ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

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Document Title:

Summary Site Profile for Sandia National 

Laboratories in Livermore, California

Document Number: ORAUT-TKBS-0053

Revision: 00 

Effective Date: 05/01/2007

Type of Document TBD

Supersedes: None

Subject Expert(s): Laura McDowell-Boyer, Janet Johnson, Theresa Lopez, and Randy Whicker

Site Expert(s): N/A

Approval: Signature on File Approval Date: 04/19/2007

Laura McDowell-Boyer, Document Owner

Concurrence: Signature on File Concurrence Date: 04/19/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 04/23/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 04/19/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 05/01/2007

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-0053 Revision No. 00 Effective Date: 05/01/2007 Page 2 of 95

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

05/01/2007 00 Approved new summary site profile for Sandia National Laboratories 

in Livermore, California. No meetings with union members have been 

held to date. Incorporates formal internal and NIOSH review 

comments. Attributions and Annotations section added. There is an 

increase in assigned dose and a PER is required. Training required: 

As determined by the Task Manager. Initiated by Laura McDowellBoyer.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 3 of 95

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ...............................................................................................................7

1.0 Introduction ..............................................................................................................................9

1.1 Purpose........................................................................................................................9

1.2 Scope.........................................................................................................................10

2.0 Site Description......................................................................................................................11

2.1 Purpose and Scope....................................................................................................11

2.2 Introduction.................................................................................................................11

2.3 Site Activities ..............................................................................................................14

2.3.1 Tritium Research Laboratory...........................................................................16

2.3.2 Depleted Uranium Machining..........................................................................17

2.3.3 Radiography, Accelerators, and Neutron Generators......................................18

2.4 Site Processes ...........................................................................................................18

2.4.1 Incidents .........................................................................................................18

2.4.2 Effluents..........................................................................................................19

2.5 Radiological Protection and Access Control ...............................................................19

3.0 Occupational Medical Dose....................................................................................................22

3.1 Examination Frequency..............................................................................................22

3.2 Equipment and Techniques ........................................................................................22

3.3 Organ Dose Estimates................................................................................................23

3.3.1 Parameters and Estimation Method ................................................................23

3.3.2 Organ Dose Estimates....................................................................................25

3.4 Uncertainties ..............................................................................................................26

4.0 Occupational Environmental Dose .........................................................................................29

4.1 Purpose and Scope....................................................................................................29

4.2 Introduction.................................................................................................................29

4.3 Internal Dose From Onsite Atmospheric Radionuclide Concentrations .......................29

4.3.1 Radionuclides of Concern ...............................................................................29

4.3.2 Source Terms for Internal Dose ......................................................................31

4.3.3 Annual Intake of Radioactivity .........................................................................31

4.3.3.1 Inhalation Intakes .............................................................................32

4.3.3.2 Ingestion Intakes ..............................................................................35

4.4 External Exposure To Onsite Atmospheric Radionuclide Concentrations ...................35

4.4.1 Locations of Concern ......................................................................................35

4.4.2 Gamma-Emitting Radionuclides......................................................................36

4.4.3 Neutron Exposure...........................................................................................40

5.0 Internal Dosimetry..................................................................................................................42

5.1 Summary Of Potential Internal Radionuclide Exposures.............................................42

5.1.1 Tritium Research Laboratory...........................................................................42

5.1.2 Uranium Alloy Machining ................................................................................42

5.2 Bioassay Programs ....................................................................................................43

5.2.1 Tritium Bioassay Programs .............................................................................43

5.2.2 Uranium Bioassay Programs...........................................................................44
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5.3 Bioassay Laboratory Procedures And Minimum Detectable Intakes ...........................45

5.3.1 Tritium.............................................................................................................45

5.3.2 Uranium ..........................................................................................................47

5.4 Bioassay Results........................................................................................................48

5.5 Method Of Calculating Dose From Urine Bioassay Data.............................................49

5.6 Airborne Radionuclide Concentrations........................................................................49

5.7 Unmonitored Dose Estimates .....................................................................................49

5.8 Radiation Dosimetry Reports......................................................................................50

5.9 Summary Tables ........................................................................................................51

6.0 Occupational External Dose...................................................................................................53

6.1 Purpose......................................................................................................................53

6.2 Scope.........................................................................................................................53

6.3 Introduction.................................................................................................................53

6.4 Dose Reconstruction Parameters ...............................................................................54

6.4.1 Site Historical Administrative Practices ...........................................................54

6.4.1.1 Administrative Practices Before 1989 ...............................................54

6.4.1.2 Administrative Practices After 1989 ..................................................54

6.4.2 Site Dosimetry Technology .............................................................................56

6.4.2.1 Dosimetry Technology Before 1989..................................................57

6.4.2.2 Dosimetry Technology After 1989.....................................................58

6.4.3 Calibration.......................................................................................................59

