ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 20

Document Title:

Lawrence Livermore National Laboratory 

– Occupational External Dose

Document Number: ORAUT-TKBS-0035-6 

Revision: 00

Effective Date: 10/07/2005

Type of Document: TBD

Supersedes: None

Subject Expert: Paul Szalinski

Document Owner

Approval: Signature on File Approval Date: 10/03/2005

Jay J. Maisler, Team Leader

Approval: Signature on File Approval Date: 10/03/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/03/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 10/07/2005

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-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 2 of 20

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

10/07/2005 00 New technical basis document for the Lawrence Livermore National 

Laboratory – Occupational External Dose. First approved issue. 

Training required: As determined by the Task Manager. Initiated by 

Jay J. Maisler.
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 3 of 20

TABLE OF CONTENTS

SECTION TITLE 

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

PAGE

4

6.1 Introduction ..............................................................................................................................5

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

6.3 Dose Reconstruction Parameters.............................................................................................6

6.3.1 Historical Administrative Practices................................................................................ 7

6.3.2 Dosimetry Technology..................................................................................................8

6.3.3 Calibration ....................................................................................................................9

6.3.4 Workplace Radiation Fields ..........................................................................................9

6.3.5 Recorded Dose Practices ........................................................................................... 11

6.3.6 Interpretation of Reported Data .................................................................................. 12

6.4 Adjustments to Recorded Dose.............................................................................................. 12

6.5 Missed and Unmonitored Dose .............................................................................................. 13

6.6 Bias and Uncertainty..............................................................................................................13

6.7 Dose Reconstruction..............................................................................................................14

6.8 Organ Dose............................................................................................................................15

References ........................................................................................................................................16

Glossary ............................................................................................................................................18

LIST OF TABLES

TABLE PAGE

6-1 LLNL dosimeter type, period of use, exchange frequency, MDL, and 

 potential annual missed dose...................................................................................................7

6-2 Selection of beta and photon radiation energies and percentages ......................................... 10

6-3 Selection of neutron radiation energies and percentages....................................................... 11

6-4 Recorded dose practices........................................................................................................ 11

6-5 Interpretation of reported data................................................................................................ 12

6-6 Adjustments to recorded dose................................................................................................ 13

6-7 Bias and uncertainty...............................................................................................................14
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 4 of 20

ACRONYMS AND ABBREVIATIONS

CR-39 Canadian Resin Number 39, a polycarbonate material used for neutron dosimetry

DCF dose conversion factor

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

EEOICPA Energy Employees Occupational Illness Compensation Program Act

Hp(d) personal dose equivalent at tissue depth d (d = 10 mm or 0.07 mm)

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

LBNL Lawrence Berkeley National Laboratory

LLNL Lawrence Livermore National Laboratory

MDL minimum detection level

MED Manhattan Engineer District

NVLAP National Voluntary Laboratory Accreditation Program

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

WB whole body
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 5 of 20

6.1

Site Profiles, which include Technical basis documents (TBDs), are not official determinations made 

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

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

dose reconstructions at particular sites or categories of sites. They will be revised when additional 

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

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

INTRODUCTION

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

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

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

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 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 DOE employees 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 [probability of causation] guidelines established under subsection (c) ....” 42 

U.S.C. § 7384n(b). Neither the statute nor the probability of causation guidelines (nor the dose 

reconstruction regulation) define “performance of duty” for DOE employees with a covered cancer or 

restrict the “duty” to nuclear weapons work.

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

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

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

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

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

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

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

This Site Profile documents historical practices at the Lawrence Livermore National Laboratory (LLNL) 

and provides information for the evaluation of internal and external dosimetry data for unmonitored 

and monitored workers; it can serve as a supplement to, or substitute for, individual monitoring data.

Workers at the Lawrence Livermore National Laboratory (LLNL) were exposed to radiation from a 

variety of radioactive materials and radiation-producing machines. In addition, many LLNL workers 

worked at the Nevada Test Site or were involved with other weapons tests where they could have 
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 6 of 20

received radiation exposures. Personnel dosimeter records are generally available for all periods at 

LLNL for workers who had any potential for occupational radiation exposure. The operations and 

radiation safety staff routinely reviewed dosimeter results for compliance with radiation control limits 

and investigated doses approaching annual or quarterly dose limits. The External Dose 

Reconstruction Implementation Guideline (NIOSH 2002) identified dosimeter records as the highest 

quality records for a retrospective dose assessment.

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 

beginning about 1940 (Morgan 1961; Taylor 1971).

