ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 79

DOE Review Release 03/29/2013

Document Title:

Los Alamos National Laboratory –

Occupational External Dose

Document Number: ORAUT-TKBS-0010-6

Revision: 03

Effective Date: 03/21/2013

Type of Document: TBD

Supersedes: Revision 02

Subject Expert(s): Donald N. Stewart

Approval: Signature on File Approval Date: 03/13/2013

Donald N. Stewart, Document Owner

Concurrence: Signature on File Concurrence Date: 03/12/2013

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 03/13/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 02/12/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/21/2013

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 2 of 79

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

05/10/2005 00 New Technical Basis Document for the Los Alamos National 

Laboratory – Occupational External Dose. First approved issue. No 

training required. Initiated by Jack E. Buddenbaum.

05/30/2007 01 Approved revision to update document. Adds Purpose and Scope 

sections. Constitutes a total rewrite of document. Adds Attributions 

and Annotations section. Incorporates formal internal and NIOSH 

review comments. This revision results in an increase in assigned 

dose and a PER is required due to the addition of median photon 

adjustment values in Tables 6-22 and A-8; otherwise, the maximum 

values were replaced with lower 95th percentiles, and therefore 

would lower doses. An exception to the lower 95th-percentile values 

is the value for Other Operations, which is higher than the previously 

reported maximum value of annual geometric means. This change 

will result in a higher dose. Changes in Table A-7 will increase 

skin/shallow dose for LAMPF/LANSCE workers. Training required: 

As determined by the Task Manager. Initiated by Jack E. 

Buddenbaum.

11/23/2009 02 Approved revision was initiated to make the following changes: 

updated NIOSH references in Sections 6-1, 6-2, A-6 and Reference 

Section. Added details to the Scope in Section 6.1. In Section 6.2,

“Reporting Practices” was modified to distinguish between LANL 

records with non-penetrating dose listed as “SHALLOW” vs. those 

listed as “SKIN.” Several changes were made to remove dose 

reconstruction methods and information relating to periods when it 

has been determined to be infeasible to estimate external doses. 

Specifically, according to the HHS letter adding a class of LANL 

employees to the SEC (1943-1975), partial dose reconstructions may 

include external doses from gamma emitters in the years 1946-1975, 

neutrons from 1946-1975, and beta-emitters in the years 1949-1975. 

Incorporates formal internal and NIOSH review comments. No 

sections were deleted. Training required: As determined by the 

Objective Manager. Initiated by Donald N. Stewart.

03/21/2013 03 Revised to incorporate dose reconstruction limitations that relate to a 

class of LANL employees added to the SEC comprised of all workers 

from January 1, 1976, to December 31, 1995, SEC-00109 (NIOSH 

2012). Attachment B has been added, which contains an analysis 

that evaluates the neutron-to-photon ratios promulgated in earlier 

revisions of the TBD. No changes were made to the neutron-tophoton ratios as a result. Incorporates formal internal and NIOSH 

review comments. For the years 1976-1980, NIOSH has agreed, in 

addressing comments on the TBD, to assign neutron dose using the 

95th-percentile value of the NP ratio to address uncertainties 

associated with the transition from film to TLD dosimeters in 

response to the LANL ABRWH Work Group. Constitutes a total 

rewrite of the document. Training required: As determined by the 

Objective Manager. Initiated by Donald N. Stewart.
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TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

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

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

6.1.3 Special Exposure Cohort............................................................................................ 11

6.2 Dose Reconstruction Parameters........................................................................................... 12

6.2.1 Administrative Practices ............................................................................................. 13

6.2.1.1 Assignment of Dosimeters to Workers ......................................................... 13

6.2.1.2 Dosimeter Exchange Frequencies ............................................................... 15

6.2.1.3 Use of Control Dosimeters ........................................................................... 17

6.2.1.4 Reporting Conventions................................................................................. 19

6.2.1.5 Recordkeeping............................................................................................. 22

6.2.1.6 Quality of External Dosimetry Data .............................................................. 23

6.2.2 Dosimetry Technology................................................................................................ 24

6.2.2.1 Beta/Photon Dosimeters .............................................................................. 25

6.2.2.2 Neutron Dosimeters ..................................................................................... 27

6.2.3 Calibration .................................................................................................................. 30

6.2.3.1 Beta/Photon Dosimeters .............................................................................. 31

6.2.3.2 Neutron Dosimeters ..................................................................................... 34

6.2.3.3 Workplace Beta/Photon Dosimeter Response.............................................. 36

6.2.3.4 Workplace Radiation Fields.......................................................................... 39

6.2.3.5 Adjustments to Measured Dose Shallow Dose Components........................ 40

6.2.3.6 Workplace Neutron Dosimeter Response .................................................... 40

6.2.3.6.1 Plutonium Processing Areas (TA-1, TA-21, TA-55)..................... 41

6.2.3.6.1.1 Neutron Energy Spectrum ...................................... 41

6.2.3.6.1.2 Neutron-to-Photon Dose Ratio................................ 41

6.2.3.6.2 LAMPF (TA-53)........................................................................... 43

6.2.3.6.3 Critical Assembly Testing (TA-18)............................................... 44

6.2.3.6.3.1 Neutron Energy Spectrum ...................................... 44

6.2.3.6.3.2 Gamma-to-Neutron Dose Ratio .............................. 44

6.2.3.6.4 Reactor Areas (TA-2).................................................................. 45

6.2.3.6.5 CMR Building (TA-3)................................................................... 46

6.2.3.7 Neutron Dose Fraction ................................................................................. 46

6.2.3.8 Uncertainty in Neutron Dose ........................................................................ 47

6.3 Adjustments to Recorded Photon Dose.................................................................................. 47

6.4 Adjustments to Recorded Neutron Dose ................................................................................ 48

6.4.1 Neutron Dose Adjustments......................................................................................... 48

6.4.2 Neutron Weighting Factor........................................................................................... 48

6.4.3 Neutron Correction Factor .......................................................................................... 49

6.4.4 Neutron-to-Gamma Dose Factors............................................................................... 50

6.5 Missed Dose .......................................................................................................................... 51

6.5.1 Photon Missed Dose .................................................................................................. 51

6.5.2 Neutron Missed Dose ................................................................................................. 52
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6.6 Organ Dose............................................................................................................................ 52

6.7 Attributions and Annotations .................................................................................................. 53

References ........................................................................................................................................ 54

Glossary ............................................................................................................................................ 60

ATTACHMENT A, GUIDANCE FOR DOSE RECONSTRUCTORS: OCCUPATIONAL

EXTERNAL DOSE FOR MONITORED WORKERS............................................. 71

ATTACHMENT B, LANL NEUTRON-TO-PHOTON (NP) DOSE RATIO ANALYSIS.......................... 76

LIST OF TABLES

TABLE TITLE PAGE

6-1 Volumes of Photodosimetry Evaluation Book......................................................................... 10

6-2 Annual external radiation doses, 1944 to 2008....................................................................... 15

6-3 Typical exchange frequencies for dosimeters in use during different periods ......................... 17

6-4 Recorded dose practices over time (LASL 1959, 1969, 1977, 1979; LANL 1986, 1989, 

1996, 2001, 2003).................................................................................................................. 21

6-5 External dosimetry events...................................................................................................... 24

6-6 Photon and beta dosimeters used over time .......................................................................... 28

6-7 Neutron dosimeters used over time at LANL.......................................................................... 30

6-8 TLD and film badge energy response to beta radiation, about 1977....................................... 33

6-9 DOELAP irradiation techniques and effective energies .......................................................... 34

6-10 General energy distribution of photons in 1962 at DP West Site ............................................ 37

6-11 Estimated photon spectra for several plutonium sources in DP West Site.............................. 38

6-12 Calculated errors from use of the Cycolac badge with low-energy photons and a 

correction factor of 0.07 ......................................................................................................... 38

6-13 Results of a 6-month comparison of film badges, NTA film, and TLDs at DP Site .................. 38

6-14 Selection of beta and photon radiation energies and percentages ......................................... 39

6-15 Approximate NCFs and dose fractions for neutron sources.................................................... 41

6-16 Results of 6-month comparison of film badges, NTA film, and TLDs at DP Site ..................... 43

6-17 Percent of total kerma by energy interval from critical assemblies during experiments........... 45

6-18 Gamma-to-neutron ratios measured with TLDs on the front of plastic man manikins at 

various distances from the Hydro critical assembly ................................................................ 45

6-19 Laboratory-measured dose fractions from PuF4 ..................................................................... 46

6-20 Neutron quality or weighting factors ....................................................................................... 48

6-21 Facility dose fractions and associated ICRP Publication 60 correction factors ....................... 49

6-22 Recommended distributions for neutron-to-gamma ratio........................................................ 51

6-23 Photon dosimeter period of use, type, MDL, exchange frequency, and potential annual 

missed dose...........................................................................................................................51

6-24 Neutron dosimeter period of use, type, MDL, exchange frequency, and potential annual 

missed dose...........................................................................................................................52

A-1 Worker gamma dose statistics ............................................................................................... 72

A-2 Recommended adjustments to reported photon doses .......................................................... 73

A-4 Recommended uncertainty factors for reported doses ........................................................... 74

A-4 Beta/photon dosimeter period of use, type, MDL, exchange frequency, and potential 

annual missed doses .............................................................................................................75
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A-5 Recommended DCFs for dose assessments ......................................................................... 75

B-1 Statistical parameters for selected periods of LANL recorded doses...................................... 78

B-2 Statistical analysis of LANL recorded doses........................................................................... 78

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Number of workers monitored as a function of time................................................................ 17

6-2 Comparison of Hp(10) for neutrons with energy responses of NTA film and neutron 

albedo dosimeter containing a thermoluminescent neutron dosimeter chip of 6

LiF and 

shielded by cadmium ............................................................................................................. 31

6-3 Geometric mean of neutron-to-photon ratios for records with both deep dose and 

neutron dose of 50 mrem or greater, 1979 to 2008 ................................................................ 37

6-4 Energy dependence of Eastman Kodak Type K film for exposure of 0.1 R............................. 37

6-5 Neutron spectrum of hydrofluorination of 239PuO2 at Plutonium Facility .................................. 42

6-6 Neutron spectrum of 238Pu ball milling process at Plutonium Facility ...................................... 42

6-7 Neutron spectrum near Door K in SNM Vault at Plutonium Facility ........................................ 42

6-8 Neutron spectrum in ER-1 Area of LAMPF with proton beam stopped in the carbon 

beam block, as determined from unfolding codes .................................................................. 43

6-9 Neutron spectrum at Station 6 near TA-18 ............................................................................. 44

B-1 LANL worker GM and 95th-percentile NP dose ratio, 1950 to 2008 ....................................... 77
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ACRONYMS AND ABBREVIATIONS

ABS acrylonitrile butadiene styrene

AEC U.S. Atomic Energy Commission

CFR Code of Federal Regulations

CHP Certified Health Physicist

Ci curie

CIH Certified Industrial Hygienist

cm centimeter

CMR Chemistry and Metallurgy Research (Building or Division)

d day

DCF dose conversion factor

DE dose equivalent

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ERDA U.S. Energy Research and Development Administration

eV electron-volt

EVM Evaluation Method

g gram

GM geometric mean

GSD geometric standard deviation

H Division Health Division

HT Heat Treatment (building)

HYPO Water Boiler Reactor in high-power configuration

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kVp peak kilovoltage

LAMPF Los Alamos Meson Physics Facility

LAMPRE Los Alamos Molten Plutonium Reactor Experiment

LANL Los Alamos National Laboratory

LAPRE Los Alamos Power Reactor Experiment

LANSCE Los Alamos Neutron Science Center

LOPO Water Boiler Reactor in low-power configuration

m meter

MDL minimum detection level 

MED Manhattan Engineer District

MeV megaelectron-volt, 1 million electron-volts

mg milligram

min minute

mm millimeter

mrad millirad
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mrem millirem

mSv millisievert

NADE neutron ambient dose equivalent

NBS National Bureau of Standards

NCF neutron correction factor

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NIST National Institute of Standards and Technology 

NP neutron-to-photon

NTA nuclear track emulsion, type A

NTP nuclear track plate

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

PNL Pacific Northwest Laboratory

PNNL Pacific Northwest National Laboratory

POC probability of causation

PTB Physikalisch-Technische Bundesanstalt

PVC polyvinylchloride

R or r roentgen, an early unit of radiation exposure

SEC Special Exposure Cohort

SNM Special Nuclear Material

SUPO Water Boiler Reactor in highest (Super) power configuration

TA Technical Area

TBD technical basis document

TED track-etch dosimetry or dosimeter

TLD thermoluminescent dosimeter

UHTREX Ultra High-Temperature Reactor Experiment

U.S.C. United States Code

WB whole-body

wk week

yr year

μC microcurie

μm micrometer

§ section or sections
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6.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] guidelines established under 

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

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

referred to in 42 U.S.C. § 7384n(b) to nuclear weapons work (NIOSH 2010).

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
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6.1.1 Purpose

When the Los Alamos site became operational in 1943, it had a single mission – the design and 

manufacture of the first nuclear weapons (Hoddeson et al. 1993). In 1947, Los Alamos Laboratory 

(Project Y) became Los Alamos Scientific Laboratory, which in 1981 became Los Alamos National 

Laboratory (LANL; University of California 2001); for simplicity, except in reference citations, this 

technical basis document (TBD) uses LANL for all periods. LANL assignments in the early 1940s 

included:

• Performing the final purification of plutonium received 

• Reducing plutonium to its metallic state

• Determining the relevant physical and metallurgical properties of plutonium

• Developing weapon component fabrication technologies (Hammel 1998)

Processes included nuclear fuel fabrication; nuclear criticality experimentation; nuclear reactor 

operations; radiochemical separations; refining, finishing, and storing plutonium and various 

enrichments of uranium; processing large quantities of other nuclear materials such as tritium, 

polonium, and lanthanum; testing nuclear device components and the devices themselves; and 

handling the associated radioactive waste (Widner et al. 2004). After World War II, LANL scientists 

and engineers were involved in development and testing of nuclear devices that were more and more 

powerful, compact, reliable, and dependably deployable in the field, and that were contained in a 

variety of delivery vehicles suited to various combat objectives. 

LANL was the lead site for U.S. nuclear component fabrication until 1949, when the Hanford 

Plutonium Finishing Plant in Washington began making pits, the central cores of the primary stages of 

nuclear devices (DOE 1997). Plutonium processing at LANL took place in D Building in 1944 and 

1945, then moved to DP Site, which housed a variety of plutonium research, design, and fabrication 

activities for more than 30 years until Technical Area (TA)-55 became operational in 1978 (Widner et 

al. 2004). After 1949, LANL was a backup production facility that designed, developed, and fabricated 

nuclear components for test devices. From time to time, LANL performed special functions in its 

backup role. For example, due to a 1984 accident at the Hanford Site, plutonium oxide was sent to 

LANL for conversion to metal (DOE 1997). 

Operations, facilities, and capabilities that were needed to support development and production of the 

various types of nuclear devices expanded in many cases to support other missions after World War II 

(Widner et al. 2004). Programs in chemistry, metallurgy, and low-temperature physics expanded into 

nonmilitary development and fundamental research. The Health Division (H Division) grew 

significantly and expanded into many areas of health physics, industrial hygiene, medicine, safety, 

and biomedical research related to people and radiation. Early reactors built to confirm critical 

masses for fissionable materials and to study properties of fission and the behavior of the resulting 

neutrons were the forerunners of a variety of reactors designed and in some cases built and operated 

at LANL. While some of these reactors served as sources of neutrons for nuclear research or 

materials testing, other designs were pursued for potential applications in power generation and 

propulsion of nuclear rockets into deep space. During World War II, accelerators were used to 

determine the critical masses for each proposed nuclear weapon design. After the war, other 

accelerators were used, including one of the world’s largest at the Los Alamos Meson Physics Facility 

(LAMPF). Some of the first significant steps toward controlled nuclear fusion as a power source 

occurred at LANL, and the plasma thermocouple program explored methods for direct conversion of 

fission energy to electricity for potential application in propulsion of spacecraft.

In more recent decades, LANL has been a large multidiscipline research institution that utilizes a wide 

variety of radioisotopic sources, radiation-producing machines, and critical assemblies (Hoffman and 

Mallett 1999a). Employees could have received occupational radiation exposure from beta, photon, 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 10 of 79

and neutron radiation. Most occupational external radiation exposure at LANL (at least during the 

1980s and 1990s) has been due to neutron radiation, which has accounted for about 60% of the 

external collective dose equivalent (DE) (Hoffman and Mallett 1999a). Neutron radiation exposures at 

LANL originate from radioisotope sources, nuclear materials handling, critical assemblies, and 

accelerators. Although they are a smaller part of the total, beta and photon radiation exposures occur 

from a larger variety of source types. In addition to those for neutron radiation, these sources include 

radiation-producing machines, medical isotopes for research and production, and others. 

LANL has used facility and individual worker monitoring methods to measure and control radiation 

exposures. Records of radiation doses to individual workers from personnel dosimeters worn by the 

worker and coworkers are available for LANL operations beginning in 1943. Doses from these 

dosimeters were recorded at the time of measurement and routinely reviewed by operations and 

radiation safety personnel for compliance with radiation control limits. The NIOSH External Dose 

Reconstruction Implementation Guidelines (NIOSH 2007a) has identified these as the highest quality 

records for retrospective dose assessments. The information in this TBD pertains to analysis of these 

records. 

Radiation dosimetry practices were initially based on experience gained during several decades of 

radium and X-ray medical diagnostic and therapy applications. These methods were generally well 

advanced at the start of the Manhattan Engineer District (MED) program to develop nuclear weapons 

on August 13, 1942. The primary challenges encountered by MED, and later U.S. Atomic Energy 

Commission (AEC, the MED successor agency), operations to measure worker dose to external 

radiation involved:

• Comparatively large quantities of high-level radioactivity

• Mixed radiation fields involving beta, photon (gamma and X-ray), and neutron radiation with 

low, intermediate, and high energies

• Neutron radiation

Many details about policies, procedures, practices, and issues dealing with external radiation 

monitoring at LANL are documented in a compilation of correspondence and reports called the 

Photodosimetry Evaluation Book. Informally, this compilation, which has nine volumes, has been 

referred to as “the Bible” of external radiation dosimetry at LANL. It was first assembled in 1956, and 

was expanded until around 2003 (Widner 2004). Table 6-1 lists the volumes that currently compose 

this resource.

Table 6-1. Volumes of Photodosimetry Evaluation Book.

Volume Approximate date range Reference

I 01/19/44–07/28/58 LASL 1959

II 01/02/60–12/23/69 LASL 1969

III 02/18/70–07/26/77 LASL 1977

IV 01/27/78–10/21/86 LANL 1986

V 06/05/45–02/08/79 Misc. docs. LASL 1979

VI 04/08/85–12/15/89 LANL 1989

VII 02/02/90–02/15/96 LANL 1996

VIII 02/08/96–12/05/01 LANL 2001

IX (active) 02/14/02–05/28/03 LANL 2003

6.1.2 Scope

Section 6.2 discusses the program for measuring skin and whole-body (WB) doses to the workers. 

The methods for evaluating external doses to workers have evolved over the years as new techniques 
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and equipment have been developed. Concepts in radiation protection have also changed. The dose 

reconstruction parameters, LANL practices and policies, and dosimeter types and technology for 

measuring the dose from different types of radiation are discussed. Attention is given to the valuation 

of doses measured from exposure to beta, gamma, and neutron radiation. Test results are tabulated 

for various dosimeters exposed at different geometries and radiation energies.

For photon and neutron dose, Sections 6.3 and 6.4, respectively, discuss the sources of bias, 

workplace radiation field characteristics, responses of the different beta/gamma and neutron 

dosimeters in the workplace fields, and the adjustments to the recorded dose measured by these 

dosimeters during specific years.

Section 6.5 presents sources of potential dose that could be missed because of the limitations of 

dosimetry systems and the methods of reporting low doses. This missed dose is discussed as a 

function of facility location, dosimeter type, year, and energy range. Section 6.6 and Attachment A 

describe the use of the external dosimetry technical basis parameters to facilitate the efforts of the 

dose reconstructors. Attachment B describes analyses performed to calculate the statistical 

parameters of lognormal distributions of neutron-to-photon (NP) dose ratios, and provides an 

annotated chronology of radiation-dosimetry related events at LANL.

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

justification, or clarification of the associated information, are presented in Section 6.7.

6.1.3 Special Exposure Cohort

The Secretary of the U.S. Department of Health and Human Services has designated three classes of 

LANL employees as additions to the SEC based on the findings and recommendations of NIOSH and 

the Advisory Board on Radiation and Worker Health Board:

• Employees of DOE predecessor agencies and their contractors or subcontractors who were 

monitored or should have been monitored for exposure to ionizing radiation during radioactive 

lanthanum (RaLa) operations at Technical Area (TA)-10 (Bayo Canyon Site), TA-35 (Ten Site), 

and Buildings H, Sigma, and U (in TA-1) for a number of workdays aggregating at least 250 

workdays during the period from September 1, 1944, through July 18, 1963, or in combination 

with workdays within the parameters that have been established for one or more other classes 

of employees in the SEC (Leavitt 2006; NIOSH 2006). 

Through the course of ongoing dose reconstruction and research, NIOSH determined that, due to 

undocumented worker movements across the site and limited claimant-specific information pertaining 

to work locations, it was unable to eliminate any specific worker from potential exposure scenarios 

based on assigned work location. NIOSH has found that a determination cannot always be made as 

to whether or not an employee worked in Technical Areas with a history of radioactive material use, or 

whether an employee should have been monitored for radiological exposures. Accordingly, NIOSH 

determined that it was necessary to remove the area-specific and monitoring criteria from the class 

description above, and to expand the SEC class definition to include all areas of LANL, and all 

employees of the DOE, its predecessor agencies, and their contractors and subcontractors who 

worked at LANL during the specified time period, regardless of monitoring. For this reason, the class 

of employees listed below was added. 

• All employees of the Department of Energy, its predecessor agencies, and their contractors 

and subcontractors who worked at the Los Alamos National Laboratory in Los Alamos, New 

Mexico from March 15, 1943 through December 31, 1975, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in 
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combination with work days within the parameters established for one or more other classes of 

employees in the Special Exposure Cohort (Sebelius 2010).

• Employees of DOE, its predecessor agencies, and DOE contractors or subcontractors who 

were monitored or should have been monitored for radiological exposures while working in 

operational TAs with a history of radioactive material use at LANL for an aggregate of at least 

250 workdays during the period from March 15, 1943, through December 31, 1975, or in 

combination with workdays within the parameters that have been established for one or more 

other classes of employees in the SEC (Leavitt 2007; NIOSH 2007b).

• Employees of the Department of Energy, its predecessor agencies, and their contractors and 

subcontractors who worked at LANL from January 1, 1976, through December 31, 1995, for a 

number of workdays aggregating at least 250 workdays occurring either solely under this 

employment or in combination with workdays within the parameters established for one or 

more other classes of employees included in the SEC (Sebelius 2012; NIOSH 2012).

