the following:

• Site-wide RadCon Managers to coordinate exposure reduction with NTS organizations.

• Each RadCon Manager was responsible for ALARA within their own organization.

• A Radiological Work Permit system to establish exposure controls and personnel monitoring 

requirements for work areas on NTS.

• Procedures that identify, monitor, and control locations where personnel can encounter 

ionizing radiation from contaminated work areas.

• A requirement that all personnel entering an area where there is a potential for exposure wear 

TLDs. Additional monitoring devices are worn where there is a potential for exposure to the 

skin, the lens of the eye, or extremities, or exposure to nonuniform and neutron radiation 

fields.

• A requirement that supplemental dosimeters, such as direct-reading and electronic-alarm 

dosimeters, be worn for entries into High Radiation Areas (HRAs) and Very High Radiation 

Areas (VHRAs).

• A requirement that a dose assessment be conducted for each instance of a lost, damaged, or 

contaminated personnel dosimeter.

• A DOELAP-accredited external personnel dosimetry system.

The four-element Panasonic UD-802 TLD was the primary dosimeter for routine use issued to all 

monitored personnel until 2001, when it was replaced by the Panasonic 809 dosimeter, which also 

contained four elements (Table 6-1). The Panasonic UD-807 TLD is used to monitor personnel 

working in situations where the likelihood of exposure to an extremity is significantly greater than 

exposure to the whole body.

In most instances, dosimeters (i.e., TLDs) were processed quarterly. Some personnel working in 

locations where high exposures were more likely to occur exchanged their dosimeters monthly. The 

NTS monitoring program was designed to ensure that personnel exposures were kept below the 

annual limit of a total effective dose equivalent (TEDE) of 5 rem.

The Panasonic UD-802 TLD was designed to identify the type and energy of detected radiation, so 

the prescribed tissue depth dose equivalents could be determined accurately. In specific, the 

four-element Panasonic 802 was intended to measure the following:

• External photon radiation from 0.010 to 1,000 rem

• Gamma energy range of 0.010 to 10 MeV

• Beta radiation from 0.030 to 1,000 rad

• Beta energy range from approximately 0.30 to 10 MeV
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Table 6-2 lists TLD element composition, filtration, and radiation type.

Table 6-2. Panasonic UD-802 dosimeter configuration. 

Element Phosphor Filtration

Rad. type & dose 

equivalent

E1 Li2B4O7:Cu Teflon/polyester, 18 mg/cm2 Gamma, beta, shallow dose

E2 Li2B4O7:Cu Teflon/polyester & Plastics, 70 mg/cm2 Gamma, beta, eye dose

E3 CaSO4:Tm Teflon/polyester, ABS Plastics, and Blak Ceramic,

645 mg/cm2

Gamma

E4 CaSO4:Tm Teflon/polyester, ABS and Lead, 1042 mg/cm2 Gamma, deep dose

The dosimeters were calibrated against known exposures to provide an accurate transition from 

measured exposure to dose equivalent. Dosimeter calibration factors were normalized to the 

corrected readings from run calibration factor (RCF) dosimeters processed with the field dosimeters. 

The computer program separated the calibration dosimeters, used the known RCF exposure value, 

and calculated RCFs that were applied to the remainder of the dosimeters in the run. The RCFs kept 

a reader in calibration over long periods and maintained consistency among different readers.

6.3.2.1.5.3 Energy Response

Panasonic UD-802 and UD-809 TLD elements have well-known energy responses. Like many 

radiation detection devices, there is an energy (about 20 keV) below which radiation does not deliver 

sufficient energy to the TLD element for detection. As photon radiation energy increases, the 

Li2B4O7:Cu elements have a relatively flat response, while the CaSO4:Tm elements overrespond 

between about 50 and 200 keV. The higher CaSO4:Tm response is due primarily to the effective Z of 

14.4 for CaSO4, compared to 7.3 for Li2B407:Cu. The lithium borate effective Z is much closer to the 

effective Z for tissue (7.4), which relationship is important to photoelectric effect interactions. The 

response of both types of elements is relatively flat to about 5 MeV, where the CaSO4 response 

increases slightly due to its higher effective Z and, in this case, pair production interactions. Figure

6-4 shows typical photon energy response curves for Li2B4O7 and CaSO4 TLD elements. Figure 6-5 

shows the response for the UD-802 TLD, including filters. The first element in the UD-802 TLD has a 

total filtration density depth of about 18 mg/cm2

, making it sensitive to beta particle radiation with 

energies of 100 keV or more. The second element responds to beta radiation with energies above 

about 300 keV. Because beta particles have a low linear energy transfer (LET) and the Li2B4O7 is 

nearly equivalent to tissue, the beta energy response is relatively flat and similar to the photon 

response shown in Figure 6-5 for photon radiation (starting at 0.1 MeV for element E1 and 0.3 MeV 

for element E2). 

Figure 6-4. Typical photon energy response curves for 

Li2B4O7 and CaSO4 TLD elements (REECo 1995b).
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 26 of 57

Figure 6-5. Panasonic UD-802 energy 

responses (REECo 1995b).

NTS algorithms identify radiation types and energies and then apply factors necessary to obtain a 

correct dose equivalent from a TLD reading. Thus, the energy response characteristics of the UD-802 

TLD, over the useable range of the TLD, are automatically factored into the dose equivalent 

determination. The validity of the algorithms for energy response correction is verified through the 

satisfactory completion of DOELAP accreditation for photon and beta particle radiation in several 

different radiation energy categories.

Unlike the Panasonic UD-802, the UD-809 introduced in 2001 has four lithium borate (Li2B4O7(Cu)). 

One uses lithium, enriched to 99.99% in 7

Li and 11B, both of which have a negligible neutron 

response. The other three employ neutron-sensitive 6

Li2

10B4O7(Cu) chips, with 6

Li enriched to 95.33% 

and 10B enriched to 94.64%. These two isotopes have high n,alpha cross sections. The elements are 

shielded with tin and cadmium on front and back, in various combinations.

6.3.2.1.5.3.1 Dose Measurement

The background, element correction factor (ECF), and RCF corrected measurements from TLD 

elements provide an indication of the dose received by the person wearing the dosimeter. However, 

because TLDs are not truly tissue equivalent in relation to dose equivalent, algorithms are used to 

convert the dosimeter response to a dose equivalent value at the specific depth in tissue.

The NTS algorithms were developed by irradiating TLDs mounted on a phantom of tissue substitute to 

specific radiation quantities and types. The dosimeter measurement was compared to the dose 

equivalent calculated by knowing the type and energy of the radiation. This process was repeated 

many times, until enough data were collected to develop equations and relationships between a 

dosimetric reading and the dose equivalent at the specified depth in tissue. Published factors for 

converting radiation exposure to dose equivalent were used in the development of the algorithms. 

Table 6-3 lists exposure-to-dose equivalent conversion factors specified in the DOELAP Standard for 

National Institute of Standards and Technology (NIST) reference photon fields.

Before 1987 unexposed control films were processed together with the personnel dosimeters. 

Readings from the control films were subtracted from the dosimeter readings to obtain a net reading 

for determining exposure (REECo 1961, 1962). Beginning in 1987, with the introduction of TLDs, the 

procedure was continued with TLD background dosimeters.
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Table 6-3. Photon exposure to dose correction factors 

for NIST reference radiations. 

NIST reference 

radiation

Conversion factor (mrem/R)

Shallow (0.07 mm) Deep (10 mm)

Filtered X-rays

M30 1.08 0.45

S60 1.15 1.07

M150 1.41 1.47

H150 1.41 1.41

K-fluorescence X-ray (keV)

16 1.08 0.38

24 1.07 0.47

34 1.07 0.99

43 1.28 1.30

58 1.47 1.54

78 1.61 1.72

100 1.59 1.74

Cesium-137 photon (keV)

662 1.03 1.03

6.3.2.1.5.3.2 Detection Limits

Panasonic UD-802 TLDs have been tested to determine their lower limit of detection (LLD). The LLD 

is the minimum evaluated dose equivalent for which the readout value of a dosimeter is significantly 

different (at the 90% confidence level) from the mean readout of unirradiated dosimeters. LLD is a 

detection limit based on the standard deviation of background measurements and a 5% chance of 

reporting a false positive value.

DOE (1986, Chapter 3) summarizes the procedure for determining the LLD for the DOELAP 

personnel dosimetry systems. It provides two alternative equations to calculate LLD. Although DOE 

(1986) requires determination of the LLD for accredited dosimetry systems, no performance criteria 

are applied to these results.

The method used to determine the LLD for the NTS personnel TLD system is consistent with the 

method described in DOE (1986). Use the following equations for determining the LLD: 

[ ( ) ] [ ( ) ] ' 2 2

D1 p 0 += .12 75 o 1/1/1/ +−+ BSBHStL

and

2 = pD StL 0

where:

LD1 = LD where the probability of reporting a false positive or false negative result is 5%.