6.2 BASIS OF COMPARISON 

Various radiation dose concepts and quantities have been in use to measure and record occupational 

dose since the initiation of the MED in the early 1940s. The basis of comparison for reconstruction of 

dose is the personal dose equivalent, Hp(d), where d identifies the depth in millimeters and represents 

the point of reference for dose in tissue. Hp(d) is an operational quantity used to determine the 

worker’s actual dose equivalent from a dosimeter reading. International Commission on Radiological 

Protection (ICRP) Publication 74 (ICRP 1996) and International Commission on Radiation Units and 

Measurements (ICRU) Report 57 (ICRU 1998) both define it as a practical method for calibration of 

instruments and dosimeters to dose equivalent. Weakly penetrating radiation is significant to shallow 

dose equivalent, which is defined at d equal to 0.07 mm and noted as Hp(0.07). Penetrating radiation 

is significant to deep dose equivalent, which is defined at d equal to 10 mm and noted as Hp(10). 

Hp(0.07) and Hp(10) are the radiation quantities used in the DOE Laboratory Accreditation Program 

(DOELAP) used to accredit DOE personnel dosimetry systems since the 1980s (DOE 1986). The 

National Voluntary Laboratory Accreditation Program (NVLAP), which is the U.S. Nuclear Regulatory 

Commission equivalent to DOELAP, uses the same operational quantities. This TBD uses Hp(10)

and Hp(0.07) as deep and shallow dose equivalents, respectively.

6.3 DOSE RECONSTRUCTION PARAMETERS

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

geometry of exposure in the workplace, and the characteristics of the LLNL dosimeter responses are 

important to the assessment of bias and uncertainty of the original recorded dose in relation to the 

radiation quantity Hp(d). Dose reconstructors can compare earlier dosimetry systems to current 

systems to evaluate their performance, based on the premise that current systems have more 

stringent criteria, as indicated in the DOELAP and NVLAP programs.

Accuracy and precision of the recorded individual worker doses depend on (Fix et al. 1997):

• Administrative practices

•

that facilities adopt to calculate and record personnel dose based on 

technical, administrative, and statutory compliance considerations

Dosimetry technology

•

, which includes the 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 respective monitoring systems and similarity of the methods of calibration to 

sources of exposure in the workplace
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• Workplace radiation fields

An evaluation of the original recorded doses, as available, combined with detailed examinations of 

workplace radiation fields and dosimeter responses to those fields is the recommended option to 

provide the best estimate of Hp(d) for individual workers.

, which can include mixed types of radiation, variations in exposure 

geometries, and environmental conditions

6.3.1

LLNL used personnel dosimeters to measure and record doses from external radiation to designated 

workers throughout the history of its operations. These dosimeters include one or more of the 

following:

Historical Administrative Practices

• Personnel whole-body (WB) beta/photon dosimeters 

• Pocket ionization chamber dosimeters 

• Personnel extremity dosimeters

• Personnel neutron dosimeters

LLNL began operations in 1952 using dosimeter and processing technical support provided by the 

Lawrence Berkeley National Laboratory (LBNL), which was then known as the University of California 

Radiation Laboratory. LBNL had implemented its dosimetry methods based on the personnel beta/

photon dosimeter design developed at the Metallurgical Laboratory at the University of Chicago 

(Pardue, Goldstein, and Wollan 1944). From 1952 to 1955, LBNL provided LLNL with photon/electron

film dosimeters and nuclear track emulsion, type A (NTA). By 1956, LLNL had its own fully functional 

personnel dosimetry program with in-house processing. 

Table 6-1 summarizes the personnel dosimeters used at LLNL over the years, along with their periods 

of use, exchange frequencies, minimum detection levels (MDLs), and estimated annual missed 

doses. 

Table 6-1. LLNL dosimeter type, period of use, exchange frequency, MDL, and potential annual 

missed dose. 

Period of use Dosimeter

MDLa

(rem)

Exchange

frequencyb Max. annual missed dosec (rem)

1952 - 1969d Photon/electron - DuPont 558 and 519 film 0.015 Weekly (n=50) 0.375

0.020 Biweekly (n=25) 0.25

0.030 Monthly (n=12) 0.18

Neutron - Kodak NTA film 0.050 Weekly (n=50) 1.25

0.050 Biweekly (n=25) 0.625

0.050 Monthly (n=12) 0.30

1969 - 1985 Photon/electron/neutron - Harshaw TLD

(TLD-100, TLD-200, TLD-600, and TLD-700)

0.010 Monthly (n=12) 0.06

0.020 Quarterly (n=4) 0.04

1985 - present Photon/electron - Panasonic 810AS and 802AS 

TLD

0.010 Monthly (n=12) 0.06

0.015 Quarterly (n=4) 0.03

0.025 Semi-annual (n=2) 0.025

Neutron – CR-39 0.010 Monthly (n=12) 0.06

0.010 Quarterly (n=4) 0.02

a. Estimated MDLs for each dosimeter technology in the workplace. Dose values were recorded at levels less than the MDL

b. Exchange frequencies were dependent on work assignment. If the exchange frequency is not evident based on trends in an 

individual’s personnel records, assume a monthly exchange frequency. 

c. Maximum annual missed dose calculated using the MDL/2 method from NIOSH (2002).

d. From 1952 to 1955, LBNL processed film dosimeters. 
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 8 of 20

Dose reconstruction parameters concerning LLNL administrative practices significant to dose 

reconstruction involve policies to:

• Assign dosimeters to workers

• Exchange 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

LLNL policies were in place for all these parameters. From its inception, LLNL had policies to monitor 

individuals with any significant potential for radiation exposure. The current practice of monitoring all 

workers entering the site, regardless of exposure potential, has been in effect since May 1958, when 

film badges became part of the security badge (Nolan 1958). Dosimeter exchange frequencies varied 

over the years depending in part on the dosimeter type in use at the time and in part on the exposure 

potential of the individual being monitored; individuals with low exposure potential tended to have less 

frequent exchange frequencies than those with high exposure potential. Doses below the detection 

limit were either recorded as zero or not recorded at all.