NIOSH has determined that it does not have access to sufficient information – including internal 

personnel dosimetry, workplace monitoring data, or sufficient process and radiological source 

information – that would allow it to estimate with sufficient accuracy the potential internal doses 

workers might have received. NIOSH has determined that there is insufficient information either to 

estimate the maximum radiation dose for every type of cancer for which radiation doses are 

reconstructed that could have been incurred under plausible circumstances by any member of the 

classes or to estimate the radiation doses of members of the classes more precisely than a maximum 

dose estimate. 

From an occupational external dose reconstruction perspective, partial dose reconstructions can

include, and are limited to, external dose from gamma emitters from 1946 onward; dose from beta 

emitters from 1949 onward, and neutron dose from 1946 onward.

6.2 DOSE RECONSTRUCTION PARAMETERS

Examinations of the beta, photon (X-ray, gamma ray), and neutron radiation type, energy, and 

geometry of exposure in the workplace, and the characteristics of the respective LANL dosimeter 

responses are crucial to the assessment of bias and uncertainty of the original recorded dose in 

relation to the radiation quantity Hp(10). The bias and uncertainty for current LANL dosimetry 

systems are well documented for Hp(10) (Hoffman and Mallett 1999a,b; LANL 1989, 1996, 2001, 

2003). The performance of current dosimeters can often be compared with performance 

characteristics of historical dosimetry systems in the same or highly similar facilities or workplaces. In 

addition, current performance testing techniques can be applied to earlier dosimetry systems to 

achieve a consistent evaluation of historical dosimetry systems. Dosimeter response characteristics 

for radiation types and energies in the workplace are crucial to the overall analysis of error in recorded 

dose. Accuracy and precision of recorded worker doses and the comparability of the recorded dose 

types depend on the following considerations (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

• Dosimetry technology, which includes physical capabilities of the dosimetry system such as 

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

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

exposure in the workplace
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 13 of 79

• Workplace radiation fields that might 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 Hp(10) and, as needed, Hp(0.07) for individual workers with the least overall uncertainty. 

6.2.1 Administrative Practices

Historically, LANL had an extensive radiation safety monitoring program using portable radiation 

instruments, contamination surveys, zone controls, and personnel dosimeters to measure dose in the 

workplace (LASL 1959, 1969, 1977, 1979; LANL 1986, 1989, 1996, 2001, 2003). This program was 

conducted directly by or under the guidance of a specially trained group of radiation monitors or 

radiation protection technologists. Results from the dosimeters were used to evaluate and record 

doses from external radiation exposure to workers throughout the history of LANL operations. 

Dosimeters that have been used fall into the following categories:

• Personnel WB beta/photon 

• Pocket ionization chamber (PIC) 

• Personnel extremity

• Personnel WB neutron

Shortly after operations began in 1943, some workers were monitored with PICs alone. In 1943, the 

phase-in of photon film dosimetry methods began at LANL. More and more groups started using film 

badges for photon dosimetry, and in 1949 a new badge was introduced to support evaluation of beta 

exposures. Beta/gamma film badge designs changed several times through the 1950s, 1960s, and 

1970s as filters of various types were used to address the energy-dependent response of film. LANL 

officially switched to the use of thermoluminescent dosimeters (TLDs) in 1980 and is currently using 

its second generation of TLD badge. 

Before 1949, the Laboratory implemented neutron dosimetry for selected workers beginning with the 

use of PICs that incorporated Bakelite chambers and graphite coatings. In 1949, nuclear track plates 

(NTPs) were first used, and badges incorporating Eastman Kodak nuclear track emulsion, type A 

(NTA) film were first used in 1951. While TLD badges were used for neutron dosimetry beginning in 

1980, NTA film continued to be used in a “piggyback” fashion with TLD badges for some workers. 

Track-etch dosimeters (TEDs) have been used for evaluation of fast neutron doses since 1995. 

LANL administrative practices significant to dose reconstruction include policies to:

• Assign dosimeters to workers

• Exchange dosimeters

• Use control dosimeters

• Estimate dose for missing or damaged dosimeters

• Replace destroyed or missing records

• Evaluate and record dose for incidents

• Obtain and record occupational dose to workers for other employer exposure

6.2.1.1 Assignment of Dosimeters to Workers

When monitoring for external radiation exposures started in 1943, PICs were assigned to “a few 

persons thought to have the highest potential for receiving exposures at or above the ‘tolerance’ limit” 

[LANL 1986 (10/6/81 questionnaire)]. At that time, the tolerance limit was 0.1 R/d based on a National 

Bureau of Standards (NBS) Recommendation [LASL 1959 (3/21/44 industrial hygiene survey and 

2/26/48 LASL memorandum)]. By 1945, when film badges were in use by a number of LANL groups, 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 14 of 79

only workers with the “higher exposure potentials” were issued dosimeter badges [LANL 1986 

(10/6/81 questionnaire)]. At the time of the earliest criticality experiments and accidents at LANL, 

workers who received the highest exposures had not yet been issued film badges. Their exposures 

were calculated from activation measurements and area film badges [LANL 1986 (10/6/81 

questionnaire)]. 

In some cases, considerable evaluation went into the determination of whether groups of workers 

should wear dosimeters. In June 1948, it was reported that workers in DP East were not wearing film 

badges [LASL 1959 (June 7 and 8 memoranda)]. After a trial period of 3 months showed no 

exposure, workers stopped wearing film badges in the spring of 1947. An assessment concluded that 

the gamma-ray intensity from polonium sources at DP East was very low in comparison with their 

alpha activity, and that film badges would be needed only for those who worked in and around the 

storage vault if it were to contain about 2,000 Ci or more [LASL 1959 (June 8 memorandum)]. For 

“urchin”-type initiators [1], gamma exposure to the fingers determined allowable exposure times. 

Wrist badges would record perhaps 1% of the dose to the fingertips, and total-body dose would be 

negligible. Because of this, tolerance times were specified as 100 Ci-min at 1 cm, and it was 

recommended that studies involving use of impressions of the ridges on the fingertips as indicators of 

radiation exposure be continued [LASL 1959 (June 7 and 8 memoranda)]. For preparation of 

polonium-beryllium (PoBe) neutron sources, fast neutron exposure was limiting. Around October 

1948, the need for film monitoring at the DP West plutonium facilities was recognized because of lowlevel gamma exposure from spontaneous fission in plutonium and the possibility of a criticality 

accident. It was emphasized at the time that AEC safety regulations required that film badges be 

worn in any area where radioactive materials were handled.

As of April 1960, the brass-cadmium film badge was being worn by about half of University of 

California employees, all of the Security Force, and 75% of Zia employees (LASL 1969). At most 

sites, film badges were reportedly issued "only to personnel who need them" (LASL 1969). At that 

time, the plan was to combine the film badge with the site security badge to increase compliance with 

the requirement to wear dosimeter badges, and it was proposed that badges be issued to all LANL 

and AEC personnel on a regular basis.

In October 1962, the new multielement Cycolac film badge was issued to about 100 persons who had 

“histories of appreciable or out of ordinary radiation exposures” [LASL 1969 (3/5/63 memorandum)]. 

The Cycolac badge was used for all film dosimetry from about 1963 to about 1977, when the use of 

some TLD badges began (Widner 2003). Between 1943 and late 1981, the number of persons 

monitored for external exposures increased from less than 100/yr to more than 5,000/yr, but 

personnel monitoring for external radiation still had not been extended to all workers at LANL. 

During the time film badges were in use, about 15,000 gamma and neutron films were developed 

each month. By June 1981, fewer than 40 NTA films were developed each month. By December 

1990, approximately 7,800 TLD badges were processed each month, plus about 200 NTA film badges 

per month that were issued to workers at LAMPF for high-energy dosimetry during accelerator 

operations (about 7 mo/yr) (LANL 1996). 

Table 6-2 lists the reported numbers of workers monitored by LANL from 1944 to 2008 along with total 

and average doses calculated from those data (LANL 2004). Estimates of the total number of 

workers employed at LANL each year are also listed [2]. “Average total dose” for each year is 

calculated as the total dose (person-rem) divided by the number of workers monitored (that is, in 

Table 6-2, column 2 divided by column 3). Figure 6-1 shows the number of workers monitored each 

year over the same period (LANL 2004). The data on which Table 6-2 are based have not been 

corrected for potential missed doses.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 15 of 79

6.2.1.2 Dosimeter Exchange Frequencies

Dosimeters were exchanged on routine schedules. In the earliest operations, daily measurements 

with PICs were performed, but results are generally not available in employee records. After film 

badges came into use in 1943, daily PIC measurements continued, but film measurements provided a 

check of the daily measurements and formed a permanent record of worker exposures. Film packets 

were exchanged and processed monthly for most workers, but were exchanged more frequently (as 

often as daily) for certain operations with high exposure potential. In February 1948, it was reported 

that brass film badges (film placed directly in a brass container, with no window of any sort or filters of 

any other metal) had been issued to all Sigma and Heat Treatment (HT) Building personnel in 

October 1947 to monitor beta and gamma exposures and had been exchanged and reissued every 

2 weeks since that time [LASL 1959 (2/26/48 memorandum)]. Health Group personnel were then 

considering the exchange of badges more often than every 2 weeks for personnel who were handling 

“abnormally large amounts” of 235U during certain periods at these facilities [LASL 1959 (2/26/48 

memorandum)].

In October 1956, beta/gamma dosimeters were exchanged at 1- or 2-week intervals, with 2-week 

intervals predominating [LASL 1959 (10/4/56 memorandum)]. NTPs were exchanged at 4-week 

intervals. There was a desire to use 2-week intervals for NTPs and read all plates, but the H-1 

monitoring group was unable to support this due to labor considerations. At that time, only half the 

issued plates were read [LASL 1959 (10/4/56 memorandum)]. In September 1988, LANL evaluated 

changing from monthly to monthly/quarterly badge issue (LANL 1989). The HSE-1 section leader

recommended against changing to monthly/quarterly issuance, stating that problems with fading 

supported continuation of monthly exchange. Monthly exchange continued to be the most common.

Table 6-2. Annual external radiation doses, 1944 to 2008

(LANL 2004, updated 2009).

Year

Total dose 

(person-rem)

No. of workers Average total 

Monitored Total dose (rem)

1944 7.89 9 1,235 (0.88)a

1945 2,153.37 812 1,235 (2.65)

a,b

1946 3,819.95 508 2,750 7.52a,b

1947 464.62 1,237 2,750 0.38

1948 343.05 2,080 1,570 0.16

1949 466.42 3,177 1,940 0.15

1950 552.49 3,895 2,420 0.14

1951 2,503.80 4,257 2,700 0.59

1952 812.72 2,366 2,790 0.34

1953 717.16 1,878 2,900 0.38

1954 920.18 2,068 2,970 0.44

1955 763.25 1,984 3,040 0.38

1956 1,280.80 2,287 3,000 0.56

1957 585.75 2,539 3,150 0.23

1958 5,704.65 3,032 3,223 1.88a,b

1959 447.09 2,930 3,214 0.15

1960 541.02 3,622 3,300 0.15

1961 299.22 3,973 3,450 0.08

1962 386.62 4,119 3,632 0.09

1963 251.83 4,176 3,428 0.06

1964 250.63 4,103 3,740 0.06

1965 364.67 4,222 3,800 0.09

1966 241.24 4,446 3,930 0.05

1967 268.31 4,072 4,050 0.07

1968 285.12 3,861 4,250 0.07
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 16 of 79

Year

Total dose 

(person-rem)

No. of workers Average total 

Monitored Total dose (rem)

1969 388.69 3,980 4,350 0.10

1970 408.92 4,031 4,100 0.10

1971 341.21 3,775 4,100 0.09

1972 338.66 3,877 4,300 0.09

1973 436.00 3,866 4,450 0.11

1974 409.64 4,337 4,860 0.09

1975 452.67 4,716 5,757 0.10

1976 393.26 5,254 6,224 0.07

1977 432.81 5,624 6,519 0.08

1978 364.53 7,045 7,162 0.05

1979 320.88 7,549 7,398 0.04

1980 375.54 7,638 5,317 0.05

1981 588.55 7,966 8,028 0.07

1982 672.83 7,997 7,639 0.08

1983 673.33 8,144 7,912 0.08

1984 798.77 8,622 8,467 0.09

1985 715.19 9,487 9,025 0.08

1986 531.67 9,612 9,265 0.06

1987 400.48 9,202 9,075 0.04

1988 391.98 9,469 9,128 0.04

1989 326.93 10,605 9,665 0.03

1990 228.85 10,796 10,806 0.02

1991 163.25 11,284 11,037 0.01

1992 132.49 11,560 11,377 0.01

1993 141.81 11,772 11,371 0.01

1994 178.44 11,783 11,386 0.02

1995 234.93 12,448 11,382 0.02

1996 188.70 10,958 11,603 0.02

1997 182.02 10,860 12,146 0.02

1998 158.21 11,167 12,829 0.01

1999 128.89 11,212 13,294 0.01

2000 87.45 10,456 12,987 0.01

2001 114.28 10,443 13,284 0.01

2002 160.06 10,871 14,332 0.01

2003 218.83 10,660 10,015 0.02

2004c 113.29 10,192 0.01

2005 149.06 10,826 0.01

2006 161.88 10,104 0.02

2007 150.50 9,745 0.02

2008 72.51 8,148 0.01

a. In accordance with the designation adding a class of LANL 

employees to the SEC, it is not feasible to reconstruct photon doses 

to LANL workers before 1956; nonpenetrating doses cannot be 

reconstructed before 1949; and neutron doses can be reconstructed 

only from 1946 onward (NIOSH 2007b). 

b. A criticality accident during this year delivered large doses to a small 

number of people, driving the average total dose up significantly.

c. For 2004 through 2007, an update to LANL (2004) was received 

during preparation of the Evaluation Report for SEC petition SEC00109 (NIOSH 2012); for 2004 through 2007 (and part of 2008), only 

the numbers of employees monitored were available.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 17 of 79

Figure 6-1. Number of workers monitored as a function of time (LANL 2004, updated 

January 2009).

In February 1992, LANL temporarily switched to quarterly badge exchange due to installation startup 

problems with a new data management system, with the exception of 22 employees who were 

involved in a special project that required weekly exchange. Monthly exchange was reinstated in April 

1992. LANL began phasing in a quarterly dosimeter exchange frequency for personnel who received 

low occupational radiation exposure beginning in April 1996 (LANL 2001). The cutoff was a less-than10-mrem collective dose for a work group (mailstop) based on 1995 exposure records. Beginning in 

July 1996, mailstops with collective doses of 50 mrem or less for 1995 were placed on quarterly 

exchange. As of February 2002, quarterly periods were used for about 40% of dosimeter issues 

(LANL 2003).

Exchange frequencies for LANL dosimeters are summarized in Table 6-3 and later in the discussion 

of methods for estimation of missed dose.

Table 6-3. Typical exchange frequencies for dosimeters in use during different periods.

Dosimeter type Date range Exchange frequency

PIC Before 1945 Daily

Film badge 1943–1979a Monthly for most, biweekly for some, up to daily for some 

operations such as radioactive lanthanum processing

TLD 1980–1995 Monthly

TLD 1996–present Monthly for most, quarterly for some

NTP 1949–approx. 1951 4-wk interval

NTA film 1951–1995 4-wk interval

TED 1995–present Quarterly

a. According to the designation adding a class of LANL workers to the SEC, it is not feasible to reconstruct 

photon doses before 1946, and infeasible to reconstruct nonpenetrating doses before 1949 (NIOSH 2007b).

6.2.1.3 Use of Control Dosimeters

A review of the documents in the Photodosimetry Evaluation Book from 1944 to the present yielded 

little information on how control dosimeters were used over time at LANL. The information that was 

found is as follows:

Year

Number of workers monitored
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 18 of 79

• An October 1944 report (in LASL 1959) received at LANL from MED about the use of lead 

cross dental film packets to monitor gamma, X-ray, and beta exposure mentioned that 

... each group of test films is developed with a calibration film. One 

unexposed test film is included in the group as a check. After processing, if 

the unexposed film shows evidence of radiation fog, as indicated by an 

image or shadow of the lead cross, the evaluation of exposures received by 

films carried by personnel is questionable, if not useless.

• A November 2, 1944, report (in LASL 1959) about film monitoring conducted at an unidentified 

MED warehouse said:

Control badges (or, in this case, films) are to be mailed from, and returned 

to, Rochester with the others but are not to be issued to any worker. 

Indeed, they are to be kept in a place which is known to be free of radiation. 

Such a control badge will then reveal whether anything goes wrong with the 

films (badges) during shipment or at any other time except while they are 

actually being worn.

The existence of the two documents cited above indicates that the use of control badges was a 

concept with which LANL personnel were familiar from the early years of operations.

• A Los Alamos “Nuclear Track Plate Evaluation Procedure” dated February 27, 1956 (in LASL 

1979), stated: 

It has been found that NTPs which are not worn, but are processed and 

“read,” will indicate some exposure. This exposure varies somewhat with 

the seasons, etc., and is attributed to cosmic radiation. Therefore, 

whenever a group of NTPs are packaged for issue, a “blank” NTP is also 

packaged and immediately sent to P-10 for processing and reading. Then 

the blank’s exposure will be subtracted from the personnel’s neutron 

exposure during the process of evaluating the NTPs.

• A November 1974 sheet of potential values for the Remarks Code used in dosimetry records 

(in LASL 1977) had a Code 025 that signified “Used as Blank,” and an August 1976 revision 

had a Code 247 that signified “Control film inadvertently issued to a visitor- D. P. 4/7/76.” 

• A July 23, 1991, memorandum from to the Dosimetry Evaluation Book (in LANL 1996) 

indicated that, in response to a finding from a DOE Laboratory Accreditation Program 

(DOELAP) onsite audit, “several 'unirradiated dosimeters' have been added to the blind audit 

program since January, 1991.”

• An April 29, 2003, memorandum (in LANL 2003) from Jeffrey Hoffman to Michael 

McNaughton, both LANL employees, discussed two environmental dosimeters that had been 

labeled “vault dosimeters” in error.

• LA-UR-03-1037, “LANL 8823 Neutron Blind Audits at TA-36” (in LANL 2003), indicated that 

the blind-testing protocol at that time included 19 WB TLDs, 13 extremity TLDs, and 13 TEDs 

each quarter, of which four, three, and three, respectively, were “non-irradiated controls.”

Based on the above excerpts and information from a former worker in the LANL dosimetry group 

(Widner 2003), a set of calibration films was developed when a batch of personnel films was 

developed. A control that had been kept in the dosimetry offices along with the calibration films 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 19 of 79

(apparently in a vault in later years), and other controls that had been kept in the film badge racks with 

the personnel films, were photographically developed at the same time. These badge racks, which 

were for storage of dosimeter badges not being worn by the persons to whom they were issued (e.g., 

overnight or over weekends), were in areas where only background radiation was expected under 

normal circumstances; control badges were also kept in these racks for the normal film badge issue 

periods. The developed “calibration film control” was used to zero the densitometer before the 

densities of the calibration film were read. The density readings of the calibration film were plotted 

against the exposure given to the calibration film to permit the evaluation of the personnel film. The 

control accompanying the personnel film was used to zero the densitometer before the densities of 

the personnel films were read. On rare occasions when the personnel control film showed some 

exposure due to fallout from Nevada Test Site operations, the control film was evaluated and its 

exposure was subtracted from all personnel film for the period. 

On rare instances, the personnel control films were “lost” or removed from the film badge racks by 

persons unknown. On such occasions, a personnel film badge that had been issued to a person who 

never entered the radiation work areas during the film badge issue period was used as the control for 

zeroing the densitometer before the other personnel films were read (Widner 2003). In those cases, 

Remark Code 25 would be attached to the zero dose reading recorded for the person whose badge 

was used as the blank. Remark Code 247 was appended to the evaluated exposure of the person 

(probably a visitor) to whom the control film had been inappropriately given.

6.2.1.4 Reporting Conventions

Monitoring for external radiation exposures at LANL began in 1943 with the use of PICs. Gamma 

exposures were recorded in units of roentgen. Before 1949, some Laboratory personnel who worked 

with the cyclotron and other neutron sources wore Victoreen pencil PICs for determination of their 

neutron dose [LASL 1969 (11/12/68 memorandum)]. The results were recorded in n units, which 

were defined as “the quantity of neutron radiation that will produce the same ionization in a 100-R 

Victoreen chamber (red Bakelite) as 1 R of gamma radiation.” While n-unit data were recorded in 

medical records of some individuals, they were apparently never converted to the computerized 

database of exposure records and will probably not be available with information provided for 

claimants (Widner 2004). See Section 6.2.2.2 for more details about the n unit. 

When brass film badges (film placed directly in a brass container, with no window of any sort or filters 

of any other metal) came into use by more and more LANL groups between 1943 and 1945, only 

gamma exposure in roentgens was evaluated.

Starting in January 1949, the following values were recorded on personnel exposure sheets: 

• PIC readings (R)

• Gamma exposure (R)

• Beta exposure (rep)

Beta exposure was reported only when “it forms a significant part of the total exposure” [LASL 1959

(1/10/49 memorandum)]. No entry in the beta exposure column indicated that a negligible amount of 

beta exposure was present. For DP Site personnel working with plutonium and for soft X-ray 

evaluations, special calibration curves were used, and gamma roentgen exposures were multiplied by 

0.6 to convert to gamma rem [LASL 1959 (7/1/56 document)]. [The comparable DOELAP roentgento-rem dose conversion value is 0.38 for 16-keV photons (DOE 1986)]. 

Starting in 1949, NTPs came into use. As of February 1956, NTPs were evaluated by assuming all 

10- to 100-μm tracks represented 3.75-MeV neutrons and all longer tracks represented the maximum 

average energy of the higher energy neutrons in the workplace. NTA film came into use in the brass-
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 20 of 79

cadmium badge in 1951. While tolerances for neutrons were stated in terms of neutrons per square 

centimeter per second in the early years, neutron doses were reported in terms of rem in 1953 and 

possibly earlier [LASL 1959 (Jan. 1953 document)].

Before April 1957, exposures for HT shop and Sigma areas (where uranium was handled) were 

recorded as pure beta exposures, but some gamma exposures should have been recorded for 

workers at those facilities [LASL 1959 (4/3/57 document)]. Based on film badge data from early 1953, 

a factor of 0.1 was used in the conversion of radiation exposure records for IBM entry to calculate 

retrospectively the gamma exposure from the gamma-plus-beta exposure measured by the 

dosimeter. The theoretical ratio of gamma to beta plus gamma was reported to be about 1:17 for 

normal uranium [LASL 1959 (4/3/1957 document)]. 

The roentgen-to-rem dose conversion factor (DCF) was changed to 0.5 from July 1957 to March 1963 

for body badges exposed to low-energy X-rays and plutonium; the use of this factor was discontinued 

around March 20, 1963 [LASL 1969 (3/20/63 memorandum)].

As of 1960, the following external radiation dose data were recorded (Littlejohn 1961):

• Gamma dose (rem)

• Beta dose (rad)

• Thermal neutron dose (from cadmium optical density minus brass optical density, rem)

• Fast neutron dose (from NTA film, rem)

On January 1, 1972, the Laboratory evaluated assignment of neutron dose equal to TLD-measured 

gamma dose for workers who handled 238Pu [LASL 1977 (12/3/71 memorandum)]. This appears to 

have been a special study for a relatively small group of workers, possibly using hand-fabricated 

badges with loose chips, because TLDs were in relatively short supply at that time (Widner 2003).