LD2 = LD corresponding to a 5% chance of reporting a false positive.

Tp = distribution factor for n-1 degrees of freedom and a probability value of 0.95 = 1.68488.

So = [{∑(Xio – Ho)

2

}/(n-1)]1⁄2 = standard deviation of measurement

Ho = Ho’ = (Xio)/n = mean dose value

Xio = unirradiated dosimeter value background dose

n = number of dosimeters number of measurements
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B = bias for DOELAP categories

S = standard deviation for DOELAP categories

The LLD is primarily a function of the standard deviation in background dose rates. Table 6-4 lists 

selected NTS background dose equivalent rates. The values can vary over time. However, the 

temporal variation has been small since the end of atmospheric testing. 

Table 6-4. Background dose equivalent rates. 

Location

Dose equivalent rate (mrem/day)

Shallow Eye Deep

NTS Area 12, Camp 0.375 0.311 0.283

NTS Area 6, Building 2 0.327 0.246. 0.211

NTS Area 6, Building 50 0.255 0.182 0-150

NTS Area 23, Building 650 0.271 0.199 0.168

Test Range Complex, DODa Facility 0.249 0.193 0.169

Las Vegas, REECo Highland Building 0.181 0.139 0.121

Mean dose rate 0.276 0.212 0.184

Standard deviation 0.067 0.059 0.057

a. DOD = U.S. Department of Defense.

Table 6-5 lists bias and standard deviation values for DOELAP categories determined during the 1990 

accreditation process. The LLD selected as most appropriate for NTS operations is LD2, which is the 

most applicable value for determination of a dose likely to be higher than background levels and 

which increases the probability that a dose will be evaluated as positive. Therefore, the LLD values of 

0.1 mrem/day for deep and eye dose equivalents and 0.13 mrem/day for shallow dose are used for 

NTS personnel TLD measurements. The LLD for a specific TLD reading is determined by the TLD 

processing software and is the appropriate LLD value multiplied by the number of days between TLD 

annealing and reading. For the routine calendar quarter, the LLD is about 9 mrem for the eye and 

deep dose equivalent and about 12 mrem for shallow dose. A nominal value of 15 mrem is 

recommended for dose reconstruction (Table 6-1).

Table 6-5. NTS Panasonic UD-802 detection limits for DOELAP categories (mrem/day). 

Category

Shallow dose Eye dose Deep dose

LD1 LD2 LD1 LD2 LD1 LD2

I. Low-energy photons – X-ray high dose N/Aa N/A N/A N/A 0.21 0.21

II. High-energy photons – high dose N/A N/A N/A N/A 0.21 0.21

IIIa. Low-energy photons – X-ray general 0.31 0.13 0.23 0.10 0.23 0.10

IIIb. Low-energy photons X-ray – plutonium environments 0.33 0.13 0.25 0.10 0.24 0.10

IV. High-energy photons 0.28 0.13 0.21 0.10 0.21 0.10

V. Beta particles – general 0.27 0.13 N/A N/A N/A N/A

a. N/A = not applicable.

6.3.2.1.5.3.3 Calibration

Several Quality Assurance (QA) operations are performed on new dosimeters before they are 

approved for field use. The numerical coding on each TLD label is verified to ensure accurate 

tracking. Before use, and annually thereafter, dosimeter phosphor elements are calibrated by 

determining an ECF for each element. In accordance with applicable procedures, a dosimeter with an 

ECF out of tolerance is removed from use.

TLDs are calibrated with the NTS 137Cs gamma source calibration range. The 137Cs source is 

traceable to NIST through a secondary standard calibrated ion chamber. An integrating electrometer, 

calibrated annually for NIST traceability, is used to verify the accuracy of each TLD and TLD reader 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 29 of 57

calibration exposure. All temperature and barometric pressure measuring instruments used for the 

calibration range are calibrated annually to NIST traceable standards.

Monthly Calibration of Gamma Radiation Fields at Inner and Outer Dosimeter Support Rings (REECo 

1995c) outlines procedures for performing gamma source exposures. The results of the exposures 

are used to ensure proper performance of the gamma sources, provide required exposure rate 

information to upgrade source calibration, and document exposure trends for quality control (QC). 

6.3.2.1.5.3.4 Quality Assurance

The QA program was established and has been maintained through adherence to QC procedures and 

practices. As required by DOE rules for protecting individuals from ionizing radiation, the program 

includes internal audits at intervals of 3 yr or less. The QC personnel for the external dosimetry 

program monitor and test program operations, data records, and performance. The effectiveness of 

the QA program is demonstrated through satisfactory completion and maintenance of DOELAP 

accreditation.

Dosimeters are carefully tested before being put into use, and their performance is routinely 

monitored. Acceptable processing and recording equipment calibration and operation is verified by 

internal QA reviews and by participation in external assessments. For example, REECo Health 

Protection Department procedure, Thermoluminescent Dosimeter Quality Assessment (REECo 

1995d), outlines the methods for assessment of the performance and adequacy of TLD issuing, 

processing, and reporting techniques.

The NTS external dosimetry program was one of the early programs accredited by DOE. 

Accreditation was initially requested in 1989 and updated in 1990. The onsite assessment was 

conducted in 1991, and accreditation was granted in 1992. DOELAP accreditation requires that 

dosimetry operations satisfy specific standards for accuracy of measurements, records, reports, and

QA activities. 

The NTS external dosimetry program has maintained accreditation in the following DOELAP 

categories:

I. Low-energy photon (high dose)

II. High-energy photon (high dose)

IIIA. Low-energy photon

IIIB. Low-energy photon (plutonium)

IV. High-energy photon

VA. Beta

VI. Neutrons (unmoderated, 252Cf)

VII. Mixtures

III + IV

III + VA

IV + VA

III + VI

IV + VI

The accredited categories are based on possible accident scenarios and probable operational 

exposure conditions at NTS and include all DOELAP categories except two, which were omitted 

because:
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• Category VB, "Beta Particles - Special" – uranium exposure environment does not exist at 

NTS. Therefore, TLDs have not been calibrated for dose measurements for beta particle 

radiation from natural or depleted uranium slabs.

• Category VI, "Neutron 252Cf (moderated)" — the unmoderated neutron category more closely 

approximates the neutron energy spectra in the NTS occupational environment.

The bias values and uncertainties in Table 6-6 are based on DOELAP performance testing for TLD 

systems within 30% at an approximate 95% confidence level (REECo 1995b).

Table 6-6. Bias values and uncertainties for NTS DOELAP accreditation categories (REECo 1995b). 

Category

Bias(B) and standard deviation (S) (mrem)

Deep dose Shallow dose

B S B S

I. Low-energy photons (X-ray) – high dose 0.036 0.065 N/Aa N/A

II. High-energy photons – high dose 0.046 0.070 N/A N/A

IIIa. Low-energy photons (X-ray) – general 0.001 0.117 0.03 0.097

IIIb. Low-energy photons (X-ray) – plutonium environments 0.022 0.142 0.011 0.134

IV. High-energy photons 0.099 0.048 0.039 0.058

V. Beta particles – general N/A N/A 0.091 0.036

a. N/A = not applicable.

6.3.2.2 Neutron Dosimeters

A small fraction of the workers at NTS had potential for exposure to neutrons. For workers with a 

possibility of neutron exposure, personnel neutron dosimeters were used to monitor exposure. 

Potential sources of neutron exposure at NTS were:

1. Direct production from a nuclear detonation

2. Spontaneous fission and subcritical multiplication in fissile materials (e.g., 235U, 239Pu)

3. Isotopic sources such as initiators and calibration sources

4. Reactor testing

Table 6-7 lists areas and operations where neutron exposure could have occurred at NTS. Personnel 

who did not work in these areas or were not directly involved in operations during the periods 

indicated were not issued personal neutron dosimeters. If workers were unmonitored for neutrons, 

based on NTS personal dosimeter issue practices, it is highly unlikely that neutron exposure occurred. 

Dose reconstructors, therefore, should not consider missed neutron dose for unmonitored personnel.

No single individual had access to areas in which there was potential for neutron exposure. Work in 

these areas always involved pairs of workers; two knowledgeable persons had to be involved with 

potential operational activities. This group probably numbered in the hundreds and was limited to 

persons who worked on specific tasks in specific areas. Neutron doses, for the most part, were low. 

In addition, neutron exposure was not possible without a concomitant gamma exposure; neutron-togamma dose ratios were less than or equal to 1. However, dose reconstructors must not ignore a 

positive indication of neutron exposure from an individual’s dose or work records.

Table 6-7. Areas and operations where neutron exposure was possible. 