6.3.2

The LLNL dosimetry methods evolved during the years as improved technology was developed and

the complex radiation fields encountered in the workplace were better understood. The adequacy of

the respective dosimetry methods to accurately measure radiation dose as discussed in later sections

depends on radiation type, energy, exposure geometry, etc. The dosimeter exchange frequency of the 

dosimeters was gradually lengthened and corresponded generally to downward reductions in the

radiation protection guidelines (Morgan 1961; Taylor 1971). The dosimeter designs accommodated 

the numerous radiation field types that workers might encounter throughout the LLNL complex. 

Dosimetry Technology

6.3.2.1 Beta/Gamma Dosimeters

From 1952 through 1955, LBNL supplied and processed beta/gamma dosimeters. These film 

dosimeters were similar in design to those developed at the Metallurgical Laboratory at the University 

of Chicago (Pardue, Goldstein, and Wollan 1944). LLNL used a similar design when it began its inhouse dosimetry program in 1955. These film dosimeters, which were in use until 1969, used DuPont 

558 and 519 film.

In 1969, thermoluminescent dosimeters (TLDs) replaced film dosimeters. LLNL constructed the TLDs 

used from 1969 through 1985, which contained Harshaw TLD-100, TLD-200, TLD-600, and TLD-700

elements. The use of Panasonic 802 and 810 dosimeters began in 1985 and has continued to the 

present. 

6.3.2.2 Neutron Dosimeters 

From 1952 through 1955, LBNL supplied and processed neutron dosimeters, which used Kodak NTA 

film. LLNL-provided dosimeters used the same film from 1955 through 1969. At some atomic 

weapons employer facilities or Department of Energy facilities neutron dosimeters were not read out 

unless a minimum gamma exposure threshold was reached. There is no evidence that this was 

practiced at LLNL. In “Personnel Monitoring Procedure for UCRL, Livermore” (Block, 1954) it is 

stated that: “A film badge containing fast neutron sensitive film shall also be provided for persons who 

will be working within areas where significant neutron exposures are possible.” No definition of 
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"significant" was provided; at this time there are no clear criteria for how neutron dosimeters were 

assigned. If available, coworker data should be used. It is clear that all unmonitored workers did not 

have the potential for exposure to neutrons based on buildings worked in and job classifications, but 

the criteria for monitoring has not been quantified, Research into dosimetry practices at Lawrence 

Berkeley National Laboratory to develop the Technical Basis Documents may provided additional 

information on dosimetry practices. This document will be revised if additional evidence is found.

As far as badge exchange and processing the document states that: “For those who work in radiation 

areas, the film badges are exchanged once a week and fresh film is installed. The old film is 

developed and the radiation dosage is recorded.” No where in this document is it mentioned that not 

all neutron film badges were processed. However, only values above the limit of detection were

reported.

In 1969, TLDs replaced film dosimeters. LLNL constructed the TLDs used from 1969 through 1985, 

which contained Harshaw TLD-100, TLD-200, TLD-600, and TLD-700 elements. From 1969 through 

1975, the neutron dosimeter used a non-Albedo-type design. Badge modifications in 1975 used the 

Albedo design whereby cadmium shielded the TLD-600 and TLD-700 components on all sides except 

the side facing the wearer’s body. This design effectively shielded incident thermal neutrons from the 

TLD-600 and TLD-700 elements and detected only thermal neutrons reflected from the wearer’s body. 

The design, which minimized the over-response of the earlier design, was in use until 1985 when 

LLNL adopted Panasonic TLDs. Also in 1985, CR-39 neutron dosimeters were introduced and used 

with the Panasonic TLDs. This system continues in use at present.

6.3.3

Potential error in recorded dose is dependent on the dosimeter response characteristics to each 

radiation type, energy, and geometry; the methodology used to calibrate the dosimetry system; and 

the extent of similarity between the radiation fields used for calibration and the field present in the 

workplace. The potential error is much greater for dosimeters with significant variations in response, 

such as film dosimeters for low-energy photon radiation and the NTA film and TLDs for neutron 

radiation.

Calibration

6.3.3.1 Beta/Photon Dosimeters

Dosimeters were calibrated using 226Ra, 137Cs, or 60Co in air (i.e., no phantom) until the 1986 adoption 

of DOELAP procedures requiring calibration with phantoms. Since 1986, photon calibrations have 

been to 137Cs with a phantom. Beta calibrations were routinely to 90Sr.