Starting in early 1980 with the conversion to the Model 7776 TLD badge, the following external 

radiation dose data were recorded [LANL 1986 (3/21/80 memorandum, p. 180 of the PDF file)]:

• “Non-penetrating Rad” represented the total skin dose from external ionizing radiation; it was 

equal to old beta-rad plus gamma-rem or old gamma-R. 

• “Penetrating-rem” represented the total WB dose from external ionizing radiation; it was equal 

to old gamma-rem dose. It was evaluated based on the copper-filtered TLD chip.

• “Neutron-rem” was typically equal to the albedo (body-scattered) neutron dose from the TLD 

badge; when piggyback NTA film was used, the fast neutron dose from NTA film was added to 

the albedo neutron dose, and “NTA” was entered in the Remarks field. 

• “Total-rem” was equal to penetrating-rem plus neutron-rem plus tritium-rem (where tritium-rem 

was calculated from tritium urine assays).

Since the implementation of the Model 8823 TLD badge in 1998, the following dose quantities (in 

millirem) have been evaluated and recorded (Hoffman and Mallett 1999a,b): 

• Beta shallow DE 

• Beta eye DE 

• Gamma shallow DE 

• Gamma deep DE 

• Gamma eye DE 

• Neutron deep DE 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 21 of 79

• Total shallow DE 

• Total deep DE 

• Total eye DE 

• Total deep neutron DE 

Shallow doses, which are reported to a tissue depth of 7 mg/cm2 (Hp (0.07)), correspond to the old 

nonpenetrating doses. Doses to the lens of the eye are reported to a tissue depth of 300 mg/cm2

(Hp

(3)). Deep doses, which are reported to a tissue depth of 1,000 mg/cm2 (Hp (10)), correspond to the 

old penetrating doses. Neutron DE values correspond to albedo neutron DE (rem).

Table 6-4 summarizes quantities that have been recorded in LANL worker exposure records over 

time.

Table 6-4. Recorded dose practices over time (LASL 1959, 1969, 1977, 1979; LANL 1986, 1989, 

1996, 2001, 2003).a

Period

Values recorded in 

personnel exposure records Compliance dose quantities EEOICPA dose quantities

1943–1948 PIC reading

Gamma exposure

Shallow DE = not measured

Deep DE = gamma exposure

Electron dose = not measured

Photon exposure = gamma 

exposure

NADE = not measured

1949–1950 PIC reading

Gamma exposure

Beta exposure

Shallow DE = beta exposure + 

gamma exposure

Deep DE = gamma exposure

Electron dose = beta exposure

Photon exposure = gamma 

exposure

NADE = not measured

1951–1959 PIC reading

Gamma exposure

Beta exposure

Fast neutron dose 

Shallow DE = beta exposure + 

gamma exposure + fast 

neutron dose

Deep DE = gamma exposure 

+ fast neutron dose

Electron dose = beta exposure

Photon exposure = gamma 

exposure

NADE = fast neutron dose

1960–1979 Gamma dose 

Beta dose

Thermal neutron dose

Fast neutron dose

Shallow DE = beta dose + 

gamma exposure + nth dose + 

nf dose

Deep DE = gamma exposure 

+ nth dose + nf dose

Electron dose = beta dose

Photon exposure = gamma 

dose

NADE = nth + nf doses

1980–1997 “Non-penetrating Rad”

“Penetrating-rem”

“Neutron-rem”

“Total-rem”

Shallow DE = nonpen + pen + 

neutron

Deep DE = pen + neutron

Electron dose = nonpen

Photon deep dose = pen

NADE = neutron

1998–

present

Beta shallow DE 

Beta eye DE

Gamma shallow DE

Gamma deep DE 

Gamma eye DE 

Neutron deep DE 

Total shallow DE 

Total deep DE

Total eye DE

Total deep neutron DE

Shallow DE = total shallow DE

Deep DE = total deep dose 

DE 

Electron dose = beta shallow 

DE

Photon deep dose = gamma 

deep DE

NADE = neutron deep DE

a. DE = dose equivalent; NADE = neutron ambient dose equivalent; nonpen = nonpenetrating; pen = penetrating.

In the reporting of doses for LANL workers to NIOSH for the Dose Reconstruction Project, LANL 

personnel have used the following conventions (Widner 2005):

• Blank entries or “----” entries in tables of doses indicate a “null value” (i.e., no monitoring was 

performed for the subject individual for that period).
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 22 of 79

• Dose entries that are all zeros (0.00 or 0.000) indicate that monitoring was performed for the 

subject individual during that period, but results were below the minimum detectable dose.

SHALLOW doses reported by LANL are the shallow doses as measured by TLDs or film 

badges; they do not include neutron dose or tritium dose. This contrasts with records reported 

as “SKIN” doses, which were reported in records submitted for dose reconstruction through 

late 2005.2

• Early LANL case records contain a column labeled “SKIN.” The SKIN doses reported are 

derived by LANL as shallow dose plus neutron dose plus tritium dose.

• If, for a given month, the dose report form identifies “Badge Type” as “Monthly,” but there are 

as many as five lines of data (see example below), multiple badges were worn during the 

month. The doses for the individual badges should be added to obtain the dose totals for the 

month.3

 

External 

dose (rem) Skin Deep Neutron Tritium

June Monthly 0.000 0.000

June Monthly 0.000 0.000

June Monthly 0.010 0.010

June Monthly 0.000 0.000

June Monthly 0.040 0.020

• If the dose reported is as follows, the badge in use had two elements – dental X-ray film and 

NTA film. In the case below, the first line of data is from the NTA film and the second is from 

the dental X-ray film. In these cases, the neutron dose entry on the second line for the month 

is essentially a placeholder, and should not be corrected for missed dose.

External 

dose (rem) Skin Deep Neutron Tritium

April Monthly 0.010 0.010

April Monthly 0.060 0.010 0.000

6.2.1.5 Recordkeeping

From 1943 through 1952, external radiation evaluation results were recorded in standard “LA 

notebooks” that eventually found their way to the document room for permanent filing [LASL 1959

(1/20/51 document)]. December 9, 1946, was the first date for which a record of visitor badge 

issuance was found in LA notebooks, and the use of the notebooks for visitor badge data ended in 

January 1951 [LANL 1986 (10/30/79 memorandum)].

In August 1950, H Division started supplying PICs and film badges for use by the GMX-1 group in 

“extraordinary work” [LASL 1959 (February 1956 memorandum)]. Before that, dosimetry for the 

GMX-1 group was reportedly handled by GMX-1 personnel at GT Site, and apparently no long-term 

records of these early measurements were retained.

In January 1953, a Cardex system for filing exposure data on paper cards was put into use [LASL 

1959 (2/16/56 memorandum)]. In January 1956, LANL started noting all NTPs issued on Personnel 

 2 When an automated tool is used for assigning doses to appropriate radiation types, the dose reconstructor must choose 

whether nonpenetrating doses were assigned as “SHALLOW” or “SKIN.” 

3 For estimating missed dose, all zeroes should be included in the QC file for the multiple badges reported in one month 

to determine missed dose.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 23 of 79

Exposure Cardex records. Before that, only plates that were read were recorded [LASL 1959

(2/27/56 document)].

In 1957, a computerized system was first employed at LANL to record personnel exposures to 

radiation [LASL 1959 (4/3/57 memorandum)]. In 1959, IBM equipment was first used to evaluate film 

exposures (Littlejohn 1961). This recordkeeping has been computerized since that time. 

In September 1978, it was reported that for the previous 10 years no entry was made for a visitor’s 

film badge in the computerized records system for film badge exposures less than 0.04 rem [LANL 

1986 (9/28/78 document)]. For reporting to the primary employers of visitors, nonzero exposures 

recorded by the LANL Cycolac film badge were defined to be total exposures of 0.04 rem or more. 

(For regularly issued film badges, all measured values from 0.00 rem upward were recorded and 

attributed to the worker in the H-1 dosimetry records system.) This procedure for visitor film badges 

was used so reporting of doses less than 0.04 rem was not required, which would have been an 

enormous undertaking due to the need to find each visitor to obtain employer name and address for 

reporting purposes. On January 1, 1976, LANL had to start reporting zero exposures for all U.S. 

Energy Research and Development Administration (ERDA; a DOE predecessor agency) visitors 

because ERDA had just established a radiation exposure record system for all its employees [LANL 

1986 (10/30/79 memorandum)].

As of October 1989, the policy for NTA film dose entries was as follows: For workers who wore 

piggyback NTA dosimeters with their TLDs, NTA doses were summed with TLD neutron doses for 

monthly and annual reports. Before implementation of the new external dosimetry data management 

system, entry of NTA neutron doses used the same guidelines as those for TLD neutron doses: 

doses less than or equal to 4 mrem were entered as 0 rem, 5 to 14 mrem were entered as 0.01 rem, 

15 to 24 mrem were entered as 0.02 rem, 25 to 34 mrem were entered as 0.03 rem, 35 to 44 mrem 

were entered as 0.04 rem, and so on (LANL 1989; Widner 2004). 

Between 1981 and September 1990, shallow, deep, or neutron exposures less than 10 mrem were 

reduced to 0 mrem [LANL 1996 (9/14/90 document)]. LANL decided to change associated 

procedures such that shallow doses, deep doses, and neutron doses less than 5 mrem were rounded 

to 0 mrem, and doses between 5 and 10 mrem were rounded to 10 mrem effective with September 

1990 TLD evaluations. These changes were not fully implemented until 1991 [LANL 1996 (2/5/91 and 

4/19/91 memoranda)]. 

In early 1992, LANL installed and started a new data management system called the External 

Dosimetry Badge System. 

6.2.1.6 Quality of External Dosimetry Data

At LANL, dosimeters were selected, issued to workers, and processed; the resulting measurements 

were recorded and used to estimate doses. There appears to be no use of recorded notional doses, 

although there are issues of missed dose for low-dosed dosimeters (see Section 6.5) and recorded 

doses for individual dosimeters at levels less than the statistical minimum detection level (MDL). 

LANL dosimetry capabilities during the early years were in line with the stage of development of 

associated technologies at the time. Radiation dosimetry technology developed considerably during 

the period of LANL operations, and LANL kept up with technological developments as they became 

accepted. Administrative practices are described in the Photodosimetry Evaluation Book (LASL 1959, 

1969, 1977, 1979; LANL 1986, 1989, 1996, 2001, 2003) and LANL technical reports, and detailed 

information for each worker is in the NIOSH claim documentation. The claim documentation provides 

specific information to be evaluated on the recorded dose of record. There do not appear to have 

been significant administrative practices that jeopardized the integrity of the dose of record. 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 24 of 79

6.2.2 Dosimetry Technology

LANL health physicists learned from experiences at other MED sites that followed the general 

evolution in dosimeter technology using PICs in addition to one- or two-element dosimeters in the 

1940s and early 1950s, multielement film dosimeters in the later 1950s, and TLDs from the 1970s to 

the present. LANL dosimeter designs differed somewhat from the designs at the MED Metallurgical 

and Clinton Laboratories in the early to mid-1940s, but capabilities to measure doses were similar.

The adequacy of the dosimetry methods to measure radiation dose accurately is determined from the 

radiation type, energy, exposure geometry, etc., as described in later sections. The dosimeter 

exchange frequency was gradually lengthened, and generally corresponded to the period of 

regulatory dose controls (LASL 1959, 1969, 1977, 1979; LANL 1986, 1989, 1996, 2001, 2003). At the 

beginning of Laboratory operations, a dose control of 1 mSv/d (100 mrem/d) was in effect. This was 

changed to 3 mSv/wk (300 mrem/wk) and later to 50 mSv/yr (5,000 mrem/yr) in the late 1950s. 

These changes were in accord with AEC regulations at the time and with the appropriate NBS 

handbooks (Widner 2003). Table 6-5 summarizes major events in the LANL personnel dosimetry 

program. 

Table 6-5. External dosimetry events (LASL 1959, 1969, 1977, 1979; LANL 1986, 1989, 1996, 2001, 

2003).

Period Event

1943 Monitoring for external radiation exposures began, using PICs.

1943 Film badges first used at LANL by the GMX-1 Radiography Group. These badges 

apparently used Lead Cross Type K dental film in a brass holder. 

Early 1944 LANL H Division health physicists gathered information about film monitoring practices from 

other MED sites.

May 1944 Film badges, made at LANL, were distributed to some workers by H Division on a trial basis.

Feb 1945 System of monitoring exposure of personnel with photographic film was established by H 

Division, with badges exchanged monthly. Continued to use PICs.

Late 1945 After Omega Site criticality accident, “catastrophe badges” were developed and assigned to 

workers who could be involved in a criticality accident.

Aug 1947 “Brass Film Badge” in use included three types of film in a brass container with no windows 

or filters of other metals. 

Sep 48–Mar 49 A new badge was phased in. Brass clip provided only partial shielding of film surface.

29 Aug 1949 First NTPs issued for evaluation of fast neutrons. 

20 Apr 1950 Started changeover to new badge that used brass and lead filters plus unfiltered area. 

7 Mar 1951 Studied ring badges and wrist badges and variability of ratios between them. 

24 Mar 1951 New finger film badge developed. Used Eastman Kodak Type K or DuPont sensitive Type 

552, brass and cadmium filters.

Apr 1951 Changed to brass and cadmium badge. Brass and cadmium filters, plus open window. 

Used DuPont 502 dental film and Eastman Kodak Type B packet that contained NTA fast 

film and Fine Grain Positive film.

1954 Thermal neutron measurements started. 

1955 Routine measurement of thermal neutrons was discontinued.

1956 Routine evaluation of thermal neutron exposures restarted.

Oct 1962 Multielement Cycolac film badge was put into use. It included multielement filter, another 

multielement filter with Li-6, and unfiltered area. Two film packets: DuPont 543 (with 

DuPont 502 film) and Kodak Type B (with NTA fast film and Fine Grain Positive film).

31 Jan 1963 Practice of using wrist-to-finger ratios greater than 1 was discontinued. 

Mar 1963 Comparison of Cycolac and brass-cadmium badges was conducted.

Aug 1968 Tried to duplicate Battelle PNL badge performance study of 1966, in which it did not 

participate.

Sep 1970 Started issuing film badges with TLD ribbon containing plastic insert to selected individuals 

at DP Site. Purpose of these badges with TLD-600 and TLD-700 ribbons was to evaluate 

individual exposures between film development and evaluation. 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 25 of 79

Period Event

1971–1972 Prototype albedo-neutron TLD badge was studied at DP West and compared to Cycolac film 

badge data.

Jul–Dec 1972 Performance of film badges, NTA film, and prototype albedo-neutron-TLDs were compared 

at DP West plutonium facilities over a 6-month period.

May 1978 Began issuing Model 7776 TLD badges to visitors. Certain LANL groups got them starting in 

September. Additional LANL groups changed from film to TLD throughout 1978 and 1979, 

as TLD cards and badges were acquired. 

1978–1979 Participated in National Research Council/University of Michigan pilot dosimeter testing 

program.

1980 At LAMPF, TLD badge was supplemented by Kodak NTA film placed in a Cycolac piggyback 

holder attached to TLD badge. 

1 Jan 1980 Dosimetry section completed changeover from film dosimetry badges to TLD badges as 

dosimeter of record. Model 7776 badge had cadmium and noncadmium versions. Cadmium 

version was initially used at LAMPF.

Late 1980 Changed to use of noncadmium badge at LAMPF to gain sensitivity.

Jun 1981 Concurrent monitoring with NTA film and TLDs was conducted.

Jun 1981 Discontinued routine issue of NTA film for LAMPF personnel. 

Dec 1984 Applied to participate in pilot testing of DOELAP.

16 Jan 1985 Selected for participation in pilot testing of DOELAP.

Jul 1987 TLD plus NTA badges used with other detectors during special neutron spectral 

measurements at LAMPF.

May 1988 Neutron measurements were made at seven locations in TA-55 Room 209 using ESP-2 

neutron instrument and TLD badges on Lucite phantom.

Jun 1991 Expressed interest in participating in some categories of DOELAP testing of extremity 

dosimeters (not the categories with low-energy beta emitters or neutrons).

Aug 1992 Comparison conducted of bubble dosimeters with Model 7776 TLDs.

1993 Purchased automated reader for use with chemically etched CR-39 detectors.

1995 Introduced TED as replacement for NTA film as high-energy neutron dosimeter at LAMPF.

22 Feb 1996 Moved up deployment of new eight-element TLD (worn with 7776 TLD badge) for use in 

dynamic determination of employee-specific NCFs. 

Apr 1996 Began to phase in quarterly dosimeter exchange frequency for personnel who received “low”

occupational radiation exposure.

1996 Submitted track-etch neutron dosimeters for irradiation at the Paul Scherrer Institute in 

Switzerland.

Feb 1998 Review of Model 8823 TLD as neutron dosimeter issued.

1 Apr 1998 Model 8823 became dosimeter of record. Eight-element TLD (called Model 8823 by 

manufacturer Harshaw/Bicron) uses Model 7774 TLD cards. 

Oct 1998 Began using wrist dosimeter for routine monitoring. 

1999 LANL TED passed DOELAP performance testing for neutrons.

1999 Participated in 4th Intercomparison of Personal Dosimeters Used in U.S. DOE Accelerator 

Facilities. 

2000 Submitted TLDs and track-etch badges for irradiation by monoenergetic acceleratorproduced neutrons from 0.144 MeV to 19 MeV at PTB in Germany.

6.2.2.1 Beta/Photon Dosimeters

The following paragraphs describe LANL dosimeters and periods of routine use to provide the 

recorded dose of record.

Pocket Ionization Chambers. Monitoring for external radiation exposures at LANL began in 1943 

when PICs were issued to a few persons thought to have the highest potential for receiving exposures 

at or above the “tolerance” limit [LANL 1986 (10/6/81 document)]. As of January 19, 1944, Victoreen 

PICs were in use. In August 1950, H Division furnished Keleket Pocket Chambers along with film 

badges for use by GMX-1 in “extraordinary work.” 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 26 of 79

Brass Film Badge, 1943 to 1948. Before 1951, external dosimetry was provided by three different 

organizations at LANL – H Division, the CMR (Chemistry and Metallurgical Research; also refers to a 

building)-12 Radiochemistry Group, and the GMX-1 Radiography Group. The first use of film badges 

at LANL was apparently in 1943 by the GMX-1 Radiography Group (Widner 2004). 

On January 19, 1944, H Division personnel inquired with Colonel Stafford Warren of MED at Clinton 

Laboratories (which became Oak Ridge National Laboratory) about using strips of film to check 

exposures to radiation. They requested film and instructions for use in checking their Victoreen PICs. 

By May 18, 1944, LANL had started to distribute film monitors to various people [LASL 1959 (5/18/44 

memorandum)]. On June 14, 1944, Colonel Warren gave authorization to order Lead Cross Type K 

Dental Film from Eastman Kodak Company. In February 1945, H Division instituted a system to 

monitor exposure of workers with photographic film [LASL 1959 (2/17/45 memorandum)]. The 

badges apparently used Eastman Kodak Lead Cross Type K Dental Film in a brass holder. Film 

packets of that type were ordered with a 0.5-mm-thick lead cross over one face (covering about 60% 

of that face), with the edges of the cross folded over to cover a small portion of the opposite side of 

the packet [LASL 1959 (6/13/44 drawing attached to 10/27/44 memorandum)]. In the MED method 

for use of this film, penetrating photon doses were estimated from the areas on the edge of the film 

packets where the ends of the lead cross folded over the edge of the packet; the combined thickness 

of the lead and the 0.033-in.-thick brass holder absorbed beta rays [LASL 1959 (report attached to 

10/27/44 memorandum)].

In the film badge in use in August 1947, film was placed directly in a 0.4- to 0.5-mm-thick brass 

container with no window or filters of other metal [LASL 1959 (8/4/47 memorandum)]. These badges, 

which were intended to measure only gamma radiation, were made at LANL. Three types of film were 

initially used together: Eastman Kodak Industrial Type K X-Ray Safety Film (0.01 to 6.0 R), DuPont 

D-2 Special Type X-Ray Film (two films, together covering 0.13 to 30 R, and DuPont Adlux Film (65 to 

1,500 R). Beginning around July 1947, only Eastman Kodak Type K film was used routinely, but all 

films continued to be used “in cases where an operation is considered at all hazardous” [LASL 1959

(8/4/47 memorandum)]. 

There is some uncertainty about the point in time when LANL switched from the film badge that used 

the lead cross film to the version in use in August 1947 that used films with no windows or filters of 

any type. 

Early LANL film packets were collected and read once a month and new films were supplied. Film 

badges did not replace the daily measurements of the PICs; they provided a check of daily 

measurements and formed a permanent record of exposure. About 1 week after the August 8, 1945, 

fatal criticality accident at TA-2, “catastrophe badges” that could measure as much as 3,000 R with 

film, red phosphorous capsules for measuring fast neutron flux, and brass in the badge itself for 

measuring slow neutron flux were assigned to “anyone who could be involved in a criticality accident” 

[LASL 1959 (6/14/45 document)].

Brass Clip Badge, 1949 to 1950. About January 1949, there were changes in the type of film badge 

used at the Laboratory. The new badge had a DuPont film packet inserted in a brass clip that 

provided only partial shielding of the film surface. The brass absorbed almost all beta radiation while 

having negligible effect on gamma radiation. By comparing the optical densities under brass with 

those not under brass, a separate evaluation of beta exposure could be made. 

Brass-Lead Film Badge, 1950 to 1951. On April 20, 1950, the Laboratory started changeover to a 

new film badge. It used brass and lead filters plus an unfiltered area. The two-filter system was used 

to support energy determination for beta and gamma radiation so appropriate correction factors could 

be applied. This badge was used until April 1951.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 27 of 79

Brass-Cadmium Film Badge, 1951 to 1962. In April 1951, the Laboratory initiated changeover to a 

brass and cadmium film badge. This badge incorporated brass and cadmium filters and an open 

window. It used two film packets: a DuPont 543 packet that contained DuPont 502 dental film and an 

Eastman Kodak Type B packet that contained NTA fast neutron film and Fine Grain Positive film.

Cycolac Film Badge, 1962 to 1978. In October 1962, a new multielement Cycolac film badge was 

issued to about 100 persons who had histories of appreciable or out-of-ordinary radiation exposures. 

More widespread use followed. Named after the brand of plastic used in its holder, the Cycolac 

badge used a multielement filter, a second multielement filter with 6

Li, and an unfiltered area. It used 

two film packets: a DuPont 543 packet that contained DuPont 502 dental film and an Eastman Kodak 

Type B packet that contained NTA fast neutron film and Fine Grain Positive film. In addition, it 

contained indium, gold, and sulfur foils for accident dosimetry. The Cycolac badge had the 

advantages of relative gamma and X-ray energy independence (within 30%) from about 30 keV to 

1,400 keV, the ability to evaluate thermal neutrons in the presence of X- and gamma radiations below 

400 keV, and improved directional independence. 