Area Operation Neutron sources Beginning Final

5 Low-level waste site TRU waste Before 1962 Present

6 Nuclear device assembly Fission neutrons Before 1962 1992

25 NRDS and BREN tower calibrations Fission neutrons & neutron 1966 1973
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and operation sources 252Cf. PuBe, AmBe

26 PLUTO Reactor (nuclear-powered 

ramjet engine) 

Fission Neutrons & neutron 

sources 252Cf. PuBe, AmBe 

Mid 1960 Late 1960

27 Nuclear explosive assembly using 

special nuclear material

Fission neutrons & neutron 

sources 252Cf, PuBe, AmBe

Before 1982 1975

Various Down-hole well logging 238PuBe or 252Cf isotopic 

sources

Before 1962 Present

Various Neutron detection instrument 

calibration facilities

238PuBe or 252Cf isotopic 

sources

Before 1962 Present

Exposure to neutrons from nuclear detonations, while a theoretical possibility, was for all practical 

purposes nonexistent. Strict measures were taken to ensure that personnel were not exposed to the 

prompt radiation from the detonation and, at the locations where personnel could have been exposed 

to prompt neutrons from the blast, air attenuation and similar attenuation mechanisms would have 

reduced the energy and the fluence of the neutrons to negligible levels. Therefore, reconstructors 

should ignore exposure from this source.

Small neutron exposures were possible in the vicinity of test shapes or other significant quantities of 

fissionable materials, but levels in this case would also probably be negligible and, in any case, 

overshadowed by the gamma radiation field. An exception would be the few individuals associated 

with final assembly, arming, and firing of test weapons. Assembly operations were the only times that 

workers, in close proximity with the weapons components, were exposed to neutrons emanating from 

bare pits.

The specific workplace neutron fields for specific types of nuclear weapons components are 

classified. Unclassified information on neutron spectra from nuclear weapons components is not 

available, but there are two sources, both with significant components above 2 MeV. Before about 

1960, nuclear weapons could have contained 210PoBe initiators and 238PuBe (DOE 1997) with a higher 

neutron dose component relative to the measured photon dose. However, these were not used after 

1960.

For exposures after 1960, if neutron dose information is not specifically available for those involved 

with final assembly and arming operations, photon exposure records, together with neutron-to-photon 

dose ratios can be used. The neutron-to-photon ratios can be derived from the experience at Pantex 

where weapons assembly operations were conducted. Analysis of dose records for each Pantex 

worker with a positive neutron dose greater than 50 mrem for the period 1993 to 2003 yields a 

geometric mean of 0.81 and GSD of 1.51. An upper 95th-percentile value of 1.6 should be used for 

the neutron-to-photon dose ratio (ORAU 2004b).

Assuming that 100% of the neutron doses were delivered by neutrons in the 0.1 to 2 MeV range is 

claimant-favorable. Although there are neutrons with higher energies at the NTS, which are more 

penetrating, the probability of causation for deeper organs, such as the liver, is much larger in the 

0.1 to 2 MeV energy range than any other energy group, thus offsetting the higher dose at depth for 

the more energetic energy ranges.

A more significant potential source of neutron exposure was from isotopic neutron sources such as 

238PuBe. These sources were used in specific activities such as instrument calibration and well 

logging. Only a few highly trained and specialized individuals, however, had access to such sources. 

The final source of potential neutron exposure was reactor test operations. These occurred in specific 

areas (Areas 25 and 26) designated for that purpose. Again, the number of individuals with the 

potential for neutron exposure was relatively small, and the potential for neutron exposure was further 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 32 of 57

mitigated by the fact that simultaneous gamma exposures were much greater and thus likely to be the 

governing factor for exposure control. Personnel were always at locations remote from the reactors 

where neutron exposures would be low or negligible during reactor operations or test periods.

The dose from neutrons is a function of neutron fluence and neutron energy. Fewer so-called fast 

neutrons (i.e., neutrons with kinetic energies exceeding several hundred electron volts) are required to 

produce a given level of dose in comparison with slow or moderated neutrons. This is reflected in the 

use of quality factors or neutron weighting factors applied to the absorbed dose to arrive at dose 

equivalent. These energy-dependent factors have changed in the last 50 yr as a result of new 

information on the relative biological effectiveness (RBE) of neutrons. Table 6-8 summarizes 

historical changes in the factors used in the United States to adjust measured absorbed doses for the 

higher radiological impact associated with neutron exposures. They are shown in the neutron energy 

groups used in dose reconstruction. 

Table 6-8. RBE, quality factors, or weighting factors for neutrons. 

Neutron 

energy

(MeV)

Dosimetry 

guidelinea 

RBE

NCRP (1971)

quality 

factorsb

Average 

quality factors 

used at NTS

ICRP (1990) 

neutron 

weighting 

factor wr

Factor to be 

applied to NTS 

neutron dose

2.5 × 10-8 3 2

2.35 5 2 

1 × 10-7

10

2

1 × 10-6 2

1 × 10-5 2

1 × 10-4 2

1 × 10-3 2

1 × 10-2 2.5 5.38 10 2

1 × 10-1 7.5

5 × 10 10.49 20 2 -1 11

1 11

2 10

7.56 10 1.35

2.5 9

5 8

7 7

10 6.5

14 7.5

20 8

40 7 Not applicable 5 Not Applicable

60 5.5

a. Trilateral meeting in 1949 radiation protection guidelines (Fix, Gilbert, and Baumgartner 1994). 

b. Recommendations of NCRP Report 38 (NCRP 1971).

c. ICRP Publication 60 (ICRP 1990).

The RBE was used until 1971, after which quality factors were used. The ICRU introduced the 

radiation weighting factor wr in 1990 as part of its definition of Hp(10). Although wr has not been 

adopted in U.S. regulations, it is necessary to convert from neutron doses obtained using NCRP 

Report 38 quality factors to Hp(10) (NCRP 1971). Figure 6-6 shows the ratio of conversion 

coefficients for personal dose equivalent, Hp(10, 0°) to conversion coefficients for dose equivalent, H

(NCRP 1971). Using this figure, the last column of Table 6-8 lists factors to make this conversion for 

each of the four energy groups for dose reconstruction at NTS (none of the neutron-producing 

facilities generated a significant number of neutrons above 20 MeV). If the neutron spectra are not 

known, dose reconstructors should treat the exposure as having come from neutrons in the 0.1 to 2 

MeV range, and should use a factor of 2 to convert from NCRP Report 38-based values to Hp(10). 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 33 of 57

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02

Neutron energy - MeV

Hp(10)/H

Figure 6-6. Ratio of fluence to dose conversion coefficients for personal dose 

equivalent Hp(10, 0°) to conversion coefficients for dose equivalent H. 

The neutron spectra at NTS would have been generally unmoderated and rich in fast neutrons, so 

much so that doses from thermal neutrons would have been trivial. However, in some operational 

situations, considerable scattering could have occurred, resulting in some softening of the spectra. 

Section 6.3.4.3 contains details on neutron spectral characteristics.

6.3.2.2.1 Film Dosimeters

From 1966 to 1971, nuclear track emulsions in the form of Kodak NTA film packets were used for 

personnel fast neutron dosimetry at NTS. The NTA film packet was incorporated in the newly 

designed combination personnel dosimeter and security credential along with a DuPont type 556 film 

packet for beta photon monitoring, an identification plate, criticality accident components, the security 

credential, and a snap-type clip (Brady and Iverson 1968).

The NTA film packet consisted of a single thick-coated film of dental size wrapped in several layers of 

light-tight paper. Fast neutrons interact with the hydrogen in the emulsion and film, producing recoil 

protons that, having kinetic energy and being charged, travel through the emulsion, creating a string 

of exposed individual grains of AgBr along their path. On development, these show up as tracks of 

grains of developed AgBr. Track length is a function of the energy of the recoil proton and the angle 

of travel with respect to the plane of the film. Fast neutron dose is determined by direct visual 

counting, usually via a microscope, of the number of proton recoil tracks in a predetermined (for 

statistical purposes) number of microscope fields, usually 100. A countable track must have a length 

of at least three grains (i.e., there must be three grains in a row for a track to be registered).

Tracks can be produced in NTA film emulsion either by protons produced by the 14N(n,p) reaction with 

low-energy neutrons (less than 10 eV) or by direct recoils from energetic neutron interactions with 

hydrogen in the film. The relatively low thermal neutron fluences, the low (n,p) reaction cross-section, 

and the large fluence of thermal neutrons per millirem compared to fast neutrons (2,200 compared to 

7.5) rendered thermal neutron dosimetry impractical.