6.3.3.2 Neutron Dosimeters

Neutron dosimeters were calibrated using PuBe sources prior to 1970. Since 1970, unmoderated and 

moderated 252Cf has been used.

6.3.4

The radiation fields at LLNL are highly variable. They include radiation from a variety of radiationproducing machines such as electron accelerators, X-ray machines, cyclotrons, neutron generators, 

and a research nuclear reactor. In addition, many different radioactive materials have been used at 

LLNL. Tables 6-2 and 6-3 list many of the radiation sources that might have been encountered at 

LLNL over the years. All beta particle energies are greater than 15 keV. In general, beta and/or 

charged particle exposure around the accelerators is limited to shallow dose. Due to the manner in 

Workplace Radiation Fields
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 10 of 20

which accelerated charged particles were contained, individuals would not have been exposed to 

energetic charged particles. The neutron energy selection of 0.1 – 2.0 MeV indicated in Table 6-3 

was chosen because neutron sources were shielded and moderated to this energy range; the energy 

range of 0.1 – 2.0 MeV has the highest quality factor, as provided in 10 CFR 835.2, which is claimant 

favorable.

Table 6-2. Selection of beta and photon radiation energies and percentages. 

Buildings -

old numbersa Buildings -new numbersa Description

Radiation 

type

Energy 

selection (keV) Percentage

101, 102, 106, 

117, 118, 147, 

176, 192

222, 221, 223, 

224, 234, 232, 

233, 167, 168, 

169

Chemistry: radioactive materials 

including Co-60, fission products, 

enriched uranium, depleted uranium, 

natural uranium, U-233, Cm-244, Pu239, Am-241, others

Beta >15 100

Photon 30–250 

> 250

50

50

153, 154, 157, 

173, 180, 194

210, 212; 171, 

173-177; 241, 

243; 435, 442, 

443; 194

Physics: accelerators, activation 

products, H-3, others

Beta >15 100

Photon 30–250 

> 250

25

75

103, 114, 125, 

127, 174, 175

215, 321, 419, 

514, 243, 253

Lab Services: radioactive materials Beta >15 100

Photon 30–250 

> 250

75

25

110 261 Criticality Test Facility Beta >15 100

Photon 30–250 

> 250

50

50

115 327 Radiography Beta >15 100

Photon 30–250 

> 250

25

75

121 412 Hot cells: high beta waste, Sr-90 Beta >15 100

Photon 30–250 

> 250

75

25

170 131 Weapons Beta >15 100

Photon <30

30–250 

50

50

171 332 Metallurgical Chemistry: AKA 

Plutonium Facility

Beta >15 100

Photon <30

30–250 

50

50

172 331 Gaseous Chemistry: AKA Tritium 

Facility

Beta >15 100

Photon <30 (tritium brems) 100

182 162, 165, 166 Lab Services: 55 Ci Co-60 (1958) Beta >15 100

Photon 30–250 

> 250

25

75

190 251 Chemistry Heavy Elements Facility: 

Cm-244, Am-241, U-233, Pu-239, 

others

Beta >15 100

Photon <30

30–250 

> 250

25

50

25

193 281 Reactor Beta >15 100

Photon 30–250 

> 250

25

75

Site 300 Site 300 Explosives Testing: linear 

accelerators, depleted uranium, H-3, 

radiography

Beta >15 100

Photon 30–250 

> 250

25

75

a. LLNL 2005
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Table 6-3. Selection of neutron radiation energies and percentages. 

Buildings -

old numbersa Buildings -new numbersa Description

Radiation 

type

Energy 

selection (MeV) Percentage

101, 102, 106, 

117, 118, 147, 

176, 192

222, 221, 223, 

224, 234, 232, 

233, 167, 168, 

169

Chemistry: radioactive materials 

including Co-60, fission products, 

enriched uranium, depleted uranium, 

natural uranium, U-233, Cm-244, Pu-239, 

Am-241, others

Neutron 0.1 – 2.0 100

153, 154, 157, 

173, 180, 194

210, 212; 171, 

173-177; 241, 

243; 435, 442, 

443; 194

Physics: accelerators, activation 

products, H-3, others

Neutron 0.1 – 2.0 100

103, 114, 125, 

127, 174, 175

215, 321, 419, 

514, 243, 253

Lab Services: radioactive materials Neutron 0.1 – 2.0 100

110 261 Criticality Test Facility Neutron 0.1 – 2.0 100

115 327 Radiography Neutron 0.1 – 2.0 100

121 412 Hot cells: high beta waste, Sr-90 Neutron 0.1 – 2.0 100

170 131 Weapons Neutron 0.1 – 2.0 100

171 332 Metallurgical Chemistry: AKA Plutonium 

Facility

Neutron 0.1 – 2.0 100

172 331 Gaseous Chemistry: AKA Tritium 

Facility

Neutron 0.1 – 2.0 100

182 162, 165, 166 Lab Services: 55 Ci Co-60 (1958) Neutron 0.1 – 2.0 100

190 251 Chemistry Heavy Elements Facility: 

Cm-244, Am-241, U-233, Pu-239, others

Neutron 0.1 – 2.0 100

193 281 Reactor Neutron 0.1 – 2.0 100

Site 300 Site 300 Explosives Testing: linear accelerators, 

depleted uranium, H-3, radiography

Neutron 0.1 – 2.0 100

a. LLNL 2005

6.3.5

Table 6-4 summarizes the terminology used to describe dosimeter measured and compliance dose 

quantities. 