Model 7776 Thermoluminescent Dosimeter Badge, 1978 to 1998. A prototype albedo-neutron 

TLD was studied at DP West in the last half of 1972. The Laboratory began issuing TLD badges to 

visitors in May 1978. Certain Laboratory groups began to receive them in September because it took 

several months for complete conversion. The Model 7776 TLD badge had cadmium and 

noncadmium versions. It incorporated copper, Cycolac plastic, and cadmium filters, and used three 

TLD-700 chips (one covered with copper, one with thin Cycolac plastic, and the third with thicker 

plastic) and one TLD-600 chip (enriched in 6

Li). In the cadmium badge, the third TLD-700 chip and 

the TLD-600 chip were shielded by cadmium pockets, as opposed to plastic covers in the 

noncadmium badge. The Dosimetry section completed the changeover from film dosimetry badges to 

TLD badges as the dosimeter of record on January 1, 1980. The Model 7776 dosimeter was not 

designed to perform low-penetrating beta dosimetry and was not accredited by DOELAP for lowenergy beta particles or beta and low-energy photon mixtures (Hoffman and Mallett 1999a). 

Model 8823 Eight-Element Thermoluminescent Dosimeter, 1998 to Present. The Model 8823 

TLD is a custom LANL design that contains two Harshaw/Bicron-NE TLD cards (Hoffman and Mallett 

1999a,b). The Model 8823 cardholder is made of black acrylonitrile butadiene styrene (ABS) plastic 

to prevent the exposure of light-sensitive TLD elements. The holder contains a 21-mil-thick cadmium 

box that is painted red in which the neutron TLD card is placed. The cadmium box has an open 

window under positions 7 and 8 (next to the body of the wearer) and over positions 5 and 6 (toward 

the incident radiation) to facilitate the combined albedo and anti-albedo design. The technical 

advantage of the moderating box is that it reduces the dependence of TLD neutron response on the 

distance of the dosimeter from the worker’s body. The holder is designed to provide 600 mg/cm2 ABS 

plastic filtration over positions 1 and 4, mainly to determine photon deep dose. Positions 2 and 3 are 

beta windows that are covered with one and two layers of aluminized Mylar, respectively. The 

aluminized Mylar is coated with black paint on the back to maximize light attenuation.

The photon and beta dosimeters used over time at Los Alamos are summarized in Table 6-6.

6.2.2.2 Neutron Dosimeters 

LANL has used five general types of neutron dosimeters that differ dramatically in their response to 

neutron radiation. The following paragraphs describe the personnel neutron dosimeters used at LANL 

and their periods of use. 

Pocket Ionization Chamber, Before 1949. Before 1949, some personnel who worked with the 

cyclotron and other neutron sources wore the Victoreen pencil PIC to determine their neutron doses 

[LASL 1969 (11/12/68 memorandum)]. The chamber consisted of a Bakelite cylinder with a clear 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 28 of 79

Lucite top and an aluminum cap and ring on the bottom. The chambers were 5 in. long and 0.5 in. in 

diameter and had a pocket clip. The chambers were not self-reading. An internal electrode and the 

inner wall of the Bakelite chamber were coated with graphite. Doses received were recorded in 

n units. An n unit was defined as “the quantity of neutron radiation that will produce the same 

ionization in a 100-r Victoreen chamber (red Bakelite) as 1 R of gamma radiation.” A 1968 study of 

the response of the Bakelite dosimeters indicated that 1 n unit corresponded to about 5.9 rad of 

neutron exposure (±25%) or 59 rem if a quality factor of 10 is applied [LASL 1969 (11/12/68 

memorandum)]. A conclusion of that study was that the Bakelite dosimeters lacked adequate neutron 

sensitivity to be useful for the evaluation of neutron exposures. Dale Hankins of H-1 stated that “large 

errors can be made if the gamma-ray contribution to the response of the dosimeter was assumed to 

be neutron dose or that the neutron contribution to the dosimeter readings was assumed to be 

gamma-ray dose” [LASL 1969 (11/12/68 memorandum)].

Nuclear Track Plates. In late August 1949, the first NTPs were issued for the evaluation of fast 

neutrons. The first written report of fast neutron exposure was in May 1950. From 1943 to 1949, no 

monitoring for fast neutron doses was performed with the exception of activation analyses performed 

for several criticality accidents. As of October 4, 1956, NTPs with 1,000-μm emulsion on glass 

backing supplied by Ilford were used for fast neutron dose evaluation. They were thought to have 

been accurate to within a factor of 2 [3].

Nuclear Track Emulsion. In April 1951, changeover to a brass and cadmium film badge began. 

This badge incorporated an Eastman Kodak Type B packet that contained NTA fast neutron film and 

Fine Grain Positive film. In October 1962, the new multielement Cycolac film badge was first issued 

at the Laboratory. The Cycolac badge included an Eastman Kodak Type B packet that contained 

NTA fast neutron film and Fine Grain Positive film. It also contained indium, gold, and sulfur foils for 

accident neutron dosimetry. The main concerns with NTA film were that latent-image tracks faded 

Table 6-6. Photon and beta dosimeters used over time (LASL 1959, 1969, 1977, 1979; LANL 1986, 

1989, 1996, 2001, 2003; Widner 2003, 2004).

Period Type badge Film or media used

1943–1944 PIC Quartz fiber

1943–June 

1946

Brass film badge Eastman Kodak Lead Cross Type K Dental Film packet

July 1946–

1948

Brass film badge – no windows or 

filters

Eastman Kodak Industrial Type K X-Ray Safety Film, 

DuPont D-2 Special Type X-Ray Film, and DuPont 

Adlux Film. After June 1947, only Eastman Kodak 

Type K was used routinely. 

1949–1950 Brass “clip” badge – provided only 

partial shielding of film surface

Eastman Kodak Type K X-Ray Safety Film

1950–1951 Brass film badge – brass and lead 

filters plus unfiltered portion

Eastman Kodak Type K X-Ray Safety Film

1951–1962 Brass-cadmium badge – brass and 

cadmium filters plus open window

Two film packets: DuPont 543 (contained DuPont 502 

dental film), Eastman Kodak Type B (contained NTA 

fast neutron film and Fine Grain Positive film)

1962–1978 Cycolac plastic badge – multielement 

filter, a second multielement filter with 

Li-6, and an unfiltered area

Two film packets: DuPont 543 (contained DuPont 502 

dental film), Eastman Kodak Type B (contained NTA 

fast neutron film and Fine Grain Positive film)

1978–March 

1998

Model 7776 TLD badge (cadmium and 

noncadmium versions) – copper, 

Cycolac plastic, and cadmium filters

Three TLD-700 chips (one covered with copper, one 

with plastic) and one TLD-600 chip (enriched in Li-6). 

In the cadmium badge the third TLD-700 chip and the 

TLD-600 chip were shielded by cadmium pockets; they 

were covered by plastic in the noncadmium badge.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 29 of 79

Period Type badge Film or media used

April 1998–

present

Model 8823 TLD badge Two TLD cards hold eight TLD elements. One card 

has three TLD-700 elements and one TLD-400. Two 

elements are filtered with ABS plastic. Two have 

minimal filtration. The second card is in a cadmium 

box; two elements form a classic albedo detector with a

TLD-600/-700 pair surrounded by cadmium except for 

an opening toward the body. Two positions are an 

incident thermal neutron detector (“an anti-albedo 

detector”) with a TLD-600/-700 pair surrounded by 

cadmium except for an opening away from the body. 

rapidly when humidity was high and that the lowest energy neutrons detectable in routine evaluations 

were about 0.8 to 1 MeV (Hankins 1973). NTA film did provide useful information about doses from 

neutrons with energies above this threshold, particularly when used in conjunction with TLDs from 

about 1980 until 1995. After some study of fading issues at LANL, NTA film was sealed in plastic with 

a desiccant to minimize fading due to high humidity [LANL 1989 (6/22/89 memorandum)].

Thermoluminescent Dosimeters. LANL began issuing TLD badges to visitors in May 1978, and 

certain groups began to receive them in September. The Model 7776 TLD badge had cadmium and 

noncadmium versions. It incorporated copper, Cycolac plastic, and cadmium filters, and used three 

TLD-700 chips (one covered with copper, one with thin Cycolac plastic, and the third with thicker 

plastic) and one TLD-600 chip (enriched in 6

Li). In the cadmium badge, the third TLD-700 chip and 

the TLD-600 chip were shielded by cadmium pockets, as opposed to plastic covers in the 

noncadmium badge. TLD badges were used by a number of groups in 1978; the LANL Dosimetry 

section completed the changeover from film dosimetry badges to TLD badges on January 1, 1980. 

The Model 7776 dosimeter relied heavily on the use of site- and operation-specific neutron correction 

factors (NCFs) for neutron dosimetry. The technique used an essentially bare TLD-600 and TLD-700 

pair in a quasi-albedo arrangement (i.e., without a cadmium or other neutron-absorbing shield anterior 

to the TLD elements). The albedo dosimeters were sensitive to the intermediate and lower energy 

fast neutrons that other dosimetry methods could not detect, but their net neutron signal was highly 

energy-dependent and required the use of site-specific NCFs to convert the response to dose. NCFs 

could vary by more than an order of magnitude. As a consequence, they were assigned at very 

conservative values such that neutron doses were typically overestimated by a factor of 2 to 3 

(Blackstock et al. 1978).

A detachable holder for NTA film was included in the Model 7776 TLD badge design to facilitate fast 

neutron measurement. This configuration was unacceptable, however, because it would invalidate 

the calibration of the badge and its DOELAP accreditation and it did not address fading problems 

(Mallett et al. 1990).

The Model 8823 TLD is a custom LANL design that contains two Harshaw/Bicron-NE TLD cards 

(Hoffman and Mallett 1999a,b). The holder also contains a 21-mil-thick cadmium box that holds the 

neutron TLD card. The cadmium box has an open window under positions 7 and 8 (next to the body 

of the wearer) and over positions 5 and 6 (toward the incident radiation) to facilitate the combined 

albedo and anti-albedo design. 

In early 1996, LANL decided to issue the Model 8823 badge in conjunction with the LANL Model 7776 

WB TLD before the acceptance of the Model 8823 as the dosimeter of record. During this interim 

period, the dedicated neutron portion of the Model 8823 was used to calculate employee-specific 

NCFs to be applied to the Model 7776 WB TLD net neutron reading [LANL 2001 (2/22/96 document)]. 

The Model 8823 dosimeter received its first DOELAP accreditation after it successfully passed 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 30 of 79

performance testing in the spring of 1997 (Hoffman and Mallett 1999a,b). The Model 8823 became 

the dosimeter of record for LANL on April 1, 1998.

The LANL Model 8823 dosimeter is DOELAP-accredited in all applicable categories (no exceptions 

are required). The general beta category (VA), which includes 90Sr/90Y and 204Tl, is selected over the 

special contact geometry uranium category (VB) and special beta category (VC), which include only 

90Sr/90Y or 204Tl. 

Track-Etch Dosimeter. The LANL TED (or PN3) contains three dosimetry-grade CR-39 track-etch 

plastic foils (Hoffman and Mallett 1999a). The foils are placed in a hemispherically shaped ABS 

plastic case on the sides of a triangular polystyrene pyramid to minimize angular dependence of the 

TED. The LANL TED, which is sensitive only to neutron radiation, is used for special field conditions. 

When issued to personnel, it is used in combination with the Model 8823 dosimeter (Hoffman and 

Mallett 1999a). The LANL TED is entered in the DOELAP pure 252Cf and moderated field category 

(Category VI). 

These five general types of neutron dosimeters that have been most widely used at LANL, which are 

summarized in Table 6-7, differed significantly in their response to neutrons of different energies, as 

shown for NTA film and albedo TLD dosimeters in Figure 6-2 (IAEA 1990). While calibration factors 

were used to adjust the response curves for 6

LiF and NTA film upward to better align with Hp(10), this 

did not remove the fundamental energy dependence of the two types of dosimeters.

6.2.3 Calibration

Potential error in recorded dose is dependent on the characteristics of the dosimetry technology 

response to each radiation type, energy, and geometry; the methodology used to calibrate the 

dosimetry system; and the similarity between the radiation fields used for calibration and that in the 

workplace. The potential error is much greater for dosimeters with significant variations in response, 

such as film dosimeters with low-energy photons and NTA or Model 7776 TLDs with neutrons. 

Table 6-7. Neutron dosimeters used over time at LANL (LASL 1959, 1969, 1977, 1979; LANL 1986, 

1989, 1996, 2001, 2003; Widner 2003, 2004).

Period Type badge Film or media used

1943–1949 No monitoring of fast neutron exposure None

1943–1953 No monitoring of thermal neutron 

exposure

None

1949–1951 Nuclear track plates NTPs with 1,000-μm emulsion on glass backing 

supplied by Ilford 

1951–1962 Brass-cadmium badge – brass and 

cadmium filters plus open window. 

(Thermal neutron monitoring stopped in 

1955, restarted in 1956.)

Two film packets: DuPont 543 (contained DuPont 

502 dental film), Eastman Kodak Type B (contained 

NTA fast neutron film and Fine Grain Positive film)

1962–1979 Cycolac plastic badge – multielement 

filter, a second multielement filter with 

6

Li, and an unfiltered area.

Two film packets: DuPont 543 (contained DuPont 

502 dental film), Eastman Kodak Type B (contained 

NTA fast neutron film and Fine Grain Positive film)

1980–1998 Model 7776 TLD badge (cadmium and 

noncadmium versions) – used with NTA 

film in a piggyback holder for some 

(<40) workers 1981 – 1995 (measured 

thermal, intermediate, and fast neutrons, 

with no separation of fast neutrons)

Essentially bare TLD-600 and TLD-700 pair in a 

quasi-albedo arrangement (i.e., without a cadmium 

or other neutron-absorbing shield anterior to the 

TLD elements)

1995–present TED Plastic hemispherical case encompassing a 

polystyrene pyramidal detector holder. The holder 

supports three CR-39 detectors at 35° angles. 

1998–present Model 8823 TLD badge (used with the Two elements form a classic albedo detector with a 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 31 of 79

Period Type badge Film or media used

TED by those with potential high-energy 

neutron exposure) (measures thermal, 

intermediate, and fast neutrons, with no 

separation of fast neutrons)

TLD-600/-700 pair in cadmium except for an 

opening toward the body. Two positions are an 

incident thermal neutron detector with a TLD600/-700 pair in cadmium except for an opening 

away from the body. 

Figure 6-2. Comparison of Hp(10) for neutrons 

with energy responses of NTA film and neutron 

albedo dosimeter containing a thermoluminescent 

neutron dosimeter chip of 6

LiF and shielded by 

cadmium (IAEA 1990).

6.2.3.1 Beta/Photon Dosimeters

Brass Film Badge: Film badges used at LANL were initially calibrated with a radium source [LASL 

1959 (8/4/47 memorandum)].

Brass-Cadmium Film Badge: In accordance with a 1956 procedure [LASL 1959 (2/16/1956 

document)], calibrations were done with radium for gamma radiation and a sheet of depleted uranium 

for beta radiation. The optical density of the filtered area was subtracted from the optical density of 

the unfiltered area, a radium calibration curve was applied, and a factor of 5/3 was used to estimate 

beta exposure in rep [LASL 1959 (2/16/1956 document)]. For nonpenetrating dose in plutonium 

facilities, a 78-kVp X-ray calibration was used rather than the depleted uranium slab. 

The standard method of film calibration as of April 1955 used a 195-mg radium source that was 

calibrated by NBS. Overall error in the radium exposures delivered reportedly did not exceed 2%, 

including uncertainties in source strength, distance, and exposure time (Kalil 1955).

By 1960, film badges were calibrated to gamma radiation with a 60Co source (Littlejohn 1961). The 

dose rate from the calibration source was measured with a Victoreen r-chamber calibrated by NBS. 

Calibrations were performed on a circular Masonite table in a relatively scatter-free room. Badges 

were calibrated with the front area of the badges toward the source, and workers were instructed to 

wear the badges in the same orientation. 

Periodic calibrations were made to nickel-coated and uncoated plutonium. The uncoated plutonium 

was encased in 0.02-in. polyvinylchloride (PVC). Surface dose rates for both sources were 

determined by extrapolation chamber measurements. Films were calibrated in direct contact with the 

source and with various filtering materials between the source and the film to simulate dry-box 

conditions. Because the plutonium calibration curves were parallel to the 60Co curve, open-window 

10|-7 10|-6 10|-5 10|-4 10|-3 10|-2 10|-1 10|0 10|1

Neutron Energy (MeV)

10|0

10|1

10|2

10|3

10|4

Relative Response

Hp(10)

NTA Film

6

LiF with Cd Shield
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 32 of 79

densities from plutonium exposures were evaluated from the 60Co curve and then corrected by

applying factors of 0.11 for coated plutonium and 0.065 for uncoated plutonium (Littlejohn 1961).

As of 1960, a 4-in.-square plate of depleted uranium was used to calibrate the film to beta radiation 

(Littlejohn 1961). The surface dose rate of the depleted uranium plate was measured with an 

extrapolation chamber. Films were normally placed in direct contact with the source or, if required, 

filtering materials such as the PVC used in the wrist badge were inserted between the source and 

film. 

An August 24, 1962, memorandum (LASL 1969) refers to the use of 60Co calibration curves for 

evaluation of beta, gamma, and thermal neutron exposures. The 60Co calibration curve was used to 

evaluate beta exposures with a relative sensitivity factor of 1/3 based on calibration with depleted 

uranium of a known surface dose rate. In addition, plutonium source calibrations were done to 

measure the response to the unfiltered area of the film. The use of nickel-coated plutonium as a 

calibration source was discontinued in January 1963 [LASL 1969 (3/20/63 memorandum)].

The brass-cadmium badges worn in plutonium areas at DP West were calibrated by placing the 

badge in contact with a 100-g piece of uncoated plutonium metal encased in PVC [LASL 1969 (3/5/63 

memorandum)]. The surface dose rate from the encased metal was measured with an extrapolation 

ionization chamber. Exposures were evaluated by reading the net optical density from the unfiltered 

area of the film, reading the exposure from a 60Co calibration curve, applying a correction factor of 

0.08 to correct for unfiltered film sensitivity to the mixture, and dividing the result in roentgen by 2 to 

get rem, based on the assumption that only 50% of the X-ray dose penetrated to the depth of the 

gonads or 1 cm of tissue [LASL 1969 (3/5/63 memorandum)]. Beginning in January 1963, the factor 

of 2 for plutonium and soft X-ray evaluations was abandoned based on a 1962 study that indicated 

that results from the brass-cadmium badge underestimated expected (calculated) doses in DP West 

plutonium areas by a factor of 2 for fields unfiltered by glass or steel and by 3 for fields that were 

filtered by these materials [LASL 1969 (3/5/63 memorandum)]. 

Cycolac Film Badge: The Cycolac badge was calibrated to the 100-g piece of uncoated plutonium 

metal encased in PVC and to a fluorescent X-ray source that emitted 71% 20-keV, 22% 60-keV, 5% 

100-keV, and 2% greater-than-200-keV X-rays, which simulated the radiations from plutonium at DP 

West [LASL 1969 (3/5/63 memorandum)]. 

For mixtures of higher and lower energy photons, the net optical density under a multielement filter 

was used with a 60Co calibration curve to estimate penetrating photon dose. The nonpenetrating dose 

was calculated from the measured net optical density under the unfiltered window based on the 60Co 

response, the penetrating dose was subtracted, and the remainder was multiplied by 0.07. This factor 

includes a correction for unfiltered film sensitivity (0.2) and conversion from roentgen to rem (0.35). 

The factor of 0.35 [which corresponded to the 1 rem = 0.5 R factor used with the brass-cadmium 

badge before January 1963 but was based on “better data” published by Watson (1959)] accounted 

for the fraction penetrating to the gonads based on Hanford Site research per a November 27, 1963, 

memorandum (LASL 1969). The DOELAP DCF for 17-keV photons is 0.38 rem/R (DOE 1986).

Around July 9, 1971, LANL started issuing film badges to selected individuals at DP Site that included 

an insert containing TLD ribbons [LASL 1977 (7/9/71 memorandum)]. The TLDs were calibrated on 

the P-6 gamma range [4] and with a plutonium fluoride neutron source. The purpose of these badges 

with TLD-600 and TLD-700 ribbons was to evaluate individual exposures between film development 

and evaluation. 

Cycolac Film Badges and TLDs: As of 1977, LANL photon dosimeters were calibrated with three 

types of sources: (1) K fluorescent X-rays from 10 to 100 keV, (2) heavily filtered X-ray beams from 

100 to 250 keV, and (3) gamma rays from isotopes above 250 to 1,000 keV (Storm, Cortez, and 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 33 of 79

Littlejohn 1977). The K fluorescent X-rays were produced using a 300-kV constant potential X-ray 

unit to produce 10- to 100-keV X-rays emitted from the primary tungsten target that caused K-shell 

X-rays to be emitted from secondary targets that varied in atomic number from 29 to 92. The heavily 

filtered X-ray beams were obtained with the primary X-ray beam by varying the potential applied to the 

tube and using large amounts of tin filtration to obtain relatively narrow-spectrum X-rays. The gamma 

rays were primarily the 412-keV, 662-keV, and 1,170- and 1,330-keV photons from 198Au, 137Cs, and 

60Co, respectively. 

For beta dose, personnel dosimeters were calibrated by exposure either in air to a high-dose-rate 

90Sr(90Y) source or in contact with a low-dose-rate uranium source. Both emit beta rays with 

maximum energies of about 2.3 MeV. A comparison of dosimeter response with the two sources was 

conducted. In both cases, TLDs were given a total exposure of 100 mrad based on the dose rate 

from the strontium-yttrium source as measured by the ion chamber and the dose rate from the 

uranium measured by the extrapolation chamber. The TLDs were calibrated to gamma rays from a 

60Co source. TLDs mounted in Cycolac plastic yielded readings of 64 mrad when exposed in air to 

strontium-yttrium and 51 mrad in contact with uranium. 

By 1977, calibrations were performed with dosimeters in air, on a phantom, and in a phantom (Storm, 

Cortez, and Littlejohn 1977). A free-air ionization chamber was the primary standard used in the 

measurement of photon radiation. Thimble-sized ionization chambers calibrated to the free-air 

chamber served as secondary standards. Electron radiation was measured with an end-window 

ionization chamber with a 7 mg/cm2 Kodapak wall. The dosimeters were calibrated to determine 

penetrating doses by placement of the secondary chamber 1 cm deep in a phantom and the 

personnel dosimeter on the surface, with a filter over the TLD to simulate 1-cm depth (Storm, Cortez, 

and Littlejohn 1977). Nonpenetrating dose calibrations were measured by placement of the chamber 

and a “lightly filtered” dosimeter on the surface of the phantom. 

Around 1977, the energy response of dosimeters to electrons was measured with beta-emitting 

isotopes that varied in maximum energy from 770 to 2,300 keV – 204Tl, 32P, and 90Sr(90Y) (Storm, 

Cortez, and Littlejohn 1977). The responses of the various dosimeters are listed in Table 6-8.

Table 6-8. TLD and film badge energy response to beta radiation, about 1977.