Based on theoretical considerations, a minimum neutron energy of about 450 keV is needed to 

produce a proton recoil track, although in practice the minimum detectable energy or threshold energy 

was about 800 keV (Figure 6-7). Thus, NTA film is essentially insensitive to neutrons below 500 to 

800 keV, depending on the quality of processing. Therefore, it is suitable only for occupational 

environments in which the majority of the dose comes from neutrons with energies above 1 MeV, or in 

which the fraction of neutrons below 1 MeV is reasonably well known. It is important to calibrate NTA 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 34 of 57

film dosimeters with a source having a neutron energy spectrum similar to that in which the individual 

could be exposed. Initially, a PuBe neutron spectrum was used for calibration. The PuBe neutron 

spectrum has an average energy of about 4 MeV and is considerably richer in fast neutrons than the 

fission spectrum (with average energy of about 1 MeV), and so is likely to result in a 

Figure 6-7. NTA film neutron energy response (IAEA 1990).

calibration factor (i.e., tracks/cm2

-n) that would underestimate the fast neutron dose. Any moderation 

of the fission spectrum encountered under field conditions would further exacerbate the 

underestimate.

NTA film has a number of severe limitations that lead to large uncertainties in neutron dosimetry. As 

noted, the response of NTA film was highly dependent on neutron energy and angle of incidence with 

respect to the plane of the film (Cheka 1954; Lehman 1951; Kathren, Prevo, and Block 1965; Kathren 

1967). Another limitation of NTA film is its use dose range. The LLD for fast neutrons typically 

corresponds to a dose of about 100 mrem. For fast neutron fluences corresponding to a few rem, the 

track density becomes so great that accurate dosimetry becomes difficult if not impossible.

NTA films respond to other radiation qualities (photons, betas, etc.). Concomitant photon dose, which 

results in film blackening, can render track counting difficult, introduce errors and, if the optical density 

produced by the photon exposure is sufficiently great, can obscure proton recoil tracks altogether. 

Track counting itself is questionable; different persons given identical sections of an exposed NTA film 

to count often produce highly variable track counts. It has been long recognized that, in the dosimetry 

laboratory, human factors associated with reading large numbers of neutron films under a microscope 

can significantly affect neutron dosimetry results. A research conducted at the Rocky Flats Plant in 

1994 reevaluated neutron doses for selected plutonium workers. This research indicated that the 

original evaluations of films could have contained significant errors and that the resulting neutron 

doses could be significantly higher or lower than the doses actually received. The degree of variation 

is a function of the track length and the personal style of the individual doing the counting. 

Uncertainty from this source alone could easily be as great as a factor of 2.

The latent image produced by proton recoil tracks in NTA film is highly susceptible to fading before

development. Fading is a function of time after exposure and is particularly severe in high humidity 

(as much as 75%/wk) but can be minimized and largely controlled if the films are sealed in a 

moisture-proof pouch before use. Humidity-induced latent image fading is probably not a significant 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 35 of 57

source of uncertainty or dose underestimation at NTS because of the low humidity, although some 

latent image fading is likely if the wearing interval exceeds a month.

Given these limitations, neutron dose estimates made with NTA film are likely to have a high degree 

of uncertainty and generally will underestimate the actual dose from fast neutrons. As noted in 

Table 6-1, this leads to the need for a relatively high bias for neutron data obtained with NTA film 

when detailed information about the neutron fields is not available (see Section 6.3.4.3).

6.3.2.2.2 Thermoluminescent Albedo Dosimeters

The NTS albedo dosimeters adopted in 1979 were based on the design of Hankins (1977). They 

were used until 1987 and consisted of four TLD chips each of 6

LiF and 7LiF, placed in a cadmium box 

to suppress response to thermal neutrons. Albedo dosimetry depends on the detection of low-energy 

neutrons reflected from the body (albedo neutrons) with a thermal neutron detector. Normally, a TLD 

with 6

LiF (TLD 600) is used to detect neutrons, while a companion 7LiF (TLD 700) detector that is 

insensitive to neutrons is used to subtract the photon/beta contribution.

The primary advantages of TLD albedo dosimeters are the high sensitivity compared to NTA film and 

the availability of automated TLD readers for rapid reading. The primary disadvantage is that the 

energy response does not match the personal dose equivalent response, so they are highly energydependent (Figure 6-8). The energy response can be improved slightly by dosimeter encapsulation, 

as used in the NTS dosimeter design. However, the response is still highly dependent on the neutron 

spectrum.

Figure 6-8. Neutron energy response of NTS albedo dosimeters

(IAEA 1990).

Until January 1987, dosimeters were routinely exchanged each month for all individuals and on exit 

from a radex area if it was suspected that an individual received 100 mR or more of exposure. In 

addition, personnel entering radex areas were issued self-reading PICs. Measured exposures were 

added to the yearly and quarterly accumulated exposures.

For albedo dosimeters, neutron fields can be put in four categories based on their relative spectra 

characteristics: (1) reactors, linear accelerators, and accelerators for medical therapy; (2) nuclear fuel 

fabrication areas; (3) radioactive neutron sources; and (4) high-energy accelerators with little or no 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 36 of 57

shielding. Within a neutron spectral class, neutron response relative to Hp(10) does not vary by more 

than a factor of 2. The large energy dependence is still a big disadvantage. The advantage to TLD 

albedo detectors, compared to film and to a lesser degree Columbia Resin-39 (CR-39), is that they 

detect neutrons of all energies and have simple automatic TLD readouts.

Calibration curves have been established for working areas that can reduce workplace-dependent 

changes of albedo response within ±30%. Depending on the neutron field, the lowest detectable dose 

using albedo TLDs varies from 5 to 20 mrem. Albedo dosimeters can be combined with TEDs for 

separate measurement of fast neutrons. In a combination detector, the albedo detector serves as the 

basic neutron detector for screening.

6.3.2.2.3 Track Etch Detectors

The gamma-insensitive CR-39 TED was introduced for personnel neutron dosimetry at NTS in 1987 

and was based on the dosimeter developed at Lawrence Livermore National Laboratory (Hankins, 

Homann, and Westermark 1987; Hadlock et al. 1988). The CR-39 TED offered better energy 

response characteristics for occupational monitoring than the TLD albedo dosimeter. The response 

curve is relatively flat, between 0.1 and 4 MeV (Figure 6-9). Once a foil has been properly etched, 

acquiring information is a nondestructive process.

Figure 6-9. Neutron energy response of NTS TED (REECo 1995b).

The TED is a dosimetry-grade polymer called Columbia Resin-39. When fast neutrons interact with 

the plastic, submicroscopic damage trails are created. These trails can be enlarged using chemical 

etching techniques to form tracks visible under a 400-power microscope. The tracks can be made 

more visible using electrochemical etching that causes breakdown to form subsurface trees that are 

easily counted under low magnification.

The NTS TED dosimeter consists of three CR-39 foils (i.e., individual dosimeter pieces), cut from 

0.025-in. (0.0635-cm)-thick sheets covered on both sides with 0.005-in. thick polyethylene film. They 

are heat sealed under an opaque blister to a plasticized card. A bar code label is applied, and (for 

personnel service) the assembly is placed in a cavity in the NTS Personnel Dosimeter.

A three-step etching process is used to develop the damage tracks that resultfrom neutron 

interactions. After removing the polyethylene film, the track detectors are first etched for 45 minutes 

at 60°C in 6.5N KOH. This is followed by the electrochemical etching step. An alternating potential of 

3 kV is applied across the track detector in the etch bath at a frequency of 60 Hz for 3 hours. The 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 37 of 57

resultant tracks are amplified, forming “trees” under the track detector surface. A third step using 3 kV

but at 2,000 Hz produces tracks that are more uniform and easily recognized as tracks. The final step 

is referred to as the blow-up stage of the process. This produces well-defined round or elliptical holes 

large enough to be seen and counted with low (4 power) magnification. The tracks are recorded with 

a television camera interfaced to a commercial bacteria colony counter that can be used to count 

several standard 0.09-cm2 fields. The dose equivalent is determined by a software program that 

converts the number of foil net tracks to a dose equivalent.

The TEDs were initially calibrated with an unmoderated 252Cf source at Pacific Northwest 

Laboratories. Secondary calibration is provided by the NTS 238PuBe neutron source. Secondary 

calibration-check TED foils are processed with each batch of personnel monitoring foils to determine 

the RCF for the processing run. Primary and secondary calibration data are recorded in units of 

millirem per track per square centimeter in the TED program. The tracks from the calibration check 

foils processed with each batch of foils are compared to the primary and secondary calibration data to 

determine the dose equivalent conversion factor for the process batch.

Background exposure is subtracted from TED results to eliminate the TED exposure that is not part of 

the individual’s occupational dose. The natural or background neutron radiation level is extremely low 

and, therefore, NTS uses the standard practice of using control TEDs. Control TEDs are prepared 

along with batches of TEDs for issue and are retained in the low-background dosimetry operations 

facility until the corresponding personnel TEDs are processed. Control, calibration-check, and 

personnel foils are processed together, and the number of tracks in control foils is subtracted from the 

number of tracks in personnel and calibration-check foils. Then calibration-check results are used to 

determine the correct dose equivalent calibration factor.