Recorded Dose Practices

Table 6-4. Recorded dose practices. 

Year Dosimeter measured quantities Compliance dose quantities

Photon/electron film dosimeter + NTA neutron dosimeter

1952 - 1963 Beta (B)

Gamma (G)

Neutron (N)

WB = gamma + neutron + betaa

H = hand extremity dose (HN = hands) 

1963 - 1969 Beta (B)

Gamma (G)

Neutron (N)

WB = gamma + neutron

S = shallow (SK = skin) = gamma + neutron + beta

H = hand extremity dose (HN = hands) 

Photon/electron/neutron - Harshaw TLD

1969 - 1985 Beta (B)

Gamma (G)

Neutron (N)

WB = gamma + neutron

S = shallow (SK = skin) = gamma + neutron + beta

H = hand extremity dose (HN = hands) 

Photon/electron/neutron - Panasonic TLD + CR-39 neutron dosimeter

1985 - present B/P/N Skin = NPEN + WB

WB = PEN + SN + FN

1995-2003 Shallow (SH or SK) 

Deep photon (PH DP) 

Deep neutron (NU DP)

Skin = beta + photon + neutron (B/P/N)

WB = Photon deep + neutron deep

a. Prior to Oct. 18, 1963, beta exposures were recorded as a component of the whole-body dose. 
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 12 of 20

6.3.6

Dose reconstructors can use the information in Table 6-5 to assist in the interpretation of LLNL 

external dosimetry summary reports. Because much of the data prior to 1963 was not adequately 

defined to directly report shallow and neutron doses the dose reconstructor should use the guidance 

found within External Dose Reconstruction (NIOSH 2003a), Interpretation of Dosimetry Data for 

Assignment of Shallow Dose (NIOSH 2005a) and Assignment of Missed Neutron Doses Based on 

Dosimeter Records (NIOSH 2005b.)

Interpretation of Reported Data

Table 6-5. Interpretation of reported data. 

Period

Reported 

quantity Description

Interpretation 

of zeroes

Interpretation 

of blanks 

(no data)

Rollup of 

individual and 

annual data Monitored/unmonitored

1952 - 

1963

rem Reported WB doses 

include gamma and 

beta, sometimes 

qualified with “G” for 

gamma only. Neutron 

doses were 

designated with “N.”

Zeroes were generally 

not reported. Reported 

zero should be 

interpreted as meaning 

less than MDL.

No data or blanks 

should be 

interpreted as 

individual was 

monitored with 

zero result.

Photon WB dose

Neutron WB dose

Shallow skin dose

Total deep WB

dose

All employees with 

significant exposure 

potential were monitored 

(more than 95% of 

employees were 

monitored before 1958). 

After 1958, all employees 

were monitored 

continuously.

1963 -

1985

rem Reported WB doses 

qualified as either “G” 

for gamma or “N” for 

neutron. Beta 

reported with “S” 

and/or “SK” for skin. 

Zeroes were generally 

not reported before 

1980. Reported zero 

should be interpreted as 

meaning less than MDL.

No data or blanks 

should be 

interpreted as 

individual was 

monitored with 

zero result.

Photon WB dose

Neutron WB dose

Shallow skin dose

Total deep WB

dose

All employees were 

monitored continuously.

1985 - 

present 

rem Photon deep, neutron 

deep, and shallow 

dose reported. 

Zeroes were typically 

reported. Reported 

zero should be 

interpreted as meaning 

less than MDL.

No data or blanks 

should be 

interpreted as 

individual was 

monitored with 

zero result.

Photon WB dose

Neutron WB dose

Shallow skin dose

Total deep WB

dose

All employees were 

monitored continuously.

6.4 ADJUSTMENTS TO RECORDED DOSE

Photon Dose

No adjustment to recorded photon doses is recommended. Prior to 1986, dosimeters might have 

been calibrated in air (i.e., no phantom). Dose reconstructors should use roentgen-to-organ dose 

conversion factors (DCFs) when assessing reported photon doses prior to 1986. Since 1986, deep 

dose equivalents at LLNL have been based on DOELAP calibration to Hp(10). Use Hp(10)-to-organ 

DCFs beginning in 1986. Table 6-6 lists adjustments to recorded dose.