Dosimeter type

Maximum beta energy

0.77 MeV 1.16 MeV 2.27 MeV

Unfiltered TLD 0.43 0.53 0.64

TLD badge in 60 mg/cm2 Cycolac plastic 0.07 0.14 0.42

Film badge (unfiltered) 0.06 0.18 0.59

In December 1984, LANL applied to participate in the DOELAP pilot program. LANL did not seek 

accreditation for category VB (beta, uranium) or for the low-energy component of category VA (beta). 

While local testing indicated that routine procedures would adequately evaluate beta doses from 

contact with a sheet of uranium, LANL believed that the calibration method was not consistent with 

the geometry under which most beta doses were received. “At Los Alamos we are aware of no 

persons working with and receiving beta exposures from 204Tl” [LANL 1986 (12/14/84 memorandum)].

The LANL Model 7776 TLD badges in the pilot DOELAP test exhibited a positive bias, reportedly due 

to the results of backscatter from the Lucite calibration phantoms used in the DOELAP irradiations 

[LANL 1989 (2/14/86 memorandum)]. The phantom used by DOELAP personnel was reportedly 

much thicker than the one Storm used in developing the TLD calibration procedure, and the X-ray 

spectra used by DOELAP and Storm reportedly differed, even though both spectra had the same 

average energies (Widner 2003). The interplay of these two factors probably accounts for the 

“positive bias” in the LANL TLD badge response. LANL and all other participants in the pilot DOELAP 

test used the test results to modify their dose evaluation procedures to become compliant with 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 34 of 79

DOELAP testing procedures (Widner 2003). Backscatter increases the badge response by about 

10% for 137Cs gammas and more than 10% in the photon energy range from 40 to 150 keV. Thus, for 

TLD badges calibrated in free air, the exposure evaluation for badges exposed on a Lucite phantom 

(i.e., according to DOELAP procedures) is high by at least 10%. Because LANL lacked the time and 

personnel resources to reevaluate the response of the LANL TLD badge over the full range of photon 

and beta energies and the total fading response, the interim correction in 1986 was to remove the 

10% fading correction for photons and beta dose evaluations because the two effects compensate 

[LANL 1989 (2/14/86 memorandum)]. The 10% fading correction was retained for neutron exposure 

evaluation because the neutron calibration factors were established using a Lucite phantom. These 

changes were reportedly made operational for the January 1986 TLD badges.

Model 8823 TLD Badge: The general algorithm adopted for the Model 8823 was designed to 

determine the responses for each of the eight elements in the Model 8823 dosimeter through the use 

of a full set of DOELAP irradiation categories. At least 10 dosimeters were irradiated to each of the 

DOELAP techniques listed in Table 6-9 at Battelle, Pacific Northwest National Laboratory (PNNL) in 

late 1996. These irradiations included photons with effective energies ranging from 17 to 662 keV, 

betas with maximum energies of 760 and 2.27 MeV, and bare and moderated fission neutron sources. 

Table 6-9. DOELAP irradiation techniques 

and effective energies.

DOELAP Irradiation 

Categories Energy

K17 17 keV

M30 20 keV

S60 36 keV

K59 59 keV

Am-241 59 keV

M150 70 keV

H150 120 keV

Cs-137 662 keV

Betas

Tl-204 760 keV

Sr/Y-90 2.27 MeV

The Model 8823 dosimeter is DOELAP-accredited in all applicable categories. The general beta 

category (VA), which includes both 90Sr/90Y and 204Tl, is selected over the special contact geometry 

uranium category (VB) and special beta category (VC), which use only 90Sr/90Y or 204Tl. 

6.2.3.2 Neutron Dosimeters

Historical aspects of the calibration of LANL film, NTA, and thermoluminescent dosimeters are 

described in the Photodosimetry Evaluation Book (LASL 1959, 1969, 1977, 1979; LANL 1986, 1989, 

1996, 2001, 2003), technical basis documents, and LANL technical reports. 

Film Badges: Thermal-neutron calibration data were obtained from film badges exposed in the south 

thermal column of the LANL homogeneous reactor (Kalil 1955; Littlejohn 1961). Thermal neutron 

fluxes for a given reactor level were known to within ±10%, and relative values were known even 

better (Kalil 1955). Gold foils were used to measure the thermal neutron flux, and a nomogram was 

employed in the evaluation of film exposed to thermal neutrons (Littlejohn 1961). The nomogram 

indicated thermal neutron exposure (rem) as a function of normalized cadmium minus brass film 

optical density. It also indicated gamma exposure and “open-window equivalent gamma exposure 

contributed by thermal neutrons and gamma rays,” a parameter that was used in the evaluation of 

beta exposures (the difference between this parameter and the open-window radium-equivalent 

gamma exposure was taken as an estimate of the beta exposure in rad) (Littlejohn 1961). 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 35 of 79

Nuclear Track Emulsion: The Kodak Type B film used for measuring fast neutron exposures as of 

August 1960 (Littlejohn 1961) was developed by J. S. Cheka of Clinton Laboratories (Cheka 1954). 

LANL calibrated this film to neutron energies of 0.5, 0.6, 0.7, 0.8, 1, 1.5, 2.5, 4, 5, 8, 14, 17, and 20 

MeV (Littlejohn 1961). The 2.5- and 14-MeV neutrons were from a Cockcroft-Walton accelerator, and 

the others were from two Van de Graaff accelerators that the Laboratory had at the time. Data from 

these calibrations indicated that between the energies of 1 and 20 MeV the film was energydependent within +100% or –50% of the stated exposure, and each proton recoil track that appeared 

in a 1-mm2 portion of the cadmium-filtered area was assigned an average value of 0.008 rem 

(Littlejohn 1961). Fundamentally, the NTA dosimeter is capable of an accurate dose estimate only for 

neutron radiation greater than about 1 MeV because it has a lower energy threshold of about 

700 keV.

As of October 16, 1987, NTA badges were calibrated with a 238PuBe neutron dose of 500 mrem. A 

procedure was written for the processing of Kodak NTA film. Because TLDs are relatively insensitive 

to neutrons above about 3 MeV, the capability to process NTA films was retained. TLD badges with 

NTA film in piggyback holders were issued to Group HSE-11 at LAMPF. The combination badges 

were used with other detectors during neutron spectrum measurements in the ER-1 at LAMPF on 

July 22, 1987. 

A 1990 report stated that NTA film was frequently calibrated at LANL using a 238PuBe source or a 

bare 252Cf source with an average neutron energy of 4.5 MeV or 2.3 MeV, respectively (Mallett et al. 

1990). The film was irradiated in the Cycolac plastic holders through the use of an NBS slab phantom 

backing (40 by 40 by 15 cm methylmethacrylate slab). Based on neutron energy spectrum 

measurements performed at LAMPF in 1987 and the sensitive energy range of NTA film, it was 

concluded that the NTA dosimeter primarily measured exposure to neutrons in this facility from 

approximately 10 to 60 MeV; a response factor of 4 mrem/track/mm2 was conservatively chosen 

(Mallett et al. 1990). 

Earliest TLD Neutron Measurements: Around July 9, 1971, when LANL started issuing film badges 

at DP Site that contained TLD inserts, the TLDs were calibrated on the P-6 gamma range and with a 

PuF4 neutron source [LASL 1977 (7/9/1971 memorandum)].

Model 7776 TLD Badge: A method for calibrating the albedo-neutron dosimeter was developed that 

did not require a detailed knowledge of the neutron energy spectrum (Hankins 1973). A BF3 

proportional counter, centered in separate measurements in 3-in. and 9-in. polyethylene spheres 

containing thin shells of cadmium, was used to measure neutrons from a variety of sources. The ratio 

of the 9- to 3-in.-sphere counting rates was plotted against sensitivity (the ratio of the TLD-600 

response less the TLD-700 response in units of 60Co milliroentgen to the neutron millirem as 

measured on the 9-in. sphere). This yielded a linear relationship on a log-log plot. Under this 

method, the ratio of 9- to 3-in.-sphere count rates was measured in each potential neutron exposure 

area. A neutron calibration factor (the inverse of the sensitivity factor) was obtained from a plot of the 

9- to-3-in. count ratio to sensitivity generated for the albedo-neutron badge. The use of the 9- to-3-in. 

ratio technique allowed determination of an “average” neutron energy for a neutron source and an 

appropriate neutron calibration factor. In addition to using 9- to 3-in. sphere ratios for determination of 

NCFs, LANL compared responses of TLDs mounted on a phantom to responses from an adjacent 

9-in., neutron rem detector-moderated, BF3 tube-based, neutron survey instrument. 

Model 8823 TLD Badge: The general algorithm technique adopted for the Model 8823 was designed 

to determine the responses for each of the eight elements in the Model 8823 dosimeter through the 

use of a full set of DOELAP irradiation categories. A minimum of 10 dosimeters was irradiated to 

each DOELAP category at PNNL in late 1996. These irradiations included bare and D2O-moderated 

252Cf fission neutron sources in 1997 (Hoffman and Mallett 1999a,b). 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 36 of 79

Model 8823 and Track-Etch Dosimeter: A March 9, 2000, report (LANL 2001) described the results 

of irradiation of LANL TLD and TED badges to monoenergetic, accelerator-produced neutrons from 

0.144 to 19 MeV at Physikalisch-Technische Bundesanstalt (PTB), the national institute of natural and 

engineering sciences and technical authority for metrology and physical safety engineering of the 

Federal Republic of Germany. Neither the TLD nor the TED individually performed satisfactorily over 

the entire range of monoenergetic neutron energies. The TLD significantly under-responded to 

neutrons of 1.2 MeV and above; the monoenergetic PTB neutron fields were substantially different 

from bare and moderated fission sources for which the dosimeter algorithm was calibrated. The 

LANL TED (sometimes called the PN3) significantly under-responded below 1.2 MeV. By summing 

the results of the two dosimeter types, which was the practice at the time, these limitations are 

minimized. The combination results were within 40% of the delivered dose at 565 keV and above. 

As of early 1998, LANL TEDs were calibrated with a bare fission source [LANL 2001 (3/30/98 

memorandum)], apparently a 242Cf fission source with no scattering material between the source and 

the dosimeter (Widner 2004).

Figure 6-3 shows a plot of the geometric mean of the NP ratio for each year from 1979 to 2008 based 

on LANL records that contain both deep dose and neutron doses of 50 mrem or greater. 

This is the period after the use of NTA film was phased out at LANL. Along the top of the graph, the 

dosimeters in use over time are identified.

6.2.3.3 Workplace Beta/Photon Dosimeter Response

The energy response of Eastman Kodak Type K film for an exposure of 0.1 R is shown in Figure 6-4.

In a March 5, 1963, memorandum, the primary radiation fields “in general” at DP West were described 

as listed in Table 6-10 (LASL 1969). More detailed photon energy breakdowns from the same report 

are listed in Table 6-11.

Based on film badge data from uranium processing areas from early 1953, a factor of 0.1 was 

selected to determine gamma exposure from measurements of gamma-plus-beta exposure in 

preparing exposure records for computer entry [LASL 1959 (4/3/57 memorandum)]. The theoretical 

value for normal uranium was said to be about 1/17 gamma to beta-plus-gamma. The factor of 0.1 

indicated that beta doses were about 9 times gamma doses in those uranium processing areas.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 37 of 79

Figure 6-3. Geometric mean of neutron-to-photon ratios for records with both 

deep dose and neutron dose of 50 mrem or greater, 1979 to 2008 (LANL 2004, 

updated January 2009). 

Figure 6-4. Energy dependence of Eastman Kodak Type K film for exposure of 

0.1 R (Storm 1951).

Table 6-10. General energy distribution of photons in 

1962 at DP West Site.

Percentage of dose Photon energy

65 to 70 L X-rays (~17 keV and 26 keV)

10 to 20 60 keV

1 to 10 100 keV

0 to 7 Greater than 200 keV

0.00

0.50

1.00

1.50

2.00

2.50

3.00

3.50

1979 1984 1989 1994 1999 2004 2009

Year

Neutron/Photon Ratio.

Model 7776 TLD Model 8823 TLD
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 38 of 79

Table 6-11. Estimated photon spectra for several plutonium sources in DP West Site.

Energy 

(keV)

Pu in 

glovebox

Pu with 

20% Pu-240 

in glovebox

Pu with 

20% Pu-240, 

through glass

Pu with 

20% Pu-240, 

through steel

Pu/Am electrorefining 

residues in milk 

carton and plastic

20 72 68 8.9 0 75

60 22 13 35 9.3 24

100 4.6 13 38 52 1.1

>200 2.1 6.4 19 38 0

That 1963 memorandum described a study in which the brass-cadmium badge and the Cycolac 

badge were simultaneously exposed to three different plutonium sources at DP West that had known 

or calculable spectra (LASL 1969). With knowledge of the spectra and film sensitivities, expected 

results from each badge were calculated. The brass-cadmium badge underestimated the dose at DP 

West by a factor of 2 when the radiation field was not filtered by glass or steel and a factor of 3 in the 

filtered cases. The Cycolac badge measured the dose within ±10% when the source was unfiltered. 

When the source was filtered by glass or steel, the hard component was measured correctly by the 

Cycolac badge but the soft component was overestimated, which caused the total dose to be in error 

(high) by as much as 60%.

Errors that were expected with the use of the Cycolac badge with low-energy photons (<45 keV) when 

a correction factor of 0.07 was used were as listed in Table 6-12, based on the 1962 study at DP 

West [LASL 1969 (3/5/63 memorandum)]. Guidance in that memorandum was that “when significant 

exposures to soft radiation (<45 keV) are measured or anticipated, a special evaluation will be made.”

Table 6-12. Calculated errors from use of the Cycolac badge with lowenergy photons and a correction factor of 0.07.

Energy 

(keV)

Dose (R) from lower 

energy photons

Reported 

dose (rem)

True dose 

(calculated)

Percent 

difference

10 0.3 0.02 0.01 50% high

20 2 0.14 0.50 360% low

30 10 0.7 0.75 7% low

A 1972 study at DP Site compared results from film badges and TLDs over a 6-month period [LANL 

1996 (2/15/96 memorandum); Fix 2012]. The study used data from 38 plutonium workers in these 

areas: 239Pu recovery, 239Pu areas only, 238Pu areas, and the PuF4 area. While the lowest dose was 

294 mrem, most were around 1,000 mrem. Relevant results of the study, averaged over 6 months, 

are listed in Table 6-13. The film badge readings were about a factor of 3 higher than the TLDs. In 

his 1996 memorandum discussing the 1973 study [LANL 1996 (2/15/96 memorandum)], Hankins said 

the method used to calibrate for plutonium did not consider the effect of the glovebox or buildup of 

241Am in the gloveboxes. 

In a study documented in 1975, D. E. Hankins showed that LiF TLDs exposed on a phantom (or a 

person) had an over-response to gamma ray energies around the 60-keV characteristic of 241Am

(Hankins 1975).

Table 6-13. Results of a 6-month 

comparison of film badges, NTA film, and 

TLDs at DP Site.

Area of DP site

Ratio of film dose to 

TLD gamma dose

Pu-239 recovery area 3.4 (range 2.4 to 4.2)

Pu-239 areas 1.9 (range 0.45 to 2.7)

PuF4 areas 3.0 (range 2.7 to 3.4)

Pu-238 areas 1.5 (range 10 to 2.2)
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 39 of 79

6.2.3.4 Workplace Radiation Fields

Table 6-14 lists LANL beta and photon energies and percentages. Table A-3 summarizes the 

recommended uncertainties to apply to the reported photon doses.

Table 6-14. Selection of beta and photon radiation energies and percentages.

Process/

Buildings Description

Operations

Radiation

type

Energy 

selection 

Begin End (keV) Percentage

Reactors 

and reactor 

fuels

During operation: Dispersed fields of higher energy photons from 

fission, activation, and fission products. Narrow beams of higher energy 

neutrons from test ports, etc., into reactor core. Potential for significant 

airborne nuclides and significant higher energy beta radiation. 

Beta

Photon

>15

30–250

>250

100

25

75

Not in operation: Dispersed fields of higher energy photon radiation 

fields from activation and fission products. Insignificant neutrons. 

Possible higher energy beta radiation during maintenance work due to 

fission products.

LOPO Water Boiler (TA-2) May 1944 Nov 1944

HYPO Water Boiler (TA-2) Dec 1944 Feb 1951

SUPO Water Boiler (TA-2) Mar 1951 Jun 1974

Plutonium Fast Reactor (Clementine, TA-2) Dec 1946 Dec 1950

Omega West Reactor (TA-2) Jul 1956 Dec 1992

LAPRE I (TA-35) Feb 1956 Oct 1956

LAPRE II (TA-35) Feb 1959 May 1959

LAMPRE I (TA-35) Early 1961 Mid-1963

UHTREX (TA-52) Dec 1956 Feb 1970

Irradiated fuel work (including TA-21, TA-35, TA-48 in the 1960s and 

CMR Wing 9 in the 1960s through about 1986).

Uranium 

production

Processing and machining: Depleted and enriched uranium. Beta

Photon

>15

30–250

100

100 TA-1 (Sigma, HT Buildings) 1943 1953

TA-3 (Sigma Complex, CMR Building) 1953 Present

Plutonium 

processing

Radiochemical operations: Plutonium processed at LANL had largely 

been separated from fission products. Radiochemical operations were 

largely for recovery of fissionable material. Beta

Photon

>15

<30

30–250

100

65

35

TA-1, D Building 1943 1945

TA-21, DP West Site Nov 1945 1978

TA-55, Plutonium Facility 1978 Present

TA-3 (including CMR Building) 1953 Present

Plutonium 

production 

Plutonium component production: Pu is machined into components 

using glovebox assembly process with mostly close anterior exposures. 

Radiation fields involve significant lower energy photons and neutron 

radiation.

Photon <30

30–250

65

35

TA-1 (D Building), TA-21 (DP Site), TA-55 (PF Site)

Plutonium storage: Radiation characteristics in this area generally 

involve dispersed lower energy neutron radiation and scattered photons, 

including 60-keV Am-241 gamma ray.

TA-1, D-5 Sigma Vault 1943 1945

TA-21, Building 21 Nov 1945 1978

TA-55 Vault 1978 Present

Accelerator 

operations

LAMPF operations at TA-53: Primarily from 

residual activity induced in targets, accelerator 

structures and components, grease, oils, and 

soil.

LANSCE

1972 Present Non-Penetrating

Beta

Photons

>15

> 30 keV 

photons

50

50

Penetrating

Photon

<30

30–50

>250

1

9

90

Calibrations

LANL site calibration of instruments and 

dosimeters

1943 Present Beta

Photon

>15

30–250

>250

100

25

75

Waste Radiation characteristics highly dependent on source of waste. Beta >15 100
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 40 of 79

Process/

Buildings Description

Operations

Radiation

type

Energy 

selection 

Begin End (keV) Percentage

handling Liquid waste: TA-45 and TA-50;

Solid waste: Areas A, B, C, D, E, G, T, U, V 

Various, 

1944 on

Photon 30–250

>250

50

50

Radioactive 

lanthanum 

operations

Preparation and use of radioactive lanthanum sources: Beta 

Photon

>15

30–250

>250

100

10

90

TA-10 (Bayo Canyon) 1944 1950

TA-35, “Ten Site” 1951 1963

Sources: LASL (1959, 1969, 1977, 1979); LANL (1986, 1989, 1996, 2001, 2003).

6.2.3.5 Adjustments to Measured Dose Shallow Dose Components

Early LANL film badges had no unfiltered elements, and it has been determined that beta doses may 

only be estimated from 1949 onward. Shallow dose from photons in plutonium processing and 

production areas would have been unmonitored by the early dosimeters as well. For these cases, in 

the years prior to 1949, the recorded deep photon doses should be assigned in accordance with 

Table 6-14, and shallow dose should be assigned as low-energy photons (<30 keV), at 1.86 times the 

recorded photon dose. 

6.2.3.6 Workplace Neutron Dosimeter Response

The AEC held a series of Personnel Neutron Dosimetry Workshops to address problems experienced 

by its sites concerning accurate measurement of neutron dose. The first workshop was held 

September 23 and 24, 1969 (Vallario, Hankins, and Unruh 1969), with the stated concern: “... for 

intermediate energy (i.e., > 0.4 eV to < 700 keV) ... neutron sources, NTA personnel neutron 

dosimeters cannot be effectively used. This leaves a gap in the personnel dosimetry program which 

at many installations may be quite serious.” The significance of the underestimated neutron dose 

became evident with studies conducted to implement TLDs. At LANL, studies of that type were 

conducted in the early 1970s [LANL 1996 (2/25/96 memorandum)].

In 1994, the neutron component of the collective person-rem for LANL was 75% of the total [LANL 

2001 (2/22/96 memorandum)].

The main work areas at LANL where there has been a potential for neutron exposure include: 

• D Building (TA-1)

• DP West (TA-21)

• DP East (TA-21)

• Current Plutonium Facility (TA-55)

• Omega Site (TA-2)

• LAMPF (TA-53)

• Criticality Laboratory (TA-2, TA-18)

• CMR Building (TA-3)

• Reactor Operations (TA-35 and TA-52)

In 1989, the majority of workers who received neutron exposures worked at LAMPF (TA-53) and the 

Plutonium Facility (TA-55) [LANL 1996 (4/18/90 memorandum)].

Based on measurements with the Model 8823 dosimeter reported in a January 23, 2003, 

memorandum (LANL 2003):

• The vast majority (more than 90%) of LANL employees, including those at TA-55, received 

WB neutron doses from “well moderated fields, akin to ~2.5-4” (or greater) polymoderated 

252Cf (fission spectrum),” and 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 41 of 79

• Less than 2% of LANL positive neutron exposures were “in fields similar to bare Cf-252.” 

Based on neutron spectrum measurements performed by LANL personnel (Harvey and Hajnal 1993) 

and information from the Photodosimetry Evaluation Book [LANL 1996 (1/17/95 memorandum)], 

Table 6-15 lists the approximate correspondence between NCFs used by LANL and the dose fraction 

for the four bins of neutron energy used by NIOSH (2007a). The breakdowns by energy category are 

estimated based on the “H(%)” columns from the Harvey and Hajnal (1993) report. A lower NCF 

value indicates that the distribution of neutron energies has shifted to lower values; that is, it has 

moderated. As neutrons moderate, the dosimeter responds more per neutron, but the DE per neutron 

is lower.

Table 6-15. Approximate NCFs and dose fractions for neutron sources.

Type of source NCF

Dose fraction by energy category

<10 keV 10–100 keV 0.1–2 MeV 2–20 MeV

Bare Pu-239 ~1.0

Bare PuBe ~1.5 1 1 33 65

Bare Cf-252 ~1.3 0 0 42 58

Cf-252 through 10.2-cm Lucite ~0.15 5 1 33 61

As described above, irradiation of LANL TLD and TED badges to monoenergetic, acceleratorproduced neutrons from 0.144 to 19 MeV at PTB in Germany (LANL 2001) showed that the TLD 

significantly under-responded to neutrons of 1.2 MeV and above, the TED significantly underresponded below 1.2 MeV and, when results were combined (as was the practice at LANL), results 

were within 40% of the delivered dose at 565 keV and above. 