6.3.2.2.3.1 Dose Algorithms

The CR-39 TED is known to have a nonlinear response, with the observed or measured dose 

equivalent being a slight overresponse in comparison with the delivered dose equivalent for low 

doses, then changing to an underresponse as the dose increases. Because the nonlinearity can vary 

with CR-39 processing methods, specific data for the NTS TED system were collected. A series of 

TED exposures was made using an unmoderated PuBe source and appropriate conversion factors 

relating the source to an equivalent unmoderated 252Cf source. Two TEDs (six foils) each were 

exposed to 12 dose equivalents between 130 to 6,505 mrem. The results ranged from 126 to 2,930 

mrem, as shown in Figure 6-10.

TED algorithms were developed by exposing many TEDs to calibrated neutron exposures and 

comparing processed results with calculated and known dose equivalents. Irradiations ranging from 

about 100 to 7,000 mrem were used to determine the shape of the curve relating true dose equivalent 

to observed dose. Linear regression analysis of the data pairs (true and observed dose) showed that 

a second-degree polynomial fit the relationship between true dose equivalent and indicated for 

uncorrected dose equivalent. That polynomial, shown below, is the algorithm used by the TED 

program to calculate the neutron dose equivalent HT from the TED uncorrected dose equivalent HI: 

HT = 0.77HI + 0.000459HI

2 
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Figure 6-10. Electrochemically etched CR-39 dose response

(REECo 1995b).

HI in the above equation is calculated by the TED program from the number of net tracks and the 

millirem to track calibration factors.

Because the Panasonic UD-802 TLD has a very slight response to neutrons, two algorithms are 

necessary to determine dose equivalent when exposure includes photons and neutrons. One 

algorithm corrects TLD response for neutron interference; the other converts TED results to dose 

equivalent. The deep dose equivalent is the sum of the deep dose from photons and the dose from 

neutrons. Therefore, in all neutron exposure cases, a TLD evaluation is performed to determine the 

deep dose from photons.

6.3.2.2.3.2 Lower Limit of Detection

Control foils processed with each TED processing are used as the basis for background subtraction 

for the TED system. The mean background dose equivalent to the control foils and the variance 

among the control foil dose equivalents are calculated for each processing. The background dose 

equivalent is determined for each processing because it can differ among foil sheets and within foils 

from the same sheet. Therefore, the TED background correction rate is not a single value, but a 

variable determined with each processing.

The method for determining the LLD for the TED system, with a variable background correction, was 

determined from Chapter 7 of NCRP Report 58, A Handbook of Radioactivity Measurements 

Procedures (NCRP 1978). The appropriate equations are:

LD1 = 2.71 + 4.65√VB

and

LD2 = 2.32√VB
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 39 of 57

where

VB = background variance in units of dose equivalent squared (millirem squared).

The LD2 equation above was selected as the most appropriate LLD for the TED system. The selection 

reasons are to same as those discussed above for the TLD system.

The control foil variance (VB) increases with TED foil age and typically ranges between 5 and 

25 mrem. Therefore, the TED system LLD is normally between 5 and 12 mrem. However, based on 

practical experience with these detectors, a value of 25 to 30 mrem is more realistic. That range of 

limits is reasonable for any TED issue period because it is primarily dependent on foil age and is only 

slightly dependent on the issue period. Considering the possibility of long-term variations and other 

contributing factors, potential energy-dependent underresponse to moderated spectra, and the ratio of 

Hp(10) to dose equivalent H, dose reconstructors should use a value of 80 mrem for the MDL 

(Table 6-1).

6.3.3

6.3.3.1 Photons

Calibration

In the early 1950s, film dosimeters were calibrated to free-air exposures using 226Ra and 60Co sources 

with source strengths traceable to the NBS, which is now NIST. Exposures were quantified from 

inverse-square relationships over a fixed exposure interval at various distances and for exposures at 

fixed distances over various time intervals. Different approaches were used to compare dosimetry 

results, including processing badges exposed to calibration fields to badges exposed during test 

operations. In addition, collocated badges of different types and ionization chambers were exposed 

simultaneously during some test events to provide in-field calibration for realistic radiation spectra.

6.3.3.2 Neutrons

Neutron calibrations have been performed with isotopic sources, reactor beams, or accelerators. 

However, virtually all routine or practical radiation protection calibrations employ radioactive sources. 

The preferred neutron production mechanisms are either alpha-neutron reactions with beryllium or 

boron, or spontaneous fission (252Cf). On rare occasions, gamma-neutron reactions have been used, 

but the associated high-intensity gamma fields generally make them unacceptable for calibration of 

dosimeters and instruments that have any degree of gamma sensitivity.

Table 6-9 summarizes the properties of commonly used isotopic sources from the International 

Organization for Standardization (ISO 2001). Probably the first isotopic source to be used was 

226RaBe. However, its neutron production was accompanied by a high photon emission. As 239Pu, 

238Pu, and 241Am became available, they became the alpha-emitting radionuclides of choice. Because 

these radionuclides have similar alpha energies, the resultant neutron energies are also similar and 

can be considered nearly identical for radiation protection purposes. In particular, 238PuBe has been 

in use at NTS for several years.

Occupational neutron fields at the NTS are commonly due to fission origin neutrons. Therefore, 252Cf 

is particularly attractive as a calibration source because it offers better inherent spectral simulation 

than the alpha,n sources. However, the unmoderated californium neutron spectrum is significantly 

harder than most reactor spectra. Beginning in the 1980s, spheres filled with deuterated water (D2O) 

came into common use as moderators to soften the spectrum of neutrons from the 252Cf source in the 

center.
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Table 6-9. Characteristics of ISO reference radiation radionuclide 

sources used for neutron dosimeter calibration. 

Source

Half-life

(yr)

Flux average energy

(MeV)

241AmBe (α,n) 432.7 4.4

252Cf spontaneous fission 2.65 2.4

252Cf in 30-cm-diameter D2O sphere 2.65 0.54

6.3.4

The radiation production characteristics at the NTS have been outlined in ORAU (2004b). The 

potential for external radiation exposures arises primarily from the fission and activation products 

handled at the NTS. In addition, there is limited potential for neutron exposure from handling TRU 

radionuclides, isotopic sources, and reactor operations.

Workplace Radiation Fields

6.3.4.1 Photon Energies

The residual radiation field following detonation of a nuclear weapon consists of radiation from fission 

products, activation products, and unfissioned uranium or plutonium. During atmospheric testing of 

fission and fusion devices, differences in photon fields of residual radioactivity from detonations were 

observed. These differences are caused by the relative abundance of a few radionuclides that were 

produced in each atmospheric test. For example, a low-altitude detonation of a fusion weapon 

induces large quantities of activation products emitting high-energy gamma rays that dominate the 

residual radiation spectrum for the first few days following the detonation. In contrast, a low-altitude 

detonation of a fission weapon produces large quantities of fission products that emit a wide range of 

photon energies. With either type of weapon, depending on the design, there can be a large amount 

of activity from the 239Np produced, which can dominate the spectrum for several days.

Although the residual radiation intensity depends on a number of factors that can vary from test to 

test, relatively few radionuclides, common to all tests, contribute to the major part of the photon 

spectrum. The relative abundance of each of these radionuclides determines the spectrum. In all 

cases, the photon field is from photons with energies between approximately 100 keV and 2 MeV. 

There is very little contribution from photons with energies less than 100 keV with the exception of 

scattering from large area sources. In those cases, the scattered radiation was determined to have 

an energy of approximately 75 keV and to have contributed as much as 10% of the overall photon 

spectrum.

For external dose reconstruction, unless exposure to fresh fallout (first few days) is known to have 

occurred (early reentry teams), use the claimant-favorable assumption that photon energies are 

between 30 and 250 keV. At times less than 1 wk after detonation, it would be reasonable and still 

claimant-favorable to assume that 75% of the photon dose was from photons with energies above

250 keV (Kathren 2004; Coryell and Sugarman 1951; Nelms and Cooper 1959). For the period 1961 

to 1966 when the multielement dosimeter was introduced, a contribution amounting to 25% of the total 

dose should be included in the range of 30 to 250 keV to account for low-energy photons attenuated 

by the lead filter that covered a portion of the film.

6.3.4.2 Beta Particle Energies

This section applies only to whole-body external doses from beta radiation as could be encountered 

from a fallout field or residual fission product activity at a work location. It does not apply to situations 

in which beta radiation is deposited directly on the skin or clothing, but rather concerns in-air doses to 
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which a worker could have been exposed. Although there is a large and highly useful body of 

literature pertaining to skin doses from deposition of beta-emitting contamination on the skin, there is 

a paucity of available information on external (i.e., in-air) beta doses from fallout fields or residual 

fission product activity. Concerns were largely limited to whole-body exposure.