Neutron Dose 

To ensure claimant favorability, dose reconstructors can generally consider the neutron energies at 

LLNL to be between 0.1 and 2.0 MeV, for which the ICRP Publication 60 radiation weighting factor is 

20 (ICRP 1990). The associated dose correction factor is 1.91, which is rounded to 2. Dose 

reconstructors should apply this factor to measured neutron dose equivalent and missed neutron dose 

equivalent. See Table 6-6.
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Table 6-6. Adjustments to recorded dose. 

Period Dosimeter Facility Adjustment to reported dose

1952 - 1985 Photon dosimeters All facilities Use roentgen-to-organ dose conversion factors. 

1986 - present Photon dosimeters All facilities Use Hp(10)-to-organ dose conversion factors.

All years Neutron dosimeters All facilities Multiply reported doses by factor of 2 to account for 

ICRP 60 weighting factors (ICRP 1990).

6.5 MISSED AND UNMONITORED DOSE

The potential for missed dose exists when workers are exposed to radiation at levels below the 

detection limit of their personnel dosimeters. In the early years of radiation monitoring, when relatively 

high detection limits are combined with short monitoring durations, missed doses can be significant. 

This section discusses methodologies for estimating missed doses.

Unmonitored dose pertains to the potential dose received by workers who did not wear personnel 

dosimetry. This situation existed at LLNL only before May 1958, when dosimeters were incorporated 

in the security badges and worn by all employees regardless of exposure potential. Prior to May 

1958, workers with little potential for occupational radiation exposure were not generally monitored. 

Watson et al. (1994) examined methods analysts can consider when there is no recorded dose for a 

period during a working career. The 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 (2002) describes options to calculate missed dose for this situation. The preferred 

option estimates a claimant-favorable maximum potential missed dose as MDL/2 multiplied by the 

number of zero-dose results. Table 6-1 lists the results of these calculations. 

Specific incident reports might address significant nonroutine worker doses, such as skin 

contamination events. The dose assessments in such reports, based on investigations conducted at 

the time of the incident, should be the best estimates of dose received.

6.6 BIAS AND UNCERTAINTY

A number of factors contribute to uncertainty in measured doses. Systematic errors can occur from 

calibration and processing as well as from extraneous conditions such as moisture, heat, and fading. 

Random errors arise from variations among workers and the energy spectra and geometries of their 

exposures. This TBD analysis found no specific uncertainty assessments for LLNL dosimeter 

systems. However, the systems have much in common with systems used at other facilities, such as 

the DOE Hanford Site, for which extensive studies have been performed (Wilson et al. 1990; Fix, 

Gilbert, and Baumgartner 1994). The similarities enable reasonable comparisons for the LLNL

systems, based on experience at Hanford and elsewhere.

NIOSH (2002) provides guidance for estimating uncertainty in external dose reconstruction. Under 

good laboratory conditions, film-badge uncertainty can be at the level of 10% to 15%. The absolute 

uncertainty at 95% confidence should not be less than the MDL, which for LLNL was as high as 

0.03 rem for beta/gamma film (Table 6-1). Figure 2.1 of NIOSH (2002) shows the results from two 

methods of calculating the uncertainty factor. In the absence of any other site-specific data, dose 

reconstructors should use numerical values employed for film-badge dose reconstruction. 
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 14 of 20

An uncertainty factor of 1.2 is appropriate for dosimetry results based on the simplified dosimetry 

i=uncertainty modeling for film contained within NIOSH (2002). The uncertainty for TLDs is generally 

smaller than that for film and somewhat less dependent on energy. Dose reconstructors should use 

an uncertainty factor of 1.2 which is claimant-favorable for doses measured with both film and TLDs. 

Relatively little information is available on uncertainty for shallow dose. In view of the similar 

mechanisms between photon and beta film dosimetry, NIOSH (2002) recommends applying the 

methodology described above to the beta dose.

Table 6-7 lists bias and uncertainty factors for LLNL.

Table 6-7. Bias and uncertainty. 

Site-specific dosimetry system

Bias magnitude and range

Overall biasa Range in biasb

Photon/electron film (1952 - 1969) 1.0 0.7 - 1.3

Neutron NTA film (1952 - 1969) 1.0 0.5 – 2.0

Harshaw photon/electron/neutron TLD (1969 - 1985) 1.0 0.8 - 1.2

Panasonic photon/electron TLD (1986 - present) 1.0 0.8 - 1.2

CR-39 neutron (1986 - present) 1.0 0.6 – 1.5

a. Based on the distribution of energy levels and geometry judged most likely. Divide recorded dose by the 

table's bias value to calculate deep dose.

b. Range of overall bias factors based on alternative distributions of energy levels and geometry.

6.7 DOSE RECONSTRUCTION

As much as possible, the basis for dose to individuals should be the dosimetry records. It is important 

to distinguish between the recorded nonpenetrating and penetrating doses and the actual Hp(0.07)

and Hp(10). The following list contains appropriate guidance: 

• Worker dosimetry records that provide nonzero beta-photon values for Hp(10) and Hp(0.07)

are adequate. Dose reconstructors should consider beta energies to be greater than 15 keV 

and, unless known to be otherwise, photon energies to be within the claimant-favorable energy 

range of 30 to 250 keV.