6.2.3.6.1 Plutonium Processing Areas (TA-1, TA-21, TA-55)

6.2.3.6.1.1 Neutron Energy Spectrum

In a study released in 1978, 9-in. to 3-in. sphere ratio measurements at 13 locations at the TA-55 

Plutonium Facility yielded ratios that indicated an average neutron energy of approximately 200 keV 

(Blackstock et al. 1978). The mean value of the ratio was 0.57, with a standard deviation of 35%. 

This result was thought to possibly explain why few neutron exposures had been observed using NTA 

film at this facility, because NTA film cannot detect neutrons with energies less than about 700 keV.

Multisphere neutron spectroscopy methods were applied to measure representative working fields in 

the LANL Plutonium Facility in 1993 (Harvey and Hajnal 1993). Work in this facility has involved 239Pu 

and 238Pu. Figure 6-5 shows the neutron spectrum measured during the hydrofluorination of 239PuO2, 

Figure 6-6 shows the spectrum of the ball milling process with 238Pu, and Figure 6-7 shows the 

spectrum of the Special Nuclear Material (SNM) storage vault. 

6.2.3.6.1.2 Neutron-to-Photon Dose Ratio

The following data on NP ratios for LANL plutonium workers are available [LASL 1977 (11/9/72 

memorandum)]:

• For 239Pu workers not in fluoride areas, observed NP ratios ranged from 0.3 to 1.7 with an 

average of 0.7.

• For 239Pu-fluoride areas, an NP ratio of 2.8 was reported for a health physics technician, not a 

chemical operator.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 42 of 79

Figure 6-5. Neutron spectrum of 

hydrofluorination of 239PuO2 at Plutonium Facility

(Harvey and Hajnal 1993).

Figure 6-6. Neutron spectrum of 238Pu ball 

milling process at Plutonium Facility (Harvey 

and Hajnal 1993).

Figure 6-7. Neutron spectrum near Door K in 

SNM Vault at Plutonium Facility (Harvey and 

Hajnal 1993).

• For 238Pu workers, observed NP ratios ranged from 2.7 to 5.5 with an average of 3.9 (238Pu 

exposures began in 1969).
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 43 of 79

• For health physics technicians in 238Pu areas, observed NP ratios ranged from 1.9 to 4.6 with 

an average of 3.3.

The measured average quotient of neutron to gamma- and X-ray DE for the Plutonium Facility in 1993 

was 2.5 (Harvey and Hajnal 1993).

The 1972 study at DP Site mentioned above [LANL 1996 (2/15/96 memorandum); Fix 2012] 

compared neutron dose results from NTA and TLDs and yielded NP ratios that varied from 1.33 to 

2.30 for 239Pu areas and averaged 3.98 for 238Pu areas, as listed in Table 6-16.

Table 6-16. Results of 6-month comparison of film badges, NTA film, and TLDs at DP Site.

Area

Neutron component 

of total dose

Gamma-toneutron ratioCorresponding 

neutron-to-gamma ratio

Ratio of NTA dose 

to albedo TLD dose

Pu-239 recovery 39% to 80% (avg. 58%) 0.75 1.33 0.12

Pu-239 areas 50% to 84% (avg. 66%) 0.60 1.66 0.22

PuF4 areas 60% to 78% (avg. 71%) 4.34 2.30 0.19

Pu-238 areas 69% to 90% (avg. 80%) 0.25 3.98 0.48

6.2.3.6.2 LAMPF (TA-53)

In 1987, neutron energy spectrum measurements were made at a potentially high neutron energy 

area at LAMPF (Mundis and Howe 1987). The resulting neutron spectrum is shown in Figure 6-8; 

associated data have been interpreted as follows in the Mundis and Howe memorandum and in a 

1990 report on dosimetry at LAMPF (Mallett et al. 1990). When unfolding codes were applied to the 

measurement data, they revealed that more than 90% of the neutron DE was due to neutrons of 

energy greater than 1 MeV, and 70% was due to neutrons of energy greater than 10 MeV. The 1987 

measurements showed that the LANL Model 7776 TLD badge under-responded by a factor of 5 to 7 

for this particular neutron spectrum (Mallet et al. 1990) because the sensitivity of the albedo-type 

dosimeter falls off severely for neutrons with energies above a few MeV (Mundis and Howe 1987). 

NTA film also under-responded, but only by about 20%. The sum of the two dosimeters was reported 

to be in good agreement with the spectrum unfolding results.

Figure 6-8. Neutron spectrum in ER-1 Area of LAMPF with proton 

beam stopped in the carbon beam block, as determined from unfolding 

codes (Mundis and Howe 1987).

In another study, however, 9-in. to 3-in. sphere ratio measurements at 18 locations at LAMPF (not just 

an area of potentially high neutron energy) yielded ratios that indicated an average neutron energy of 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 44 of 79

<100 keV at LAMPF (Blackstock et al. 1978). The mean value of the ratio was 0.17, with a standard 

deviation of 30%. This result was thought to possibly explain why few neutron exposures had been 

observed with the use of NTA film at this facility, because NTA film cannot detect neutrons with 

energies less than about 700 keV.

6.2.3.6.3 Critical Assembly Testing (TA-18)

6.2.3.6.3.1 Neutron Energy Spectrum

Neutron spectral measurements were made in three areas at TA-18 in 1998 and 1999 to determine 

what NCFs should be used in conjunction with exposures from critical assembly testing [LANL 2001 

(3/17/98 and 6/24/99 memoranda)]. As a result of this work, an NCF of 0.07 was recommended for 

areas surrounding TA-18. A plot of neutron spectra from several sources, including environmental 

monitoring TLD Station 6 in the parking lot for TA-18, is shown in Figure 6-9 [LANL 2001 (6/24/99 

memorandum)].

Figure 6-9. Neutron spectrum at Station 6 near TA-18 (white line).

A 1967 study of selected criticality dosimetry methods yielded neutron energy distributions for the key 

LANL critical assemblies (Hankins 1968). Table 6-17 lists the breakdown of total kerma by energy 

interval at a distance of 3 m (or 5.9 m for Hydro) from five critical assemblies when they were active. 

Spectral data taken at these distances will generally not be useful in the assessment of external 

doses to workers who were involved with criticality experimentation other than in the special cases of 

accidental exposures. Workers are normally at remote locations while the criticality experiments are 

in progress. 

6.2.3.6.3.2 Gamma-to-Neutron Dose Ratio

The 1967 study of criticality dosimetry methods yielded estimates of gamma-to-neutron ratios for five 

critical assemblies at LANL based on measurements with TLDs and film badges placed in air, on the 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 45 of 79

Table 6-17. Percent of total kerma by energy interval from critical assemblies during experiments

(Hankins 1968).

Critical assembly

Distance from 

assembly (m)

Neutron energy

0.4–750 keV 0.75–1.5 MeV 1.5–2.9 MeV >2.9 MeV

Jezebel (outside kiva) 3.0 14 23 21 42

Jezebel (inside kiva) 3.0 9 31 22 39

Hydro 5.9 27 36 3.4 34

Flattop 3.0 41 43 11 11

Parka 3.0 45 22 23 23

Godiva IV (burst mode) 3.0 18 37 24 24

Godiva IV (extended mode) 3.0 13 41 25 25

front of “plastic man” manikins filled with sodium solution, and on the back of plastic men (Hankins 

1968). Table 6-18 lists the gamma-to-neutron and corresponding neutron-to-gamma ratios for the 

Hydro critical assembly based on measurements with TLDs on the front of plastic men at distances 

from 5.9 to 100 m. Hankins (1968) presents data for six distances from 5.9 to 19.8 m. A line was fit 

to these data for extrapolation of the ratios to greater distances.

Table 6-18. Gamma-to-neutron ratios measured with TLDs on the 

front of plastic man manikins at various distances from the Hydro 

critical assembly (Hankins 1968).

Distance from 

assembly (m)

Gamma-to-neutron 

ratioa

Corresponding 

neutron-to-gamma ratioa

5.9 1.5 0.67

6.0 1.0 1.0

8.7 1.2 0.83

12.3 1.1 0.91

16.2 1.1 0.91

19.8 0.84 1.2

25 0.83 1.2

35 0.71 1.4

50 0.58 1.7

70 0.46 2.2

90 0.39 2.6

100 0.36 2.8

a. Values beyond 20 meters are linearly extrapolated from data in Hankins (1968).

6.2.3.6.4 Reactor Areas (TA-2)

A March 4, 1982, document (in LANL 1986) describes a review of the NCFs in use for Omega West 

Reactor (CNC-5) personnel. Neutron measurements were made with a 9-in.-diameter polyethylene 

sphere, an RM-16 rate meter, and an Ortec scaler. TLD badges mounted on a 1.5-in. polyethylene 

slab were exposed beside the sphere. Measurements were made near the north face of the reactor, 

where significant personnel exposures were most likely to occur. Results indicated that a correction 

factor of 0.1 should be applied to the neutron dose rather than the 0.5 in current use. It is 

recommended that this change be made in the evaluation procedures. Data sheets attached to that 

March 4, 1982, document give penetrating radiation and neutron millirem values for five TLDs for 

each of the two runs.

Because neutron spectra have not been obtained for the LANL reactors, an assumption of 100% 

fission spectrum neutrons (0.1 to 1 MeV) is used.
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 46 of 79

6.2.3.6.5 CMR Building (TA-3)

Activities that involve plutonium at the CMR Building (also known as South Mesa Building 29, or 

SM-29) have included laboratory work on small quantities of uranium and plutonium (Widner et al. 

2004). In addition, Wing 9 of that building contains hot cells that have handled irradiated uranium and 

sometimes plutonium. 

Stack FE-19 of the CMR Building serves the glovebox processes and rooms on the south side of 

Wing 3. Since early 1974, FE-19 has been a major source of plutonium at LANL, up to 99% of the 

total released in 1980. Alpha-emitting radioactivity in liquids flowing into the TA-50 waste treatment 

plant rose sharply around 1973 because of increased use of 238Pu in the CMR Building (Widner et al. 

2004).

The neutron spectrum at the CMR Building is assumed to be similar to that of plutonium processing 

areas but with a slight increase in the fraction in the 2-to-20-MeV category due to the possibility that 

research activities involved sources that emitted more higher energy neutrons (such as PuBe or 

252Cf). The dose fraction in the 2-to-20-MeV category was raised from 33% to 40%, and dose 

fractions for the other categories evenly reduced accordingly.

6.2.3.7 Neutron Dose Fraction

The fraction of the total dose in each neutron energy group can be determined by dividing the neutron 

spectra into the four lower neutron energy groups discussed in NIOSH (2007a). The highest neutron 

energy group (>20 MeV) was not used because operations at LANL, other than in a particularly highenergy neutron area of LAMPF where worker exposures were probably uncommon, did not produce a 

significant component of neutrons of this energy. The dose for each neutron energy group was 

calculated by multiplying the neutron flux Ø (Roberson, Cummings, and Fix 1985; Brackenbush, 

Baumgartner, and Fix 1991) by the corresponding flux to DCFs in National Council on Radiation 

Protection and Measurements (NCRP) Report 38 (NCRP 1971). The neutron doses in each NCRP 

Report 38 energy interval are summed to develop the four neutron group doses. The dose fraction Df

for each neutron energy group n was calculated by dividing the neutron group dose by the total dose 

DT: 

(6-1)

where:

Ø(Ei) = neutron flux of the i-th energy bin

DCFi = NCRP Report 38 flux to dose-rate conversion factor for the i-th energy bin

DT = total dose

Table 6-19 lists the neutron dose fractions by energy group using data measured by Roberson, 

Cummings, and Fix (1985). 

Table 6-19. Laboratory-measured dose fractions from PuF4.

Neutron 

energy group

Shielding of PuF4 sourcea

0 cm (bare) 2.54 cm 5.08 cm

<10 keV 0.00 0.00 0.01

10–100 keV 0.00 0.00 0.00

0.1–2 MeV 0.06 0.85 0.89

2–20 MeV 0.94 0.15 0.10

T

i

i i

f n D

E DCF

D E

∑

=

( )

( )

φ
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 47 of 79

Default dose fractions

0.1–2 MeV 0.1 0.9 0.9

2–20 MeV 0.9 0.1 0.1

a. Thickness of acrylic shielding between source and detector.

6.2.3.8 Uncertainty in Neutron Dose

Measurement of neutron dose in the workplace is difficult (Brackenbush, Baumgartner, and Fix 1991). 

A significant under-response in recorded dose with NTA dosimeters became evident in the late 1960s 

at several sites that were preparing to implement TLD neutron dosimeters. Section 6.4 describes the 

neutron dose adjustments required, and Table A-3 summarizes the recommended uncertainties to 

apply to the recorded dose values.

6.3 ADJUSTMENTS TO RECORDED PHOTON DOSE

The following are instances for which corrections to recorded photon doses might be warranted:

• In a 1972 study with prototype albedo TLDs used alongside film badges in DP West plutonium 

areas, film badge readings were about a factor of 4 higher than TLD results for 239Pu recovery 

areas and a factor of 2 higher in other DP West plutonium areas [LASL 1977 (1/23/74 

memorandum)]. 

• In a 1963 study with the brass-cadmium film badge used alongside the Cycolac multielement 

film badge, the brass-cadmium badge underestimated dose at DP West by a factor of 2 when 

the radiation was not filtered by glass or steel and a factor of 3 when photons were filtered 

[LASL 1969 (3/5/63 memorandum); average factor of 2.5].

• From September 1961 to October 1964, a software problem resulted in treatment of 

exposures to photons greater than 200 keV in energy from the brass-cadmium badge as soft 

gamma exposures, which caused doses higher than 100 mrem to be reported low by as much 

as a factor of 4 [LASL 1969 (10/13/64 memorandum)]. 

• The Cycolac film badge used from 1962 to 1978 over-responded to photons with energies 

below 100 keV, which resulted in overestimates of the dose by as much as a factor of 2 in the 

Plutonium Facility (Storm et al. 1981). This finding agreed with the results of the Hankins 

6-month study from 1973 (Fix 2012). 

• From March 1963 to March 1973, penetrating photon doses (rem) from plutonium exposures 

were set equal to delivered exposures (roentgen) in error, which caused penetrating gamma 

doses to be reported high by about a factor of 2.9 [LASL 1977 (5/2/73 and 1/23/74 

memoranda; the factor of 2.9 corrects for inappropriate reduction in the nonpenetrating 

component to 35% of actual in calibration)]. 

The information above has influenced recommendation of a series of nonoverlapping correction 

factors (see Attachment A, Section A.3) for recommended application to photon doses reported by 

LANL. In several cases in which there is evidence that might support modification of reported doses, 

correction factors are not recommended because of the lack of incontrovertible evidence sufficient to 

justify adjustment of reported doses in manners that might not be favorable to claimants. The first, 

second, fourth, and fifth bullets above describe information that has not led to the recommendation of 

specific correction factors in Table A-2. 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 48 of 79

6.4 ADJUSTMENTS TO RECORDED NEUTRON DOSE

Adjustments to LANL recorded neutron doses are necessary to estimate doses because of the 

uncertainty associated with the recorded dose in the complex workplace radiation fields and the 

variability in exposure circumstances. 

6.4.1 Neutron Dose Adjustments

LANL incorporated the energy variation of the DE in its calibration methodology. As a result, the 

recorded DE (DER) is a combination of all neutron energies. To calculate the POC, the recorded 

neutron dose must be separated into neutron energy groups and later converted to International 

Commission on Radiological Protection (ICRP) Publication 60 methodology (ICRP 1991). 

6.4.2 Neutron Weighting Factor

Adjustment to the neutron dose is necessary to account for the change in neutron quality factors 

between historical and current scientific guidance, as described in NIOSH (2007a). LANL neutron 

calibration factors determined from National Institute of Standards and Technology (NIST)-calibrated 

sources are used directly without modification for field conditions. The quality factor is incorporated in 

the NIST calibration methodology, which used flux-to-dose-rate conversion factors for varying neutron 

energies for each calibration source. Flux-to-dose-rate conversion factors were based on NCRP 

Report 38 (NCRP 1971), which lists flux-to-dose-rate conversion factors and associated quality 

factors that vary from 2 at energies less than 1 keV to 11 at 1 MeV. To convert from NCRP Report 38 

quality factors to ICRP Publication 60 radiation weighting factors (ICRP 1991), a curve was fit that 

described the neutron quality factors as a function of neutron energy. The average quality factor for 

each neutron energy group was developed by integrating the area under the curve and dividing by the 

neutron energy range as shown in Equation 6-2: 

(6-2)

Table 6-20 summarizes changes in the quality factors and the average NCRP Report 38 quality factor 

for the neutron energy groups used in dose reconstruction (NCRP 1971). 

Table 6-20. Neutron quality or weighting factors

Neutron energy 

(MeV)

Historical 

dosimetry 

guidelinea

NCRP 38

quality factorsb

Average 

quality factor 

used at LANL

ICRP 60 neutron 

weighting factor 

(wR)

c

2.5E-8 3 2

2.35 5

1E-7

10

2

1E-6 2

1E-5 2

1E-4 2

1E-3 2

1E-2 2.5 5.38 10

1E-1 7.5

5E-1 11 10.49 20

1 11

2 10

7.56 10 2.5 9

5 8

7 7

(2.0 0.1)

( )

( )

2.0

0.1

,0.1 2.0 − =

∫ − Range

Q E dE

Q E

f

n MeV
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 49 of 79

Neutron energy 

(MeV)

Historical 

dosimetry 

guidelinea

NCRP 38

quality factorsb

Average 

quality factor 

used at LANL

ICRP 60 neutron 

weighting factor 

(wR)

c

10 6.5

14 7.5

20 8

40 7 Not applicable 5

60 5.5

a. Trilateral meeting in 1949 radiation protection guidelines (Fix, Wilson, and Baumgartner 1997). 

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

c. ICRP Publication 60 (ICRP 1991).

6.4.3 Neutron Correction Factor

Table 6-21 lists the average quality factor for the four neutron energy groups that encompass LANL 

neutron exposures. The neutron DE correction factor can be calculated by dividing the dose fractions 

for each neutron energy group Df(En) by the corresponding energy-specific average NCRP Report 38 

quality factor Q(En) and then multiplying by the ICRP Publication 60 radiation weighting factor (wR), as 

shown in Equation 6-3 (NCRP 1971; ICRP 1991). 

(6-3)

Table 6-21. Facility dose fractions and associated ICRP Publication 60 correction factors (ICRP 

1991; LASL 1959, 1969, 1977, 1979; LANL 1986, 1989, 1996, 2001, 2003) 

Process Description/buildings

Operations

Neutron 

energy

Default dose 

fraction (%)

ICRP 60 

correction 

Begin End factor

Plutonium 

production

Neutron dose was associated with overall LANL plutonium production process in which plutonium was 

purified, formed, machined, and recovered. Work was primarily conducted in gloveboxes with 

predominant close anterior exposure to workers. Neutron exposure also resulted from spontaneous 

fission neutrons from Cm-242 and Cm-244. 

Plutonium facilities

(D Building at TA-1, DP 

West Site at TA-21, 

Plutonium Facility at TA55)

1943 Present <10–100 keV

0.1–2 MeV

2–20 MeV

11

56

33

0.23

1.1

0.44

LAMPF, 

LANSCE

Neutrons of varying energies from high-energy proton linear accelerator studies.

LAMPF at TA-53 1972 Present <10 keV

10–100 keV

0.1–-2 MeV

2–20 MeV

30

30

20

20

0.64

0.56

0.38

0.26

Reactor 

operations

Neutrons of varying energies from reactor operations.

Omega Site (TA-2) 

TA-35 

TA-52 

1944

1955

1969

1992

1963

1970

0.1–2 MeV 100 1.9

Criticality 

experiments

Neutrons of varying energies from criticality testing and experimentation (in normally occupied areas, not 

including accidental exposures, which must be considered as special cases).

Omega Site (TA-2) 

TA-18 

1943

April 1946

April 1946

Present

<10–100 keV

0.1–2 MeV

2–20 MeV

3.2

59

38

0.060

1.1

0.50

Chemistry 

and

metallurgy 

research

Neutrons of varying energies from actinide chemistry and metallurgy research.

TA-1

TA-3

1943

1952

1952

Present

<10–100 keV

0.1–2 MeV

2–20 MeV

10

50

40

0.21

0.95

0.53

Table 6-21 summarizes default neutron dose fractions by energy for LANL work areas where field 

measurements of neutron spectra were performed through the use of the associated ICRP Publication 

R

n

f n

f n w

Q E

D E C E = ×

( )

( ) ( )
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 50 of 79

60 correction factors (ICRP 1991). For years after 1978, the neutron DE is calculated by multiplying 

the recorded neutron dose by the area-specific correction factors. For example, consider a 

1,000-mrem recorded neutron dose to a worker at DP West; the corrected neutron dose is calculated 

as follows:

• 1,000 × 0.44 = 440 mrem from neutrons of 2 to 20 MeV estimated to represent 33% of the 

dose fraction 

• 1,000 × 1.1 = 1,100 mrem from neutrons of 0.1 to 2 MeV estimated to represent 56% of the 

dose fraction 

• 1,000 × 0.23 = 230 mrem from neutrons of <10 keV to 100 keV estimated to represent 11% of 

the dose fraction 

Thus, the corrected neutron dose is a total of 1,770 mrem. These adjustments should be applied to 

measured dose, missed dose, and dose determined based on an NP ratio. For years before 1979 at 

LANL, multiply the annual photon dose (adjusted for missed dose) by the NP dose factor from 

Section 6.4.4 and by the area-specific ICRP Publication 60 correction factor listed above to estimate 

neutron dose (ICRP 1991). For the years 1976-1980, NIOSH has agreed, in addressing comments 

on the TBD, to assign neutron dose using the 95th-percentile value of the NP ratio to address 

uncertainties associated with the transition from film to TLD dosimeters. 

6.4.4 Neutron-to-Gamma Dose Factors

Essentially all LANL radiological work areas with significant neutron radiation also had significant 

photon radiation. For periods and workplace settings for which neutron dosimetry methods are 

thought to have been particularly unreliable or uncertain (for example, NTA film dosimetry for 

intermediate-energy neutron sources), it is sometimes possible and advisable to estimate neutron 

doses based on measured gamma doses through application of NP ratios. This is because of the 

documented underestimating of doses by NTA dosimeters due to fading and significant effects of the 

NTA energy threshold in typical workplace neutron spectra. DOE neutron dosimetry workshops 

stated strong concerns for the performance of the NTA dosimeters in typical plutonium facility neutron 

fields. A recent National Radiological Protection Board analysis of Monte Carlo N-Particle (MCNP) 

Transport Code calculations of NTA response in workplace neutron fields (NRPB 2001) reached the 

conclusion that there are few if any realistic parameters (such as calibration to a lower energy neutron 

spectra than the workplace spectra) that would result in an overestimate of the neutron dose using 

NTA dosimeters, whereas there are numerous parameters that would result in a potential significant 

under-response.