There are about 90 possible fission fragments, which, on the average, decay through three stages to 

yield a total of about 250 individual fission-produced radionuclides, most of which are beta particle 

emitters.2

Unlike gamma rays and neutrons, whose attenuation in matter is exponential, beta rays have a finite 

range in matter determined by the energy of the beta particle. Because beta particles emitted by 

radioactive species are not monoenergetic, the range is usually specified in terms of the maximum 

energy of the beta particle spectrum. For the beta produced by the decay of fission and activation 

products, the maximum energy typically does not exceed 3 MeV

 Many of these beta-emitting fission product radionuclides have very short half-lives – on 

the order of a few seconds or less – and quickly decay into other nuclides. Others decay over longer 

periods. Therefore, the composition and beta energy spectrum of fallout is not constant; it changes

over time. In addition to beta-emitting fission products, activation products are produced by a nuclear 

detonation, although their contribution to overall beta dose is relatively small. Unlike photons, which 

are emitted with discrete quantum energies, beta emission is characterized by a distribution of 

energies ranging from zero to a maximum value that is commonly used to characterize the spectrum. 

The average energy of a beta spectrum is typically about one-third of the maximum energy, and the 

total energy produced by beta particles from fission products is essentially the same as that from 

fission product gamma rays, namely about 7 MeV per fission. 

3

The range of a beta particle of known energy can be rigorously calculated by integration of the –dE/dx

equation, which shows that the rate of energy loss is a complex function directly proportional to the 

number of atoms per cubic centimeter of absorber and to the Z of the absorber, and is exponentially 

related to the kinetic energy of the beta particle. However, a reasonably good approximation of the 

range R of a beta particle of energy E (in megavolts-electron) in any medium can be calculated in 

terms of density thickness (in milligrams per square centimeter) from the following empirical equation:

, and the range of a 3-MeV particle 

in air is approximately 36 ft. Therefore, an individual at a greater distance greater than 36 ft from a 

fallout field would not receive an external dose from beta radiation associated with the decay of 

radionuclides produced by fission and fission-produced activation products. Similarly, an individual 

exposed to beta particles with energies below 70 keV would receive no beta dose to the skin because 

beta particles with energies below 70 keV have insufficient energy to penetrate the cornified outer 

layer of the skin.

R = 412E(1.265-0.0954 lnE) 

which holds over the energy range 0.01 to 2.5 MeV. For energies above 2.5 MeV, the appropriate 

empirical relationship is:

R = 530E – 106

A somewhat less exact but still reasonable approximation for the range of a beta particle in units of 

grams per square centimeter is to divide the beta particle energy by 2: a 3-MeV particle would have a 

 

2

Although the term beta particle can refer to both positron and negatron emission from the nucleus of an excited atom, as 

applied to fission produced radionuclides it refers only to negatron (i.e., electron) emission because no known fission 

products emit positrons.

3

The few exceptions are fission products with very short half-lives, which can be ignored for all practical purposes.
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 42 of 57

range of 1.5 g/cm2, which corresponds to 36 ft, approximately the same value (within about 3%) 

obtained using either of the above equations.

Because of attenuation of beta radiation in air, the external dose associated with a fallout field is 

strongly dependent on distance. As one moves away from the beta field, more and more of the lower 

energy particles in the spectrum reach the end of their range and no longer contribute to the dose; 

1 m of air is sufficient to attenuate all betas with energies below about 400 keV. And, as already 

noted, 36 ft of air is sufficient to absorb virtually all beta radiation associated with a fallout field, so 

beyond this distance there is zero beta dose. In contrast, as one moves closer to the fallout beta field, 

the external beta dose rate rises more rapidly than the concomitant gamma dose rate. Hence, the 

beta-to-gamma dose rate ratio is a function of distance from the fallout field; at distances greater than 

36 ft, the ratio is zero because there is no beta dose, as noted above. Calculations by Sondhaus and 

Bond (1955) on persons highly exposed to Marshall Islands weapons fallout contamination from the 

Bravo event indicate that, relative to the average whole-body gamma dose from a fallout field, the 

corresponding beta dose would range from about a factor of 2 at the head to about a factor of 10 at 

the bottom of the feet (bare) (standing on the contamination). At 1 m above the ground, the beta–togamma dose rate ratio was 3. This is consistent with the data obtained onboard a contaminated ship 

and described in a 1954 memorandum obtained from the DOE archives in Las Vegas which, only 

33 hr after detonation, described a beta:gamma ratio of 10 at the deck level, 8 at knee level, and 

about 5 at film badge level (about 4 ft above the deck).

The above results appear inconsistent with the observations of Kulp and Dick (1960), who reported 

beta dose rates at 1 in. from a contaminated aircraft surface to be typically 10-fold greater than 

gamma dose rates at 1 ft from the surface. When a correction is made for the distance by using 1/d

relationship for an infinite plane source to obtain the gamma dose rate at 1 in., the beta:gamma dose 

rate ratio falls to less than 1. However, the Kulp and Dick measurements involved vastly different 

geometries with opportunities for beta shielding and therefore might not be applicable to the type of 

exposures expected in the field at NTS. A more rigorous approach was taken by Broido and Teresi 

(1961) who evaluated the surface dose rate from beta radiation from a fallout field in relation to the 

gamma dose, and found the beta dose to be 13 times that of the gamma dose. Comparing the beta 

surface dose to the gamma dose at 1 m, Broido and Teresi observed a ratio of 40. In a study by 

Black (1962) in which actual measurements were made of doses to troops crawling through a 

contaminated fallout field, the beta:gamma dose ratio was approximately 7.

In these studies, the concern was the dose from contamination deposited on the skin, a situation 

rather different from the external exposure situation that could be encountered in the field at NTS and 

that would likely produce a higher beta:gamma dose ratio. Although numerous theoretical and 

empirical studies have been made of gamma radiation doses above fallout-contaminated ground, 

there have been few such studies of beta radiation doses. Review of operational monitoring logs from 

several Plumbbob detonations indicate, on the basis of survey meter readings, a beta:gamma dose 

rate ratio in the range of slightly greater than unity to a maximum of about 3, regardless of distances 

above the field (ranging to 3 ft), and for times ranging from 1 to 4 wk after the detonation. 

Beta:gamma dose rate ratios did not appear dependent on dose rate but, according to a study by 

Barnaby (1957), vary with time. At 1 m above the surface of contaminated ground, Barnaby found a 

beta:gamma dose ratio of 27 at 1.5 hr after detonation, dropping rapidly to a minimum of about 3 at 10 

to 20 days after detonation, and then increasing again to about 30 at 400 days after detonation. This 

is not necessarily inconsistent with observations at the NTS, which were mostly made from a few days 

to a few weeks after detonation.

Given the above discussion, for beta dose reconstruction in the absence of personnel monitoring data 

(i.e., exposures before 1967), the dose reconstructors should assume that a reasonable client-
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 43 of 57

favorable estimate of beta dose can be obtained by ratio with gamma dose. To this end, some 

provisional values are presented below (with the caveat that they are subject to revision as additional 

information is obtained). For beta doses incurred in a mixed beta-gamma radiation field a few days to 

a few weeks after detonation for which the gamma dose is known, a beta:gamma dose ratio of 5 

would seem to be reasonable, while allowing a measure of overstatement to ensure 

claimant-favorability. In other words, the beta dose would be 5 times the measured gamma dose. If 

the mixed field exposure occurred less than 2 to 3 days after detonation, or more than 6 wk after 

detonation, a beta:gamma dose ratio of 10 is provisionally suggested. These provisional dose ratios 

take no credit for shielding or attenuation of beta particles by clothing (which could reduce them by a 

factor of 2 or more) and assume a 1-m distance from the fission product field with no overburden, 

leaching, vegetation, or other material or weathering action that would reduce the ratio, however likely 

this could be.

For external dose reconstruction, a positive indication of beta exposure in a dose record is considered 

to be due to betas with energies greater than 15 keV.

6.3.4.3 Neutron Energies

As indicated in Section 6.3.2.2, primary NTS operations with neutron exposure potential have been:

• Low-level waste

• Nuclear device assembly

• NRDS and BREN (Bare Reactor Experiment Nevada) tower calibrations and operation

• PLUTO reactor (nuclear-powered ramjet engine) 

• Nuclear explosive assembly using special nuclear material

• Down-hole well logging

• Neutron detection instrument calibration facilities

The neutron source spectra from these operations were either from fission of uranium and transuranic 

nuclides or isotopic sources involving n,alpha reactions, such as 238PuBe or 241AmBe. Moderation or 

scattering of the source neutrons results in a “softening” of the spectrum, with an increase in the 

fraction of low-energy neutrons present. This, in turn, results in (1) lower RBEs, quality factors, or 

radiation weighting factors, and (2) changes in the effective calibration factors for the energydependent dosimeters and instruments used for personal neutron dose assessment.