• Radionuclides used in the various buildings and facilities are described in the LLNL site

description (NIOSH 2005c.) These will give the dose reconstructor additional evidence as to 

whether shallow doses are primarily from beta or low energy x-rays if a particular building is 

cited in the workers' history files. If no evidence can be found as to the source of shallow 

dose, a claimant favorable approach should be taken. In general, for most non-skin cancers 

this results in assigning shallow dose to the 30 to 250 keV photons; for skin cancers on 

exposed skin, assigning the dose to beta particles is more client favorable.

• Dose reconstructors should assign missed dose for workers for whom dosimetry records 

provide zero beta-photon values for Hp(10) and Hp(0.07), as identified in NIOSH (2002). This 

approach is conservative based on a review of historical data.

• For unmonitored workers, dose reconstructors should solicit and use coworker data. In 

general, assign the maximum reasonable coworker dose as a claimant-favorable estimate. As

an alternative, use population data.

• For unmonitored workers whose exposure potential has been determined to be low, assign the 

environmental dose.
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 15 of 20

• Prior to 1963, the reported whole body dose is a good estimate of the skin dose, since it 

includes both the penetrating and shallow components (both of which would contribute to skin 

dose). Prior to 1963, the reported “whole body” dose represents an overestimate of the deep 

dose because it included a shallow-only component. The recommendation for pre-1963 

dosimetry records would be to accept the reported whole body dose as both shallow and 

deep. This would be claimant-favorable and reasonable since it will not be possible for a dose 

reconstructor to break the reported whole body dose down into deep and shallow components. 

Also see NIOSH (2005a) for additional guidance on reconstructing non-penetrating dose.

• No evidence could be found to indicate whether elevated ambient levels of external radiation 

were subtracted from workers’ doses. However, because of the LLNL’s facility type it is 

unlikely that there would have been a significant elevated ambient dose (NIOSH 2003b.)

6.8 ORGAN DOSE

NIOSH (2002) discusses the conversion of measured doses to organ dose equivalent, and Appendix 

B of that document contains the appropriate dose conversion factors for each organ, radiation type, 

and energy range based on the type of monitoring performed. In some cases, simplifying 

assumptions are appropriate. For periods when calibrations were performed in free air, prior to 1986, 

dose reconstructors should use the exposure-to-organ DCF. For recorded doses from 1986 to the 

present, use the Hp(10)-to-organ DCF.
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 16 of 20

REFERENCES

Block, S. (University of California Radiation Laboratory), 1954, “Personnel Monitoring Procedure for 

UCRL, Livermore,” Livermore, California

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

Testing of Personnel Dosimetry Systems, DOE/EH-0027, Office of Environment, Safety and 

Health, Washington, D.C.

Fix, J. J., E. S. Gilbert, and W. V. Baumgartner, 1994, An Assessment of Bias and Uncertainty in 

Recorded Dose from External Sources of Radiation for Workers at the Hanford Site, PNL10066, Pacific Northwest Laboratory, Richland, Washington.

Fix, J. J., L. Salmon, G. Cowper, and E. Cardis, 1997, “A Retrospective Evaluation of the Dosimetry 

Employed in an International Combined Epidemiologic Study,” Radiation Protection Dosimetry, 

Volume 74, pp. 39-53.

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

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

England.

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

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

England.

ICRU (International Commission on Radiation Units and Measurements), 1998, Conversion 

Coefficients for Use in Radiological Protection Against External Radiation, Report 57, 

Bethesda, Maryland.

LLNL (Lawrence Livermore National Laboratory), 2005, LLNL Building Index: Pre 1966 and Post 

1966, facsimile to J. J. Maisler (Integrated Environmental Management, Inc.), February 15, 

2005, Livermore, California.

Morgan, K. Z., 1961, Dosimetry Requirements for Protection from Ionizing Radiation. (In:) “Selected 

Topics in Radiation Dosimetry”, International Atomic Energy Agency, Vienna, Austria, pp. 3-

23.

NIOSH (National Institute for Occupational Safety and Health), 2002, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Office of Compensation Analysis and Support, 

Cincinnati, Ohio.

NIOSH (National Institute for Occupational Safety and Health), 2003a, External Dose Reconstruction, 

ORAUT-PROC-0006, Office of Compensation Analysis and Support, Cincinnati, Ohio.

NIOSH (National Institute for Occupational Safety and Health), 2003b, Occupational Dose from 

Elevated Ambient Levels of External Radiation, ORAUT-OTIB-007, Office of Compensation 

Analysis and Support, Cincinnati, Ohio.