6.4.3.1 Calibration for Dosimeter Technologies Before 1989........................59

6.4.3.2 Calibration for Dosimeter Technologies After 1989...........................59

6.5 Dose Calculation And Reporting.................................................................................61

6.5.1 Dose Calculation and Reporting Before 1989 .................................................61

6.5.2 Dose Calculation and Reporting After 1989 ....................................................62

6.5.3 Exposure Energy Spectra ...............................................................................64

6.5.3.1 Neutron Dose Conversion Factors....................................................66

6.5.4 Missed Dose ...................................................................................................66

6.5.5 Organ Dose Conversion Factors.....................................................................67

6.5.6 Uncertainty......................................................................................................67

7.0 Attributions and Annotations .......................................................................................................68

References ........................................................................................................................................69

Glossary ............................................................................................................................................82

ATTACHMENT A, SUMMARY OF ACTIVITIES AT SITE BUILDINGS......... ......................................86

ATTACHMENT B, DOSIMETRY PROGRAM ADMINISTRATIVE

 DOCUMENTATION EXAMPLES..........................................................................89
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LIST OF TABLES

TABLE TITLE PAGE

2-1 Area information and parameters ..................................................................................... 14

2-2 Historical T2 inventory at TRL........................................................................................... 17

2-3 Effluents of 3H to air and sewer during operations and cleanup of TRL ............................ 20

3-1 Entrance kerma by procedure and period......................................................................... 24

3-2 DCFs; absorbed dose for organs at various Al HVL for radiography................................. 25

3-3 Analogs for IREP organs not specified in ICRP (1982)..................................................... 25

3-4 Organ dose estimates for chest X-rays (rem) ................................................................... 26

3-5 Organ doses from lumbar spine X-ray, 1956 to 1971 ....................................................... 27

4-1 Summary of methodology for estimating intakes of 3H, 234U, and 239Pu)............................ 32

4-2 Maximum site-wide annual median inhalation intakes, 1956 to 1970................................ 33

4-3 Maximum site-wide annual median inhalation intakes, 1971 to 2004................................ 34

4-4 External gamma radiation dose based on 2,500 hr/yr exposure duration ......................... 39

4-5 Elevated neutron dose due to LLNL perimeter neutron source, based on 2,500 

hr/yr exposure duration .................................................................................................... 41

5-1 MDAs for and frequency of tritium bioassay ..................................................................... 45

5-2 Preliminary time line for uranium in urine analyses.......................................................... 47

5-3 Range of tritium doses based on bioassay, 1979 to 1981................................................. 48

5-4 Estimated tritium doses based on bioassay, 1982 to 1995 ............................................... 48

5-5 MPCas for tritium and uranium.......................................................................................... 50

5-6 Internal dose control program........................................................................................... 51

5-7 Detection limits for urine bioassay for 3H with liquid scintillation counting.......................... 51

5-8 Detection limits for total urine bioassay ............................................................................ 51

5-9 Solubility type, fraction activity and particle size ............................................................... 52

6-1 Documented dosimetry technologies and dose quantities ................................................ 57

6-2 Circumstances and protocols for dose adjustment or estimation ...................................... 62

6-3 Monthly and quarterly LLDs for the Harshaw Model 8802 dosimeter card ........................ 64

6-4 Selection of radiation energies and percentages .............................................................. 65

6-5 Neutron dose conversion factors for applicable SNL/CA facilities..................................... 66

6-6 Estimated maximum annual missed photon, beta and neutron dose ................................ 66

6-7 Dose units for organ dose conversion factors................................................................... 67

6-8 Bias and uncertainty estimates......................................................................................... 67
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LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Boundaries of SNL/CA property ....................................................................................... 12

2-2 Map of SNL/CA facilities................................................................................................... 13

3-1 Kerma in air at 1 m from X-ray source as a function of total filtration for various 

values of a tube potential.................................................................................................. 28

4-1 Thermoluminescent dosimeter locations since 1989 ........................................................ 37

4-2 Location of LLNL perimeter gamma and neutron dosimeters ........................................... 37

4-3 Location of LLNL offsite gamma dosimeters..................................................................... 38

4-4 Livermore valley radiation background measurements in comparison with annual 

average measurements at LLNL site perimeter and LLNL Site 300.................................. 40

6-1 ALARA reporting action levels.......................................................................................... 55

6-2 Harshaw Model 8812 dosimeter assembly, element specifications, and respective 

radiation, and dose equivalent quantities.......................................................................... 59

6-3 Example Harshaw 8800 card reader calibration procedures, forms, analysis, and 

output...............................................................................................................................60

6-4 Example calibration form for the Harshaw 4000 dosimetry system................................... 61

6-5 Angular dependence testing results for the Harshaw 8800/8812 dosimetry 

system. .............................................................................................................................63

B-1 Example external dosimetry history records (output from SANDOS) ................................ 88