There is no indication that neutron spectral measurements were conducted at the NTS for radiation 

protection purposes. However, IAEA has compiled extensive neutron spectral functions, instrument 

and dosimeter response functions, and dosimetric quantity response functions (IAEA 1978, 1990, 

2001). Because the source spectra for NTS operations are limited to those discussed above, 

adequate simulation can be obtained by selecting the proper neutron production mechanism and 

moderation and scatter conditions from the spectral catalogs that best simulate those for NTS 

operations. For example, Godiva is actually the reactor used on the BREN tower.

Figures 6-11 to 6-13 show the effect of scattering and moderation on the neutron spectra. Figure

6-14 and Table 6-10 further illustrate the impact of spectral softening from scatter and moderation; 

that is, the fraction of Hp(10,0) due to neutrons with energies less than the indicated value. For 

example, with the “hard,” unmoderated 241AmBe source spectrum, about 74% of the dose is due to 

neutrons above 2 MeV. The PuBe spectrum at 1 m is probably more characteristic of an 

unmoderated calibration source spectrum, with about 63% of the Hp(10) due to neutrons above 

2 MeV. In contrast, consider the much softer 252Cf source spectrum in a 15-cm diameter D2O sphere 

where 9% of the dose comes from neutrons below 0.1 MeV and only 31% of the dose comes from 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 44 of 57

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Neutron energy - MeV

Fluence/unit log energy (lethargy)

Bare 252Cf

Bare 252Cf

w/room

scatter

Figure 6-11. Lethargy neutron spectra for 252Cf calibration source with and without room scatter (IAEA 

2001).

1.E-03

1.E-02

1.E-01

1.E+00

1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Neutron energy - meV

Fluence/unit energy (lethargy)

252Cf - ISO

252Cf in

10.16 cm

CH2

252Cf in

22.86 cm

CH2

252Cf in

22.86 cm

CH2 w/10.2

cm Fe cover

Figure 6-12. Lethargy neutron spectra for 252Cf with various thickness moderators (IAEA 2001).
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 45 of 57

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Neutron energy - MeV

Fluence/unit energy (lethargy)

241AmBe

w/o room

scatter

241AmBe

w/room

scatter

Figure 6-13. Lethargy neutron spectra for an 241Am-Be calibration source with and without room 

scatter (IAEA 2001).

0

0.2

0.4

0.6

0.8

1

1.2

1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01

Neutron energy, En - MeV

Fraction of Hp(0,10) <En

252Cf - ISO 252Cf D2O Mod. - ISO

AmBe - ISO TRU Plant, Pu Repro. Plant, Heavily shielded

Bare 238PuO2 at 100 cm Godiva Spectrum

Figure 6-14. Fraction of Hp(10,0) due to neutrons with energies less than En. 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 46 of 57

Table 6-10. Fraction of Hp(10,0) for neutrons with energies less than En. 

Energy

En - MeV

252Cf

ISO

Bare 252Cf 

w/o 

shadow 

cone

252Cf D2O 

Mod. - ISO

252Cf in D2O 

w/o shadow 

cone

241AmBe 

ISO

AmBe 

w/o 

shadow 

cone

PuBe 

at 1m

TRU Plant, Pu 

Repro. Plant, 

heavily 

shielded

Bare 

238PuO2

at 100 

cm

Godiva 

spectrum

1.00E-02 0.000 0.004 0.018 0.030 0.000 0.003 0.002 0.028 0.012 0.000

1.99E-02 0.000 0.005 0.027 0.039 0.000 0.003 0.002 0.032 0.013 0.000

5.01E-02 0.001 0.008 0.050 0.064 0.000 0.005 0.003 0.046 0.014 0.001

1.00E-01 0.003 0.017 0.089 0.106 0.001 0.010 0.005 0.080 0.021 0.008

1.99E-01 0.017 0.042 0.155 0.181 0.009 0.028 0.027 0.162 0.058 0.043

3.98E-01 0.072 0.114 0.228 0.283 0.037 0.070 0.066 0.321 0.225 0.158

5.01E-01 0.110 0.158 0.269 0.328 0.053 0.099 0.088 0.392 0.331 0.226

7.94E-01 0.236 0.292 0.361 0.462 0.099 0.169 0.158 0.549 0.544 0.394

1.00E+00 0.325 0.382 0.407 0.516 0.127 0.212 0.242 0.625 0.646 0.492

1.25E+00 0.430 0.486 0.476 0.579 0.158 0.260 0.310 0.694 0.744 0.591

1.99E+00 0.662 0.708 0.668 0.775 0.262 0.376 0.373 0.814 0.842 0.781

3.98E+00 0.927 0.939 0.916 0.961 0.660 0.683 0.744 0.931 0.911 0.958

5.01E+00 0.968 0.969 0.963 0.969 0.813 0.806 0.871 0.956 0.930 0.983

7.94E+00 0.999 0.988 0.998 0.992 0.995 0.969 0.997 0.989 0.977 1.000

1.00E+01 1.000 0.993 1.000 0.994 1.000 0.991 1.000 0.999 0.987 1.000

1.58E+01 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000

neutrons with energies greater than 2 MeV. Table 6-10 provides a detailed listing of the fraction of 

Hp(10,0) due to neutrons with energies less than the indicated value.

Figure 6-13 and Table 6-10 indicate that the Hp(10, 0) contribution from neutrons with energies less 

than 10 keV can be ignored for the purposes of NTS dose reconstruction. The primary Hp(10)

contributions fall in the energy ranges 100 keV to 2 MeV and 2 MeV to 20 MeV.

It is, therefore, recommended that dose reconstructors should make the claimant-favorable 

assumption that neutron energies are between 100 and 2,000 keV.

Table 6-10 also indicates that NTA film could underestimate doses by 55% or more because the 

detection threshold for the film is about 0.8 MeV. Therefore, it is recommended that the bias for NTA 

measurements include a factor of 2.5 to account for potential dosimeter underresponse in softer 

neutron fields. The bias should also include correction for the ratio of dose equivalent H to personal 

dose equivalent Hp(10) conversion coefficients of 0.5 in the 0.1 to 2 MeV energy range (Table 6-10). 

These two factors combine to yield a recommended bias for NTA film based dose assessment of 5 

(Table 6-1), if the spectrum is not well known.

TEDs have a lower energy threshold than NTA film. Therefore, fewer neutrons will be undetected. 

However, with a threshold of about 100 keV, the underestimate due to energy response should be 

less than 10% to 15%, so dose reconstructors should use an overall bias of 2.5, including dose 

conversion coefficient correction.

On the other hand, albedo dosimeters calibrated with isotopic sources in low-scatter conditions are 

likely to overestimate operational dose if more scatter and moderation could be involved. It is not 

recommended that credit be taken for potential overresponse of albedo dosimeters, so the bias in this 

case consists only of the dose conversion factor correction and a bias value of 2 is appropriate. This 

value is recommended for the Panasonic UD-809 measurements as well because the detection 

process is albedo based.
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 47 of 57

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Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 51 of 57

GLOSSARY

absorbed dose, D

Amount of energy imparted by radiation to unit mass of absorbing material (100 ergs per 

gram), including tissue. The unit used before the use of the International System of metric 

units is the rad; the International System unit is the gray.

accreditation

Recognition that a dosimeter system has passed the performance criteria of the DOE 

Laboratory Accreditation Program standard in specified irradiation categories.

accuracy

If a series of measurements has small systematic errors, they are said to have high accuracy. 

The accuracy is represented by the bias.

albedo dosimeter

A TLD device that measures the thermal, intermediate, and fast neutrons that are scattered 

and moderated by the body from an incident fast neutron flux.

algorithm

A computational procedure.

Atomic Energy Commission

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

Manhattan Engineering District , and a predecessor to the U.S. Department of Energy. 

backscatter

Deflection of radiation by scattering processes through angles greater than 90 degrees with 

respect to the original direction of motion.

beta particle

A charged particle of very small mass emitted spontaneously from the nuclei of certain 

radioactive elements. Most (if not all) direct fission products emit (negative) beta particles. 

The beta particle is physically identical with an electron moving at high velocity.

buildup

Increase in flux or dose due to scattering in the medium.

calibration blank

A dosimeter that has not been exposed to a radiation source. The results from this dosimeter 

establish the dosimetry system base line or zero dose value.

claimant-favorable

Process of estimation based on technical considerations of parameters significant to dose 

such that the estimated dose is not underestimated.

collective dose equivalent

The sum of the dose equivalents of all individuals in an exposed population. Collective dose is 

expressed in units of person-rem (person-sievert).
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 52 of 57

control dosimeter

A dosimeter used to establish the dosimetry system response to radiation dose. The 

dosimeter is exposed to a known amount of radiation.

curie

A special unit of activity. One curie exactly equals 3.7 x 1010 nuclear transitions per second.

deep absorbed dose (Dd) 

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

composition.

deep dose equivalent (Hd) 

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

densitometer

Instrument that has a photocell to determine the degree of darkening of developed 

photographic film.

density reading

See optical density.

DOE Laboratory Accreditation Program (DOELAP)

Accredits DOE site dosimetry programs based on performance testing and onsite reviews 

performed on a 2-year cycle.