Section 6.2.3.6.1.2 contains reported NP ratios for LANL plutonium workers. Reported neutron-togamma ratios range from 0.3 to 5.5. Section 6.2.3.6.3.2 contains reported NP ratios for LANL critical 

assembly areas. Reported neutron-to-gamma ratios for the Hydro assembly range from 1.7 to 2.8 for 

working distances thought to be representative of normal testing procedures (50 to 100 m from the 

assembly). Based on these site-specific data, estimated median and 95th-percentile values of 

neutron-to-gamma ratio are recommended for plutonium facilities and criticality experiments in the first 

through third rows of data in Table 6-22. Values for other operations, based on analysis of annual 

deep and neutron doses reported by LANL for 1979 to 2004, are listed in the fourth row of data in 

Table 6-22. These values are based on post-NTA results in which the deep and neutron doses were 

50 mrem or greater (LANL 2004). These values were reanalyzed for this revision of this TBD, and the 

results are included in Attachment B. The values were validated, though the 95th-percentile value 

was increased from 6.48 to 8.15. The calculated geometric standard deviations (GSDs) were 

included for all ratios in Table 6-22. 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 51 of 79

Table 6-22. Recommended distributions for neutron-to-gamma ratio.

Neutron source type

Neutron-to-photon dose ratio

Median GSD 95th percentile

Plutonium facilities (predominantly Pu-239) 0.7 2.3 2.8

Plutonium facilities (documented involvement with Pu-238 

operations; applies only to 1969 or later; most likely at TA-21, 

TA-3-29 [CMR Building], or TA-55)

3.9 1.23 5.5

Criticality experiments (>50 m distant) 2.3 1.05 2.5

Other operations 1.6 2.7 8.15

6.5 MISSED DOSE

There are undoubtedly missed recorded doses for LANL workers. The analysis has been separated 

according to photon and neutron missed dose. 

6.5.1 Photon Missed Dose

Missed photon dose for LANL workers would have occurred if doses received were below the limits of 

detection for the dosimeters provided and if they were based only on a recorded or assumed zero 

dosimeter result. Methods to be considered if there was a recorded or assumed zero dosimeter result 

for a period during a working career were examined by Watson et al. (1994). In general, estimates of 

missed dose can use dose results for coworkers or the recorded dose before and after the period of 

missed dose. However, these situations require careful examination. Missed dose for dosimeter 

results less than the MDL is particularly important for earlier years when MDLs were higher and 

dosimeter exchange was more frequent. 

NIOSH (2007a) describes options to calculate missed dose. The typical method is to assign dose 

equal to the MDL divided by 2 for each dosimetry result that is either zero or some positive value less 

than the MDL/2.

Analysis of missed photon dose by period according to dosimeter type and exchange frequency is 

needed to evaluate claim information, particularly if only annual dose data are available. The normally 

cited MDLs for beta and photon dosimeters are based on laboratory irradiations. Actual MDLs are 

higher because of additional uncertainty in field use and the use of dose recording thresholds. 

Table 6-23 summarizes potential missed photon dose. Reasonable MDLs are listed in this table for 

most applications for film dosimeters based on LANL documentation and reports from other DOE 

facilities (Wilson 1960, 1987; NIOSH 1993; NRC 1989; Wilson et al. 1990) and for TLDs (Fix et al. 

1997; Mallett, Hoffman, and Vasilik 1994; Hoffman and Mallett 1999a). 

Table 6-23. Photon dosimeter period of use, type, MDL, exchange frequency, and potential annual 

missed dose.

Period of usea Dosimeter

MDLa

(mrem)

Exchange 

frequency

Mean annual missed 

doseb (mrem)

February 1945–

September 1962c Brass, brass clip, or brass-cadmium film

40 Monthly (n = 12) 240

Biweekly (n =25) 500

Weekly (n = 50) 1,000

Daily (n = 250) 5,000

October 1962–December 

1979

Cycolac multielement film 40 Monthly (n = 12) 240

Biweekly (n =25) 500

Weekly (n = 50) 1,000

Daily (n = 250) 5,000

January 1, 1980–March 

1998

Model 7776 TLD 10 Monthly (n = 12) 60

10 Quarterly (n = 4) 20
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 52 of 79

Period of usea Dosimeter

MDLa

(mrem)

Exchange 

frequency

Mean annual missed 

doseb (mrem)

April 1, 1998–2003 

(ongoing)

Model 8823 TLD 10 Monthly (n = 12) 60

10 Quarterly (n = 4) 20

a. Estimated MDLs for each dosimeter technology in the workplace. Dose values were recorded at levels less than the 

MDL.

b. Mean annual missed dose calculated using MDL/2 from NIOSH (2007a).

c. In accordance with the designation adding a class of LANL employees to the SEC, it is not feasible to reconstruct 

photon doses before 1946 (NIOSH 2007b). 

6.5.2 Neutron Missed Dose

Neutron radiation was present around nine reactors at TA-2, -35, and -52; in the plutonium processing 

facilities at TA-1, -21, and -55; around LAMPF at TA-53; and at several other facilities. The approach 

recommended to calculate neutron missed dose can be divided into two periods. The first period is 

1951 to 1978, when NTA film was primarily used; the second period is 1979 and after, when TLDs 

were primarily used. Table 6-24 summarizes the reported limits of detection or dose recording 

thresholds. Estimates of missing neutron doses before 1979 are based on reported photon doses, 

adjusted in accordance with Table A-2 in Attachment A, and use of the applicable NP ratio.

Table 6-24. Neutron dosimeter period of use, type, MDL, exchange frequency, and potential annual 

missed dose.

Period of use Dosimeter

MDL 

(mrem)

Exchange 

frequency

Mean annual 

missed dose

(mrem)a

1951–1962 Brass-cadmium badge <50 Monthly (n = 12) <300b

April 1951–1978

(limited use with TLDs to 1995)

NTA film Biweekly (n = 25) <625b

Weekly (n = 50) <1,250b

1962–1978 Cycolac multielement badge Daily (n = 250) <6,250b

1979–1998

(some quarterly exchange

beginning in 1996)

Model 7776 TLD badge 10 Monthly (n = 12) 60

Quarterly (n = 4) 20

1995–present TED 20 Quarterly (n = 4) 40

1998–present

(40% were exchanged quarterly 

by February 2002)

Model 8823 TLD badge 10 Monthly (n = 12) 60

Quarterly (n = 4) 20

a. Mean annual missed neutron dose calculated using MDL/2 from NIOSH (2007a).

b. NP ratio should be used to estimate missed doses during these periods.

6.6 ORGAN DOSE

Calculation of the POC requires an estimate of the organ dose because the claim is normally specific 

to disease in an organ. This is estimated from uncertainty distributions of parameters related to 

dosimeter response, radiation type, energy, and worker orientation in the field. 

Appendix A of NIOSH (2007a) discusses conversion of measured doses to organ DE, and 

Appendix B contains appropriate DCFs for each organ, radiation type, and energy range based on the 

type of monitoring performed. The selection of worker orientation is important to the calculation of 

organ dose. Examples of common exposure orientations are listed in NIOSH (2007a, Table 4.2); 

however, an assumption of 100% anterior-to-posterior geometry is typically the most representative, 

and is generally used for dose reconstruction. For exposures in which other geometries might result 

in a more accurate dose, the dose reconstructor is referred to the guidance in NIOSH (2007a). 
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 53 of 79

6.7 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

[1] Widner, Thomas E., Certified Health Physicist (CHP), Certified Industrial Hygienist (CIH). 

ChemRisk. Health Physicist. 2004. 

The urchin initiator was a sphere that consisted of a hollow beryllium shell, with a solid 

spherical beryllium pellet nested inside. During implosion, violent turbulence quickly mixed the 

polonium and beryllium together and caused the emission of neutrons.

[2] Widner, Thomas E., CHP, CIH. ChemRisk. Health Physicist. 2007. 

The numbers of workers at LANL each year were obtained from a number of LANL sources;

these values are quite uncertain for some years. For some years, the numbers of total 

workers supplied by LANL is smaller than the number of people who received dosimeters. 

Sources of workforce data for Table 6-2 included a LANL Human Resources – Workforce Data 

& Analysis and Los Alamos publication LASL-77-25 (LASL 1978). The major source of 

uncertainty appears to stem from variability in the worker types included in the count each 

year. Worker categories include full-time regular, part-time, limited-term, students, 

subcontractors, and visitors. More recent data have worker type information recorded, but 

historical data might not. For example, early military workers on the site were categorized as 

“visitors.”

[3] Widner, Thomas E., CHP, CIH. ChemRisk. Health Physicist. 2005. 

With NTA films used at LANL, the estimated standard error was probably larger than that for 

film badge measurement of photon doses, and varied significantly with the energy of the 

neutrons. Given the lack of specific technical information obtained in relation to dosimetry 

systems for early LANL history, measurement uncertainties have been estimated based on 

reported values for contemporary systems in use at other facilities. Based on the technical 

basis document for the Hanford Site (ORAUT 2010) and additional values for NTA film 

adapted from ORAUT (2006a), an overall bias factor for NTA film is estimated at 1.5 (range in

bias 0.5 to 1.5) with a systematic uncertainty factor of 1.5.

[4] Widner, Thomas E., CHP, CIH. ChemRisk. Health Physicist. 2004. 

The P-6 gamma range was in the Physics Building. It was normally used to calibrate 

instruments rather than dosimeter badges, with 60Co and 137Cs sources of different magnitudes 

and possibly a 226Ra source for low-range calibrations. The sources were positioned in a 

concrete well and moved up and down to obtain different dose rates to the instruments at the 

surface. Dose rates were measured with Victoreen R-chambers because scattering 

contributed significantly to the radiations emerging from the well (Widner 2004).
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 54 of 79

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GLOSSARY

absorbed dose, D 

Amount of energy (ergs or joules) deposited in a substance by ionizing radiation per unit mass 

(grams or kilograms) of the substance and measured in units of rads or grays. See dose.

accreditation

For external dosimetry, the assessment of whether or not a personnel dosimetry system meets 

specific criteria. The assessment includes dosimeter performance and the associated quality 

assurance and calibration programs.

accuracy

The characteristics of an analysis or determination that ensures that both the bias and 

precision of the resultant quantity will remain within the specified limits.

albedo dosimeter

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

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

algorithm

Set of rules or steps for solving a problem, especially for calculating a value.

alpha radiation

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

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

charge of +2.

anterior-posterior

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

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

attenuation

Process by which absorption and scattering reduces the number of particles or photons 

passing through a body of matter.

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

BF3 chamber or counter

Proportional counter using a gaseous BF3 compound to detect slow neutrons through their 

interaction with boron.

backscatter

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

beta dose

Designation (i.e., beta) on some records for external dose from beta and less-energetic X-ray 

and gamma radiation, often for shallow dose or dose to the lens of the eye.
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beta particle

See beta radiation.

beta radiation

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

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

beta particle is a positron.

buildup

Increase in flux or dose due to scattering in the medium.

calibration

Adjustment or determination of the response or reading of an instrument relative to a standard 

or a series of conventionally true values.

calibration blank

Dosimeter not exposed to radiation. This dosimeter establishes the dosimetry system base 

line or zero dose value.

collective dose equivalent

Sum of the dose equivalents of all individuals in an exposed population in units of person-rem 

or person-sievert. See dose.

confidence interval or level

The interval about an estimate of a stated quantity within which the value of the quantity is 

expected to be with a specified probability. See uncertainty.

control dosimeter

Dosimeter used to establish the background dose at one of several locations at Los Alamos. 

Control dosimeters were used to set the zero points for personnel dosimeter reading. Control 

dosimeters were also called “blanks” and “unirradiated dosimeters.”.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

Cycolac

Commercial name for a plastic that was used in some LANL dosimeters, which informally took 

on that name. It contained styrene, butadiene, and acrylonitrile.

deep absorbed dose (Dd)

Absorbed dose in units of rem or sievert at a depth of 1 centimeter (1,000 milligrams per 

square centimeter). See dose.

deep dose equivalent (Hd)

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

per square centimeter). See dose.

densitometer

Instrument that uses a photoelectric cell to measure the transition of light through developed 

X-ray film to determine the optical density.
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density reading

See optical density.

DOE Laboratory Accreditation Program (DOELAP)

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

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

records, and calibration programs.

dose

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

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

units of roentgens, rads, reps, or grays.

dose conversion factor (DCF)

Multiplier for conversion of potential dose to the personal dose equivalent to the organ of 

interest (e.g., liver or colon). In relation to radiography, ratio of dose equivalent in tissue or 

organ to entrance kerma in air at the surface of the person being radiographed.

dose equivalent (DE or H)

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

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

from the absorbed dose. See dose.

dose equivalent index

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

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

within a sphere at any depth for the respective neutron energies even though the maximum 

dose occurred at different depths and discounting the outer 0.07-millimeter-thick shell. Also 

called unrestricted dose equivalent index.

dosimeter

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

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

pocket ionization chamber, thermoluminescent dosimeter, and track-etch dosimeter.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

DuPont 552 film packet

Film packet containing DuPont 502 sensitive film and DuPont 510 insensitive film.

DuPont 558 film packet

Film packet containing DuPont 508 film with sensitive and insensitive emulsions on either side.

error

Difference between the correct, true, or conventionally accepted value and the measured or 

estimated value. Sometimes used to mean estimated uncertainty. See accuracy and 

uncertainty.
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exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

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

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

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

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

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

external dose

Dose received from radiation emitted by sources outside the body.

extremities

The portion of the arm from and including the elbow through the fingertips and the portion of 

the leg from and including the knee and patella through the toes.

fast neutron

Neutron with energy equal to or greater than 10 kiloelectron-volts. This type of neutron causes 

fission in some isotopes (e.g., 238U, 239Pu). See intermediate neutron and slow neutron.

film

In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter.

film density

See optical density.

film dosimeter

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

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

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

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

densitometer. Also called film badge.

filter

Material used in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response.

gamma radiation

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

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

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

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

gamma ray, particle, or photon (γ)

See gamma radiation.
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geometric mean

The nth root of the product of all the members of a set of positive numbers, where n is the 

number of members. The sample geometric mean is designed for averaging ratio or 

proportion data. It is equivalent to taking logarithms of the sample values (i.e., transforming 

the sample), finding the arithmetic mean of the logs, and then retransforming back to the 

original scale (by taking antilogs). It can only be used when all the sample values are greater 

than zero.

geometric standard deviation (GSD)

In probability theory and statistics, the geometric standard deviation describes the spread of a 

set of numbers whose preferred average is the geometric mean

glovebox

Enclosure with special rubber gloves through which an operator can handle radioactive or 

toxic material without risk of injury or contamination normally operated at a slightly reduced 

pressure so that air leakage, if any, is inward.

gray (Gy)

International System unit of absorbed radiation dose, which is the amount of energy from any 

type of ionizing radiation deposited in any medium; 1 gray equals 1 joule per kilogram or 

100 rads.

intermediate neutron

Neutron with energy between 0.5 electron-volts and 10 kiloelectron-volts. See fast neutron

and slow neutron.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray

radiation.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

kerma

Measure in units of absorbed dose (usually grays but sometimes rads) of the energy released 

by radiation from a given amount of a substance. Kerma is the sum of the initial kinetic 

energies of all the charged ionizing particles liberated by uncharged ionizing particles 

(neutrons and photons) per unit mass of a specified material. Free-in-air kerma refers to the 

amount of radiation at a location before adjustment for any external shielding from structures 

or terrain. The word derives from kinetic energy released per unit mass.

kiloelectron-volt (keV)

Unit of particle energy equal to 1,000 (1 × 103

) electron-volts.
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kiva

One of the remotely controlled critical assembly buildings associated with the Critical 

Experiment Facility at LANL. From Hopi, ki- means house, but the meaning of -va is unknown.

luminescence

Emission of light from a material as a result of some excitation. See thermoluminescence.

Manhattan Engineer District (MED)

Subdivision of the U.S. Army Corps of Engineers that administered the World War II 

Manhattan Project to develop the first nuclear bomb. The word Manhattan was chosen to 

divert attention from the Project’s real purpose. The U.S. Atomic Energy Commission 

assumed control of MED facilities and activities in 1946.

megaelectron-volt (MeV)

Unit of particle energy equal to 1 million (1 × 106

) electron-volts.

minimum detectable activity

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

minimum detectable level (MDL)

See minimum detectable activity.

missed dose

In relation to external dose, dose to monitored workers that was not measured or recorded due 

to such factors as a missing or damaged dosimeter or a result below the detection limits of the 

dosimeter. Missed dose is especially important in the early years of radiation monitoring, 

when relatively high detection limits were combined with short exchange periods.

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.

multiple-collision neutron dose

Dose in relation to the neutron flux through tissue based on the assumption that two or more 

interactions per neutron occurs and results in greater energy deposition.

n, or n unit

Early unit of neutron dose used with pocket ionization chambers; the quantity of neutron 

radiation producing the same ionization in a 100-roentgen PIC as 1 roentgen of gamma

radiation. See pencil dosimeter.

neutron

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

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

neutron film dosimeter

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

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

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

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

radioactive by absorbing free neutrons. See neutron.

nonpenetrating dose (NP, NPEN)

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

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

See dose.

notional dose

Dose value assigned for lower dosed workers (at some facilities but not at LANL), often based 

on a small fraction of the regulatory limit.

nuclear emulsion

Thick photographic coating in which the tracks of various fundamental particles show as black 

traces after development. The number of tracks in a given area is a measure of the dose from 

that radiation. See nuclear track emulsion, type A.

nuclear track emulsion, type A (NTA)

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

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

magnification.

open window

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

light protection). See film dosimeter.

optical density

Measure of the degree of opacity of photographic or radiographic film defined as OD = log10

(I0/I), the base-10 logarithm of the ratio of the reference light intensity I0 (without film) to the 

transmitted light intensity (through the film). Also called film density and density reading.

pencil dosimeter

See pocket ionization chamber.

penetrating dose (PEN)

Dose from moderate to higher energy photons and neutrons that penetrates the outer layers of 

the skin. See dose.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon radiation

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

waves to gamma rays.

pit

Package at the center of an implosion weapon that contains the machined fissile material that 

begins the fission chain reaction. Also called primary.
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pocket dosimeter

See pocket ionization chamber.

pocket ionization chamber (PIC)

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

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

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

dosimeter.

precision

Describes dispersion of measurements in relation to a measure of location or central

tendency.

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Act, the percent likelihood, at the 99th percentile, that a worker incurred a 

particular cancer from occupational exposure to radiation.

PuF4 source

Neutron source with plutonium tetrafluoride activating material. The source was used to 

duplicate the neutron energies in plutonium facilities.

quality factor (Q)

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei.

random errors

When a given measurement is repeated and the values do not agree exactly. The causes of 

the disagreement between the values must also be the causes of their differences from the 

true value. See systematic error.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.
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rep

Historical quantity of radiation (usually other than X-ray or gamma radiation) originally defined 

as 83 ergs absorbed per gram in the body and redefined in the 1940s or early 1950s as the 

amount that would liberate the same amount of energy (93 ergs per gram) as 1 roentgen of Xor gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 8.38 milligray. The word derives from roentgen equivalent physical.

roentgen (R)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts).

scattering

Change in direction of radiation by refraction or reflection, often accompanied by a decrease in 

radiation due to absorption by the refracting or reflecting material.

shallow absorbed dose (Ds)

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

of specified geometry and composition.

shallow dose equivalent (Hs)

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

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

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

sievert (Sv)

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

skin dose

See shallow dose equivalent.

slow neutron

Neutron with energy less than 1 electron-volt. See fast neutron and intermediate neutron.

Special Nuclear Material (SNM)

Plutonium or uranium enriched to a higher-than-natural assay including 239Pu, 233U, uranium 

containing more than the natural abundance of 235U, or any material artificially enriched in one 

of these isotopes.

systematic errors

When a given measurement is repeated and the values differ from the true value by the same 

amount. See random error.

Technical Area (TA)

Area of land at LANL for assigned operations.
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thermal neutron

Neutron in thermal equilibrium with its surroundings having an average energy of 

0.025 electron-volts.

thermoluminescence

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD)

Small block or crystal of lithium fluoride in a thermoluminescent dosimeter. A TLD-600 

dosimeter contains a chip made from more than 95% 6

Li for neutron radiation detection, and a 

TLD-700 dosimeter contains a chip made from more than 99.9% 7

Li for photon and beta 

radiation detection. Also called crystals.

thermoluminescent dosimeter chip

Small block or crystal of lithium fluoride in a thermoluminescent dosimeter. A TLD-600 

dosimeter contains a chip made from more than 95% 6

Li for neutron radiation detection, and a 

TLD-700 dosimeter contains a chip made from more than 99.9% 7

Li for photon and beta 

radiation detection. Also called crystals.

tissue equivalent

Substance with response to radiation equivalent to tissue. A tissue-equivalent response is an 

important consideration in the design and fabrication of radiation measuring instruments and 

dosimeters.

track-etch dosimeter (TED)

Device for evaluation of fast neutron dose through examination of traces left by the neutrons 

on the Columbia Resin Number 39 emulsion.

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

uncertainty

Standard deviation of the mean of a set of measurements. The standard error reduces to the 

standard deviation of the measurement when there is only one determination. See accuracy, 

confidence level, and error. Also called standard error.

Van de Graaff accelerator

Direct current high-voltage accelerator in which a terminal in a pressurized tank accelerates 

the beam. The voltage is generated by charge carried on an insulating belt.

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. See

dose.
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X-ray radiation

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

particles. X-rays are produced by various mechanisms including electron shell transitions 

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

gamma rays, but gamma photons originate inside the nucleus of an atom.

Zia

Private firm created as the facilities contractor for LANL and its support community.
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TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Unmonitored Photon Dose..................................................................................................... 72

A.2 Adjustments to Reported Photon Doses................................................................................. 73

A.3 Missed Beta/Photon Dose...................................................................................................... 74

A.4 Unmonitored Neutron Dose.................................................................................................... 74

A.5 Recommended Dose Conversion Factors.............................................................................. 75

LIST OF TABLES

TABLE TITLE PAGE

A-1 Worker gamma dose statistics ............................................................................................... 72

A-2 Recommended adjustments to reported photon doses .......................................................... 73

A-4 Recommended uncertainty factors for reported doses ........................................................... 74

A-4 Beta/photon dosimeter period of use, type, MDL, exchange frequency, and potential 

annual missed doses .............................................................................................................75

A-5 Recommended DCFs for dose assessments ......................................................................... 75

ATTACHMENT A

GUIDANCE FOR DOSE RECONSTRUCTORS: OCCUPATIONAL EXTERNAL 

DOSE FOR MONITORED WORKERS

Page 1 of 5
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A.1 UNMONITORED PHOTON DOSE

Table A-1 summarizes the lognormal probability statistical parameters for LANL dosimeter results that 

are equal to or exceed 50 mrem for all years of record. These data can be used to estimate 

unmonitored doses. They can also be used to estimate unrecorded doses for which no records exist. 

The reported doses that are the basis of this table have not been corrected for potential missed 

doses. 