NIOSH (National Institute for Occupational Safety and Health), 2005a , Interpretation of Dosimetry 

Data for Assignment of Shallow Dose, ORAUT-OTIB-0017, Office of Compensation Analysis 

and Support, Cincinnati, Ohio
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 17 of 20

NIOSH (National Institute for Occupational Safety and Health), 2005b, Assignment of Missed Neutron 

Doses Based on Dosimeter Records, ORAU-OTIB-0023, Office of Compensation Analysis and 

Support, Cincinnati, Ohio

Nolan, W. E., 1958, “Progress Report on Film Badge Program,” memorandum to D. C. Sewell, 

November 13, 1958, Livermore, California.

Pardue, L. A., N. Goldstein and E. O. Wollan, 1944, Photographic Film as a Pocket Radiation 

Dosimeter, CH-1553-A-2223, Metallurgical Laboratory, Chicago, Illinois.

Taylor, L. S., 1971, Radiation Protection Standards, CRS Press, Cleveland, Ohio.

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

Doses for Workers Without Monitoring Data for Retrospective Epidemiologic Studies,” Health 

Physics, Volume 67, Number 4, pp. 402-405.

Wilson, R. H., J. J. Fix, W. V. Baumgartner, and L. L. Nichols, 1990, Description and Evaluation of the 

Hanford Personnel Dosimeter Program From 1944 Through 1989, PNL-7447, Pacific 

Northwest Laboratory, Richland, Washington.
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Glossary

beta radiation

Radiation consisting of electrons emitted spontaneously from the nuclei of certain radioactive 

elements. Most (if not all) direct fission products emit beta radiation. The beta particle is 

physically identical to an electron moving at high velocity.

curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

deep dose equivalent (Hd) 

The dose equivalent at a depth of 10 millimeters in tissue.

dose equivalent (H) 

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

The special unit is the rem. When D is expressed in grays, H is in sieverts. (1 sievert equals

100 rem.)

dosimeter

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

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

of absorbed dose or dose equivalent received by an individual. (See film dosimeter, neutron 

film dosimeter, thermoluminescent dosimeter.) 

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation.

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, or 

extremities. This includes the fabrication, assignment, and processing of dosimeters as well 

as interpretation and documentation of the results.

film

In the context of this document, a packet that contains one or more pieces of film in a lighttight wrapping. The film when developed has an image caused by radiation that can be 

measured using an optical densitometer. (See nuclear emulsion.)

film dosimeter

A small packet of film within a holder that attaches to a wearer.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus.

minimum detection level (MDL)

The minimum quantifiable dose equivalent that a given dosimetry system can reliably 

measure. 
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 19 of 20

missed dose

The potential dose equivalent that might not have been measured due to the limitation of the 

dosimeter, even though a worker was monitored.

neutron

A basic particle that is electrically neutral and has nearly the same mass as the hydrogen 

atom.

neutron film dosimeter

A film dosimeter that contains a nuclear track emulsion, type A, film packet.

nuclear track emulsion, Type A (NTA)

A film that is sensitive to fast neutrons. The developed image has tracks caused by neutrons 

that can be seen by using an appropriate imaging capability such as oil immersion and a 

1,000-power microscope or a projection capability.

occupational dose

The radiation dose resulting from a claimant’s exposure to radiation and/or to radioactive 

material while working at the LLNL site. Occupational dose does not include exposure to

background radiation, as a patient from medical practices, from voluntary participation in 

medical research programs, or as a member of the general public.

personal dose equivalent Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 millimeter and 

10 millimeters for the skin and body, respectively. These are noted as Hp(0.07) and Hp(10), 

respectively.

photon

A unit or particle of electromagnetic radiation consisting of X- and/or gamma rays.

rad

The traditional unit of absorbed dose (one rad equals 100 ergs per gram of material absorbing 

the radiation energy). The word derives from radiation absorbed dose. 

radiation

Alpha, beta, neutron, and photon radiation.

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

rem

The traditional unit of dose equivalent, which is equal to the product of the absorbed dose in 

rad and the quality factor of the radiation. The word derives from roentgen equivalent in man. 

roentgen

A unit of exposure to gamma (or X-ray) radiation. It is defined precisely as the quantity of 

gamma (or X-) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kilogram of dry 

air. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in 

soft tissue for higher (more than about 100 kilovolts-electron) energy photons.
Document No. ORAUT-TKBS-0035-6 Revision No. 00 Effective Date: 10/07/2005 Page 20 of 20

shallow absorbed dose (Ds) 

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

composition.

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.07 millimeter in tissue.

sievert (Sv)

The International System unit for dose equivalent. (1 sievert equals 100 rem.)

thermoluminescence

Property of a material that causes it to emit light as a result of being excited by heat.

thermoluminescent dosimeter (TLD)

A device used to measure radiation dose. It consists of a holder containing solid chips of 

material that when heated will release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose.

unmonitored dose

The potential unrecorded dose equivalent that could have resulted because an exposed 

worker was not monitored.

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 centimeter (1,000 milligrams 

per square centimeter); however, this term is also used to refer to the recorded dose.

X-ray

(1) Ionizing electromagnetic radiation of external nuclear origin. (2) A radiograph.