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. When D is expressed in gray, H is in sieverts. (1 sievert = 

100 rem.)

dosimeter

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

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

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

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.)

dosimetry system

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

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

as interpretation and documentation of the results.

DuPont 552

A film packet containing two pieces of film: a 502 sensitive film and a 510 insensitive film.

DuPont 558

A film packet containing a 508 film with one side having a sensitive emulsion and the other 

side insensitive emulsion.

error

A term used to express the difference between the estimated and actual value. 
Effective Date: 09/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0008-6 Page 53 of 57

exchange period (frequency)

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

exposure

As used in the technical sense, a measure expressed in roentgens of the ionization produced 

by gamma (or X-) rays in air. 

fast neutron

Neutron of energy between 10 kilovolts-electron and 10 megavolts-electron. 

film

Generally means a film packet that contains one or more pieces of film in a light-tight 

wrapping. When developed the film has an image caused by radiation that can be measured 

using an optical densitometer. (See DuPont 552, DuPont 558, nuclear track emulsion)

film density

See optical density.

film dosimeter

A small packet of film in a holder that attaches to a worker.

filter

Material used to adjust radiation response of a dosimeter to provide an improved tissue 

equivalent or dose response.

first collision dose

A dose measurement that can be determined for photons or neutrons. For neutron radiation, 

the simplest calculation is one relating dose to flux through a thin layer of tissue. The resultant

graph, sometimes referred to as the first-collision curve, is derived from the assumption that 

the probability of two or more interactions per neutron is negligible. Because of the short 

range of the charged secondary radiation from fast neutrons, the first collision dose in 

irradiated material is practically the same as the absorbed dose.

gamma rays

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

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are 

physically identical to X-rays of high energy; the only essential difference is that X-rays do not 

originate in the nucleus. 

gamma ray interactions

Interaction of gamma rays with matter occurs through three primary processes as follows:

• Photoelectric absorption

The process whereby a gamma ray (or X-ray) photon, with energy somewhat greater than 

that of the binding energy of an electron in an atom, transfers all its energy to the electron, 

which is consequently removed from the atom.

• Compton scattering 

An attenuation process observed for X-ray or gamma radiation in which an incident photon 

interacts with an orbital electron of an atom to produce a recoil electron and a scattered 

photon of energy less that the incident photon.
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• Pair production

An absorption process for X-ray and gamma radiation in which the incident photon is 

annihilated in the vicinity of the nucleus of the absorbing atom, with subsequent production 

of an electron and positron pair. This reaction occurs only for incident photon energies that 

exceed 1.02 megavolts-electron.

gray

The International System unit of absorbed dose (1 gray = 100 rad).

intermediate energy neutron

Neutron of energy between 0.5 electron volt (assumed to be 0.4 electron volt because of 

cadmium cutoff in neutron response) and 10 kilovolts-electron. 

ionizing radiation

Electromagnetic radiation (consisting of photons) or particulate radiation (consisting of 

electrons, neutrons, protons, etc.) capable of producing charged particles through interactions 

with matter.

isotopes

Forms of the same element having identical chemical properties but different atomic masses. 

Isotopes of a given element all have the same number or protons in the nucleus but different 

numbers of neutrons. Some isotopes of an element may be radioactive.

kilovolt-electron (keV)

An amount of energy equal to 1,000 electron volts.

nuclear track emulsion, type A (NTA)

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

that can be seen by using oil immersion and 1000-power microscope.

luminescence

The emission of light from a material as a result of some excitation.

Manhattan Engineering District

Agency designated to develop nuclear weapons; a predecessor to the U.S. Department of 

Energy.

minimum detection level (MDL) 

Often confused because statistical parameters necessary to its calculation are not explicitly 

defined. Nonetheless, the MDL is often assumed to be the level at which a dose is detected at 

the 2-sigma level (i.e., 95% of the time). The MDL should not be confused with the minimum 

recordable dose. 

megavolt-electron (MeV)

An amount of energy equal to 1 million electron volts.

multiple-collision neutron dose

Dose to flux through tissue based on the assumption that two or more interactions per neutron 

occur resulting in greater energy deposition. 
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nuclear emulsion

Generally refers to nuclear track emulsion, type A film.

neutron

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

neutron, fast

Neutrons with energy equal or greater than 10 kilovolts-electron. 

neutron, intermediate

Neutrons with energy between 0.4 electron volt and 10 kilovolts-electron. 

neutron, thermal

Strictly, neutrons in thermal equilibrium with surroundings. Generally, neutrons with energy 

less than the cadmium cutoff at about 0.4 electron volt. 

neutron film dosimeter

A film dosimeter that contains an nuclear track emulsion, type A film packet.

nonpenetrating dose (NP or NPen)

Designation on film dosimeter reports that implies a radiation dose, typically to the skin of 

whole body, from beta and lower energy photon radiation.

open window (OW)

Designation on Hanford film dosimeter reports that implies the use of little shielding (only that 

of the security credential). It commonly is used to label the film response corresponding to the 

OW area. 

optical density

The quantitative measurement of the density of photographic blackening defined as 

D = Log10(Io/I).

pencil dosimeters

A type of ionization chamber used by personnel to measure radiation dose. These results 

may be labeled as Pen dose. Other names: pencil, pocket dosimeter, pocket pencil, pocket 

ionization chamber, PIC.

penetrating dose (P or Pen)

Designation on film dosimeter reports that implies a radiation dose, typically to the whole body, 

from higher energy photon radiation.

personal dose equivalent, Hp(d)

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

radiological protection purposes by the International Commission on Radiological Units and 

Measurements. Represented by Hp(d), where d identifies the depth (in millimeters) and 

represents the point of reference for dose in tissue. For weakly penetrating radiation of 

significance to skin dose, d = 0.07 millimeter and is noted as Hp(0.07). For penetrating 

radiation of significance to whole-body dose, d = 10 millimeters and is noted as Hp(10). 

photon

A unit or particle of electromagnetic radiation consisting of X- and/or gamma rays. 
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quality factor (Q) 

A modifying factor used to derive dose equivalent from absorbed dose.

rad

A unit of absorbed dose equal to the absorption of 100 ergs per gram of absorbing material, 

such as body tissue.

radiation

One or more of beta, neutron, and photon radiation. 

radiation monitoring

Routine measurements and the estimation of the dose equivalent for the purpose of 

determining and controlling the dose received by workers.

radioactivity

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

neutrons from unstable nuclei

random errors

When a given measurement is repeated the resultant values, in general, do not agree exactly. 

The causes of the disagreement between the individual values must also be causes of their 

differing from the actual value. Errors resulting from these causes are called random errors.

relative biological effectiveness (RBE) 

A ratio of the absorbed dose of a reference radiation to the absorbed dose of a test radiation 

producing the same biological effects, other conditions being equal.

rem

The rem is a unit of dose equivalent, which is equal to the product of the number of rads 

absorbed and the quality factor. The word derives from roentgen equivalent in man.

rep

Historically the rep has been used extensively for the specification of permissible doses of 

ionizing radiations other than X-rays or gamma rays. Several definitions have appeared in the 

literature but in the sense most widely adopted, it is a unit of absorbed dose with a magnitude 

of 93 ergs per gram. The word derives from roentgen-equivalent-physical.

roentgen

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

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

air. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in 

soft tissue.

scattering

The diversion of radiation from its original path as a result of interactions with atoms between 

the source of the radiations and a point at some distance away. Scattered radiations are 

typically changed in direction and of lower energy than the original radiation.

shallow absorbed dose (Ds)

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

composition.
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shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 millimeter in tissue.

shielding

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

personnel or materials from radiation.

sievert

The International system unit for dose equivalent. (1 sievert = 100 rem.)

skin dose

Absorbed dose at a tissue depth of 7 milligrams per square centimeter. 

thermal neutron

Strictly, neutrons in thermal equilibrium with surroundings. Generally, refers to neutrons of 

energy less-than the cadmium cutoff of about 0.4 electron volt. 

tissue equivalent

Used to imply that radiation response characteristics of the material being irradiated are 

equivalent to tissue. Achieving a tissue-equivalent response is an important consideration in 

the design and fabrication of radiation measuring instruments and dosimeters.

thermoluminescent

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

thermoluminescent dosimeter (TLD)

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

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

chips are sometimes called crystals.

thermoluminescent dosimeter chip

A small block or crystal made of LiF used in the TLD.

TLD-600 - A TLD chip made from 6

Li (>95%) used to detect neutrons.

TLD-700 - A TLD chip made from 7

Li (>99.9%) used to detect photon and beta radiation.

whole-body dose

Absorbed dose at a tissue depth of 1.0 centimeter (1,000 milligrams per square centimeter); 

however, also used to refer to the dose recorded.

X-ray

Ionizing electromagnetic radiation of extranuclear origin.