Table A-1. Worker gamma dose statistics.

a

Yeara

LANL recorded gamma dose datab Lognormal fit

No. of workers reported 

gamma dose ≥ 50 mrem

Dose (mrem) Dose (mrem)

Mean Maximum Median 95% GSD

1946 245 3.52E+03 1.18E+05 5.07E+02 9.18E+03 5.82

1947 656 6.96E+02 1.65E+04 2.52E+02 1.85E+03 3.36

1948 683 4.82E+02 1.34E+04 1.99E+02 1.29E+03 3.13

1949 1,091 3.85E+02 1.07E+04 1.56E+02 8.82E+02 2.87

1950 1,364 3.17E+02 2.02E+04 1.40E+02 6.68E+02 2.58

1951 2,182 1.08E+03 1.36E+04 4.82E+02 4.46E+03 3.87

1952 936 7.12E+02 1.00E+04 3.10E+02 2.29E+03 3.37

1953 908 6.72E+02 1.07E+04 3.22E+02 2.28E+03 3.28

1954 1,007 8.61E+02 8.99E+03 4.04E+02 3.16E+03 3.49

1955 823 8.81E+02 8.82E+03 3.98E+02 3.01E+03 3.43

1956 1,068 1.15E+03 1.86E+04 4.92E+02 4.47E+03 3.82

1957 780 6.46E+02 1.35E+04 2.88E+02 2.08E+03 3.33

1958 1,071 4.38E+03 3.50E+06 4.67E+02 4.26E+03 3.83

1959 740 5.11E+02 5.51E+03 2.37E+02 1.63E+03 3.23

1960 1,059 4.59E+02 5.00E+03 2.05E+02 1.33E+03 3.12

1961 749 3.82E+02 4.69E+03 1.94E+02 1.13E+03 2.92

1962 937 3.82E+02 5.65E+03 2.15E+02 1.14E+03 2.75

1963 640 3.24E+02 3.28E+03 1.84E+02 9.25E+02 2.67

1964 606 3.65E+02 4.10E+03 1.98E+02 1.09E+03 2.82

1965 635 5.13E+02 9.69E+03 2.38E+02 1.62E+03 3.20

1966 568 3.69E+02 4.67E+03 2.03E+02 1.09E+03 2.79

1967 551 4.06E+02 6.24E+03 2.08E+02 1.21E+03 2.92

1968 609 3.23E+02 4.03E+03 1.66E+02 9.30E+02 2.85

1969 798 3.58E+02 4.95E+03 1.55E+02 9.99E+02 3.11

1970 590 5.79E+02 4.78E+03 2.51E+02 1.95E+03 3.48

1971 531 5.52E+02 5.00E+03 2.44E+02 1.85E+03 3.43

1972 554 4.95E+02 5.73E+03 2.27E+02 1.52E+03 3.17

1973 553 5.84E+02 4.63E+03 3.11E+02 2.00E+03 3.10

1974 685 4.01E+02 2.77E+03 2.24E+02 1.28E+03 2.88

1975 944 3.82E+02 3.17E+03 2.04E+02 1.19E+03 2.91

1976 952 3.58E+02 3.29E+03 1.90E+02 1.11E+03 2.92

1977 921 4.22E+02 4.46E+03 2.01E+02 1.31E+03 3.13

1978 938 3.33E+02 3.68E+03 1.79E+02 9.83E+02 2.82

1979 779 2.70E+02 1.87E+03 1.69E+02 7.90E+02 2.56

1980 666 2.69E+02 1.85E+03 1.73E+02 7.82E+02 2.50

1981 832 2.71E+02 2.40E+03 1.67E+02 7.87E+02 2.57

1982 821 3.25E+02 2.27E+03 1.84E+02 1.01E+03 2.82

1983 1,017 2.96E+02 2.16E+03 1.72E+02 8.68E+02 2.67

ATTACHMENT A

GUIDANCE FOR DOSE RECONSTRUCTORS: OCCUPATIONAL EXTERNAL 

DOSE FOR MONITORED WORKERS

Page 2 of 5
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Yeara

LANL recorded gamma dose datab Lognormal fit

No. of workers reported 

gamma dose ≥ 50 mrem

Dose (mrem) Dose (mrem)

Mean Maximum Median 95% GSD

1984 1,513 2.32E+02 2.47E+03 1.37E+02 6.39E+02 2.55

1985 733 3.54E+02 1.88E+03 2.10E+02 1.12E+03 2.77

1986 527 3.21E+02 1.71E+03 2.03E+02 1.00E+03 2.64

1987 420 3.14E+02 2.74E+03 2.08E+02 9.26E+02 2.48

1988 428 2.97E+02 1.71E+03 1.96E+02 9.04E+02 2.54

1989 436 2.56E+02 1.47E+03 1.75E+02 7.34E+02 2.39

1990 340 2.41E+02 2.28E+03 1.62E+02 6.60E+02 2.35

1991 298 1.87E+02 1.86E+03 1.37E+02 4.63E+02 2.10

1992 221 1.38E+02 5.65E+02 1.15E+02 2.99E+02 1.79

1993 224 1.40E+02 9.46E+02 1.14E+02 3.11E+02 1.85

1994 215 1.47E+02 6.89E+02 1.20E+02 3.31E+02 1.86

1995 193 1.26E+02 2.85E+02 1.11E+02 2.55E+02 1.66

1996 267 1.74E+02 6.00E+02 1.40E+02 4.14E+02 1.93

1997 327 2.00E+02 1.21E+03 1.52E+02 4.93E+02 2.04

1998 313 1.92E+02 1.09E+03 1.41E+02 4.76E+02 2.09

1999 253 1.58E+02 6.41E+02 1.26E+02 3.67E+02 1.91

2000 215 1.20E+02 5.96E+02 1.01E+02 2.54E+02 1.75

2001 281 1.69E+02 2.13E+03 1.29E+02 3.93E+02 1.97

2002 380 1.86E+02 1.37E+03 1.42E+02 4.61E+02 2.05

2003 453 2.50E+02 2.35E+03 1.70E+02 6.39E+02 2.24

2004 282 1.55E+02 1.23+03 1.06E+02 4.06E+02 2.26

2005 339 1.57E+02 1.27E+03 1.16E+02 9.31E+02 3.55

2006 348 1.78E+02 1.33E+03 1.17E+02 4.59E+02 2.30

2007 284 1.94E+02 9.63E+02 1.33E+02 5.22E+02 2.30

2008 194 1.58E+02 1.06E+03 1.11E+02 4.42E+02 2.32

a. In accordance with the addition of a class of LANL employees to the Special Exposure Cohort, it is infeasible to 

reconstruct gamma doses before 1946 (NIOSH 2007b). 

b. Individual dosimeter records analyzed only if gamma dose was equal to or greater than 50 mrem.

When necessary, it is recommended that dose reconstructors assign the median (that is, geometric 

mean) gamma dose from Table A-1 to an unmonitored worker for each year of employment; however, 

LANL had a strict policy for monitoring exposed employees, with some exceptions described above. 

Also, when unmonitored photon dose is assigned, the addition of missed dose is required as well. 

The dose reconstructor must make an assumption about the likely dosimeter exchange frequency, 

and assume all unmonitored dose occurred in one cycle, with missed dose for all other cycles. 

A.2 ADJUSTMENTS TO REPORTED PHOTON DOSES

Table A-2 lists adjustments to reported LANL photon doses.

Table A-2. Recommended adjustments to reported photon doses.

Period Dosimeter Facilities/operations

Adjustment to 

reported dose References

1949 Brass clip film 

badge

Plutonium areasa Multiply reported photon 

doses by 1.3.b Traub, Sherpelz, and Taulbee 2005

Sep 1961–

Oct 1964

Brasscadmium Photon exposures >200 keV 

(reactors, uranium production, 

Multiply reported photon 

doses that exceed 

LASL 1969 (10/13/64 

memorandum)

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badge accelerators, calibrations, 

waste handling, radioactive 

lanthanum, etc.) 

100 mrem by 4.

a. Plutonium areas have included TA-1, TA-21, TA-55, and some areas of TA-3. 

b. Based on calculated dose rate due to 30- to 250-keV photons, based on measured dose (P2/C7) for a generic 6-kg pit 

(5 years), from Table A.3 of Traub, Sherpelz, and Taulbee (2005). 

Based on site-specific documentation (LASL 1959, 1969, 1977, 1979; LANL 1986, 1989, 1996, 2001, 

2003) and characterizations of similar dosimeters used at other sites, the standard error in recorded 

film badge doses from photons of any energy is estimated to have been ±30%. The estimated 

standard error for recorded doses from beta radiation was probably the same, but for unknown 

mixtures of beta and photon radiations the standard error was likely somewhat larger than 30%. 

The standard error for neutron dose readings of approximately 100 mrem from NTA film is estimated 

to have been ±50%. 

Based on these estimates and analysis of DOELAP testing result summaries for the LANL Model 

7776 and Model 8823 dosimeters (Test Sessions 22 and 36, respectively), Table A-3 lists

recommended uncertainty factors to be applied to worker doses reported by LANL.

Table A-4. Recommended uncertainty factors for reported doses.

Radiation

Film badges

(1946-1979)

NTA film

(1951-1978)

Model 7776 

dosimeter

(1980-1998)

Model 8823 

dosimeter

(1999-Present)

Photon (deep) ±30% Not applicable ±14% ±19%

Beta (shallow) ±30% Not applicable ±16% ±14%

Neutron (deep) Not applicable ±50% ±8% ±8%

A.3 MISSED BETA/PHOTON DOSE

Missed dose is the dose that might not have been accounted for on an individual’s records because 

the records might have indicated zero dose due to the detection limitations of the film or TLD. 

Several exchange frequencies were in use at any time, so the dose reconstructor needs to determine 

the exchange frequency that applies to a specific worker from individual records. The values in the 

last two columns of Table A-4 can be considered maximum annual missed doses for dose 

reconstruction. Beginning with 1979, badge exchange frequencies should be assumed to be monthly 

if specific information to the contrary for the claimant is not available (ORAUT 2006b).

Missed beta/photon dose is entered into the Interactive RadioEpidemiological Program (IREP) as a 

lognormal distribution with a geometric mean consistent with Table A-4 and a GSD of 1.52.

A.4 UNMONITORED NEUTRON DOSE

There should not, typically, be significant neutron exposure of unmonitored workers, with the 

exception of 1946 to 1949 (NIOSH 2007b). However, if there is an estimation of neutron dose, the 

recommended option is to apply the NP distribution data from Table 6-22 to measured or estimated 

missed or unmonitored photon doses. This process of calculation of a neutron dose is based on the 

expectation that the neutron dose to unmonitored workers is equivalent to the median neutron dose 

ATTACHMENT A

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DOSE FOR MONITORED WORKERS

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ATTACHMENT A

GUIDANCE FOR DOSE RECONSTRUCTORS: OCCUPATIONAL EXTERNAL 

DOSE FOR MONITORED WORKERS

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measured for monitored workers. The application of the NP ratios is discussed in Section 6.4.3, and 

is facilitated with automated options in approved external dose spreadsheets. 

Table A-4. Beta/photon dosimeter period of use, type, MDL, exchange frequency, and potential 

annual missed doses.

Period of use Dosimeter

Deep 

MDLa

(mrem)

Nonpenetrating 

MDLa

(mrad)

Exchange 

frequency

Geometric mean 

annual missed doseb,c

Deep dose

(mrem)

Nonpenetrating 

dose (mrad)

July 1946–1948 Brass film badge 40 Nonpenetrating 

dose was not 

measured.

Monthly (n = 12) 240 Unmonitoredd

Biweekly (n = 25) 500

Weekly (n = 50) 1,000

Daily (n = 250) 5,000

1949–September 

1962

Brass clip, brass-lead, 

brass-cadmium film 

badges

40 40 Monthly (n = 12) 240 240

Biweekly (n = 25) 500 500

October 1962–

December 1979

Cycolac multielement 

film badge

Weekly (n = 50) 1,000 1,000

Daily (n = 250) 5,000 5,000

January 1, 1980–

March 1998

Model 7776 TLD 10 30 Monthly (n = 12) 60 180

Quarterly (n = 4) 20 60

April 1, 1998–

present

Model 8823 TLD 10 30 Monthly (n = 12) 60 180

Quarterly (n = 4) 20 60

a. Estimated MDLs for each dosimeter technology in the workplace. 

b. See Table 6-14 footnotes for guidance on calculation of unmonitored nonpenetrating dose.

c. Mean annual missed dose calculated using MDL/2 from NIOSH (2007a).

d. According to the designation adding a class of LANL employees to the SEC, it is not feasible to reconstruct nonpenetrating doses 

before 1949 (NIOSH 2007b).

The application of the data from Table 6-22 to measured or estimated photon doses can be 

accomplished through Monte Carlo simulation. 

A.5 RECOMMENDED DOSE CONVERSION FACTORS

DCFs should be selected from NIOSH (2007a) for the dose quantities specified in Table A-5 for the 

periods of interest.

Table A-5. Recommended DCFs for dose assessments. 

Period Recommended photon DCF Recommended neutron DCF

1946–1984 Exposure (R) to organ DE (HT) Deep DE [Hp,slab(10)]

1985–present Deep DE [H to organ DE (HT) p(10)] to organ DE (HT)
Document No. ORAUT-TKBS-0010-6 Revision No. 03 Effective Date: 03/21/2013 Page 76 of 79

TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Introduction ............................................................................................................................ 76

B.1.1 Purpose...................................................................................................................... 76

B.1.2 Method ....................................................................................................................... 76

B.2 Summary ............................................................................................................................... 76

B.3 Calculated NP Dose Ratio Statistical Parameters, 1944–2008............................................... 78

LIST OF TABLES

B-1 Statistical parameters for selected periods of LANL recorded doses...................................... 78

B-2 Statistical analysis of LANL recorded doses........................................................................... 78

LIST OF FIGURES

B-1 LANL worker GM and 95th-percentile NP dose ratio, 1950 to 2008 ....................................... 77

B.1 INTRODUCTION

This attachment describes analyses performed to calculate the statistical parameters of lognormal 

distributions of NP dose ratios using recorded doses for LANL workers (LANL 2004) from 1944 

through November 2008. This information extends the information in this TBD. 

B.1.1 Purpose

The purpose of this attachment is to document the method of analysis and to present the calculated 

statistical lognormal distribution parameters [i.e., geometric mean (GM), GSD, and the 95th percentile] 

for each year and selected multiyear periods. 

B.1.2 Method

The analysis of NP dose ratios used LANL recorded dose data in LANL (2004). Only annual neutron 

and photon doses that equal or exceed 50 mrem were used. The analysis results are in Fix (2011).

B.2 SUMMARY

The calculated NP dose ratio parameters for each year from 1944 through 2008 are listed in Table B2 (Fix 2011). The pattern in the GM and 95th-percentile parameters are shown in Figure B-1 with 

significant events from the history of LANL dosimetry presented in this attachment. A documented 

personal communication with a LANL external dosimetry site and subject expert, who performed 

many workplace dose measurements and evaluations, was also done (Fix 2012). 

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Figure B-1. LANL worker GM and 95th-percentile NP dose ratio, 1950 to 2008.

The statistical parameters in Table A-1 are based on 412,000 (1950–2008) measured annual dose 

results. Significant dosimetry-related events noted in Figure B-1 include:

• Implementation of the use of NTA film in 1951 (Littlejohn 1961). LANL had earlier used a 

similar nuclear track emulsion glass plate for routine dosimetry. The use of NTA to measure 

neutron dose continued until 1995 when the routine use of track-etch dosimetry was 

implemented.

• LANL had available an instrument capability to measure the neutron DE in rem that was used 

in 1963 (Hankins 1963). This capability provides a means to assess the performance of 

dosimeter-measured workplace doses with an accurate instrument measurement.

• LANL introduced thermoluminescent dosimetry capabilities over many years. The earliest 

noted studies occurred in 1970 (Hankins 1969), whereas regular assignment of TLDs to all 

workers was occurring in 1978 (Lawrence 1978).

• LANL participated in performance testing of the standard that eventually was issued as 

ANSI/HPS N13.11 in 1983 (HPS 1983). This standard formed the basis of the DOELAP for 

personnel dosimetry. LANL was DOELAP-accredited in 1985 for requested categories.

• LANL implemented updated TLD capabilities with the Bicron/Harshaw 8823 system in 1998,

which uses TED capabilities for selected workers (Hoffman and Mallett 1999b).

The data in Figure B-1 can be summarized over selected periods based on the foregoing dosimetry 

operational considerations. Several options are listed in Table B-1.

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Table B-1. Statistical parameters for selected periods of LANL recorded doses.

Period Description

NP dose ratio

GM GSD

95th 

percentile

1950–2008 All years of NTA, TLD, or TED dose data 1.04 3.06 6.43

1979–2004 ORAUT-TKBS-0010-6 (2009) period 1.55 2.73 8.27

1979–2008 ORAUT-TKBS-0010-6 (2009) plus 2005–2008 1.50 2.68 7.77

1980–2008 Era of dosimeter performance testing 1.53 2.66 7.90

1985–2008 DOELAP-accredited dosimetry systems 1.60 2.61 8.15

In the absence of a known or suspected neutron exposure, a reasonable assumption for neutron dose 

reconstruction is a GM of 1.6 and a GSD of 2.7. 

ORAUT-TKBS-0010-6

Table 6-22 of this TBD summarizes recommended NP dose ratios for identified sources of neutron 

radiation. The analyses in this Attachment primarily pertain to Neutron Source Types A and D in 

Table 6-22. It is evident from the GM and 95th-percentile values that the results of the current 

analysis and those used in the development of recommendations in this TBD are equivalent. The 

values in Table 6-22 were updated to include the changed 95th-percentile value in the last row; all 

other values were verified.

B.3 CALCULATED NP DOSE RATIO STATISTICAL PARAMETERS, 1944–2008

Table B-2 summarizes the statistical analysis of recorded doses at LANL.

Table B-2. Statistical analysis of LANL recorded doses (Fix 2011).

Year

Number of workers NP dose ratio

Total 

monitored

Positive 

recorded dose

Photon, neutron 

dose ≥50a Mean Max. GM GSD

95th 

percentile

1944 9 7 0

1945(b) 812 413 2 139.32 275.00 31.62 1.47 261.43

1946(b) 508 391 7 14.02 16.94 13.93 1.12 16.53

1947 1,237 1,004 0

1948 2,080 1,383 0

1949 3,177 1,817 21 3.32 10.88 2.25 2.32 10.63

1950 3,895 1,981 75 4.86 24.00 2.12 2.64 14.78

1951 4,257 2,538 142 3.01 25.75 1.32 3.37 11.52

1952 2,366 1,262 97 3.81 39.94 2.31 2.08 9.40

1953 1,878 1,151 131 1.53 9.18 0.80 3.04 5.45

1954 2,068 1,166 99 0.85 5.14 0.48 2.77 2.92

1955 1,984 956 100 0.63 4.80 0.39 2.11 1.62

1956 2,287 1,189 112 0.84 8.75 0.49 2.60 2.28

1957 2,539 1,054 78 2.66 31.17 0.77 4.87 10.53

1958 3,032 1,378 133 1.01 8.27 0.36 4.43 4.29

1959 2,930 1,552 186 0.75 11.57 0.37 3.16 2.30

1960 3,622 1,718 179 1.04 6.80 0.58 2.44 2.88

1961 3,973 1,395 24 0.24 0.83 0.16 2.10 0.63

1962 4,119 2,101 45 1.02 4.42 0.49 3.63 3.93

1963 4,176 1,702 71 1.01 8.40 0.40 5.36 5.46

1964 4,103 1,580 89 0.60 3.38 0.35 2.90 2.02

1965 4,222 1,352 101 0.75 10.00 0.40 2.71 2.14

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Year

Number of workers NP dose ratio

Total 

monitored

Positive 

recorded dose

Photon, neutron 

dose ≥50a Mean Max. GM GSD

95th 

percentile

1966 4,446 1,514 98 0.53 4.14 0.29 3.41 1.89

1967 4,072 1,493 84 0.95 9.60 0.44 3.22 3.42

1968 3,861 2,044 114 1.26 9.00 0.62 3.59 4.79

1969 3,980 2,679 159 1.18 9.29 0.59 3.17 3.87

1970 4,031 1,018 150 0.49 6.80 0.28 2.59 1.48

1971 3,775 806 85 0.48 2.45 0.25 3.24 1.69

1972 3,877 938 127 0.78 5.40 0.39 3.48 2.86

1973 3,866 924 202 0.91 10.20 0.48 2.71 2.67

1974 4,337 1,392 250 0.92 18.07 0.42 3.46 3.17

1975 4,716 1,874 229 0.67 10.20 0.38 2.91 2.23

1976 5,254 2,165 136 0.37 2.80 0.20 3.39 1.42

1977 5,624 1,851 130 0.47 8.67 0.22 3.47 1.69

1978 7,045 2,070 135 0.64 6.50 0.29 3.92 2.44

1979 7,549 2,511 288 1.14 11.33 0.59 3.33 4.15

1980 7,638 3,443 340 1.63 12.80 1.00 2.82 5.51

1981 7,966 5,034 463 1.96 14.50 1.30 2.14 5.35

1982 7,997 4,297 456 1.89 15.30 1.21 2.19 5.07

1983 8,144 4,419 445 1.85 11.57 1.13 2.08 5.00

1984 8,622 6,069 535 1.94 11.57 1.26 2.05 5.29

1985 9,487 4,921 434 2.63 21.40 1.76 1.94 6.36

1986 9,612 2,262 324 3.72 21.40 2.27 2.24 11.59

1987 9,202 1,621 253 3.54 17.71 1.99 2.27 10.75

1988 9,469 1,548 271 3.68 25.80 1.85 2.58 12.59

1989 10,605 1,616 244 3.24 14.33 1.93 2.43 9.58

1990 10,796 1,538 207 2.69 14.71 1.64 2.40 8.80

1991 11,284 2,356 156 3.05 19.90 1.85 2.28 8.73

1992 11,560 1,687 131 2.85 17.24 2.00 2.11 7.29

1993 11,772 1,948 132 3.23 15.03 2.34 2.29 9.18

1994 11,783 3,145 170 3.38 15.88 2.41 2.19 9.37

1995 12,448 3,219 188 3.21 20.92 2.27 2.18 8.21

1996 10,959 2,358 197 2.14 11.69 1.57 1.91 5.09

1997 10,860 2,992 215 1.86 10.78 1.31 2.50 5.61

1998 11,168 2,201 171 1.78 9.16 1.40 2.21 4.50

1999 11,212 1,720 158 1.85 8.34 1.40 2.25 4.94

2000 10,456 1,485 122 2.10 7.03 1.52 2.07 5.22

2001 10,443 1,435 156 1.74 6.91 1.22 2.61 5.39

2002 10,931 1,760 199 1.55 6.91 1.07 2.39 4.22

2003 10,752 2,101 239 1.57 7.91 1.25 2.66 4.77

2004 10,192 1,731 156 1.64 9.86 1.13 2.52 4.79

2005 10,826 2,176 193 1.71 7.47 1.29 2.24 4.42

2006 10,104 2,021 215 1.93 9.24 1.51 2.21 4.70

2007 9,745 1,525 203 1.79 8.01 1.34 2.36 4.73

2008c 8,148 941 106 1.48 5.91 1.05 2.52 4.42

a. Number of workers with recorded neutron and deep photon doses, respectively, ≥ 50 mrem.

b. Doses anticipated to have been estimated following accidents and not routinely measured.

c. Anticipated to represent <1 year of dose data (i.e., November 2008 last noted dose entry).

ATTACHMENT B

LANL NEUTRON-TO-PHOTON (NP) DOSE RATIO ANALYSIS

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