B-2 Example nonroutine dosimeter evaluation form for special radiation work case 

(potential neutron exposures)........................................................................................... 89

B-3 Example nonroutine dosimeter evaluation form for special radiation work case 

(potential neutron exposures)........................................................................................... 90

B-4 Example unreturned dosimeter investigation report.......................................................... 91

B-5 Example unreturned dosimeter questionnaire follow-up letter........................................... 92

B-6 Example portion of unreturned dosimeter questionnaire showing various codes 

used by the Personnel Dosimetry Division........................................................................ 93

B-7 Work Area Radiation Dosimetry Request Form ................................................................ 94

B-8 Example work area radiation dosimetry results report. ..................................................... 95
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ALARA as low as reasonably achievable

AP anterior-posterior

Bq becquerel

CEDE committed effective dose equivalent

CEP Controls for Environmental Pollution

CFR Code of Federal Regulations

Ci curie

cm centimeter

cpm counts per minute

CRDL Chemical and Radiation Detection Laboratory

DAC derived air concentration

DCF dose conversion factor

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ESE entrance skin exposure

g gram

GPS Gas Purification System

gU generic unit

hr hour

HTO tritiated water vapor

HVAH high-velocity air hood

HVL half-value layer

ICP-MS inductively coupled plasma mass spectrometry

ICRP International Commission on Radiological Protection

ICT Insulating Core Transformer (LLNL)

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kVp peak kilovoltage, applied kilovoltage

L liter

LAT lateral

LLD lower limit of detection

LLNL Lawrence Livermore National Laboratory

m meter

MBA Mass Balance Area
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mCi millicurie

MDA minimum detectable activity

MDC minimum detectable concentration

MDL minimum detection limit

MeV megaelectron-volt, 1 million electron-volts

mL milliliter

mrem millirem

MPCa maximum permissible concentration in air

nC nanocoulomb

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NIST National Institute of Standards and Technology

PA posterior-anterior

pCi picocurie

POC probability of causation

PRDP Personnel Radiation Dosimetry Program

R roentgen

RDC Radiation Detection Company

RESL Radiological and Environmental Sciences Laboratory

SID source-to-image distance

SNL/CA Sandia National Laboratories/California

SNL/NM Sandia National Laboratories/New Mexico

SOP Safe Operating Procedure

SPD site profile document

SSD source-to-skin distance

SWP Safe Work Permit

T2 elemental tritium

TBD technical basis document

TIB technical information bulletin

TLD thermoluminescent dosimeter

TMA/EAL Thermo-Analytical Incorporated/EAL Corporation

TRL Tritium Research Laboratory

U-nat natural uranium

VERS Vacuum Effluent Recovery System

yr year

μCi microcurie

μg microgram

§ section or sections
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1.0 INTRODUCTION

1.1 PURPOSE

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] 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 occupationally derived radiation 

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

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

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

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

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

derived:

• Radiation from naturally occurring radon present in conventional structures

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

 

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

This SPD for the Sandia National Laboratories site in Livermore, California (SNL/CA) describes

aspects of the SNL/CA site and historical activities and practices that are pertinent to dose 

reconstruction under the EEOICPA. Section 2.0 describes the site and the activities. Section 3.0

addresses the protocol and procedures associated with routine occupational medical X-rays of 

SNL/CA employees. Estimated occupational intakes associated with above-background ambient 

levels of radionuclides in the SNL/CA environment are provided in Section 4.0, along with estimated 

external ambient dose rates. Sections 5.0 and 6.0 address the technical issues associated with 

measurement of internal and external dose, respectively. 

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

justification, or clarification of the associated information, are presented in Section 7.0.
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2.0 SITE DESCRIPTION

2.1 PURPOSE AND SCOPE

This section briefly describes the physical environment of the SNL/CA site and the site activities and 

processes carried out since its establishment in 1956. The types of radioactive materials present on 

the SNL/CA site, the areas in which exposure to radioactive materials may have occurred, and the 

emissions of radionuclides to the environment are identified. Access control for radioactive areas is 

also described.

2.2 INTRODUCTION

The background information in this introductory section was taken primarily from Ullrich (2003), which 

was prepared to support DOE’s compliance with the national Historic Preservation Act. The SNL/CA 

site was established in 1956 to provide direct support for Lawrence Livermore National Laboratory 

(LLNL) nuclear weapons designs. The primary mission during the time coinciding with the Cold War 

(1956 to 1989) was the design and testing of non-nuclear components of nuclear weapons designed 

by LLNL. SNL/CA was to engineer, or “weaponize,” the nuclear physics packages designed by LLNL; 

production of parts and final weapons was accomplished at other weapons complex sites. 

Sandia Corporation, a Lockheed Martin Company, currently operates SNL/CA and Sandia National 

Laboratories/New Mexico (SNL/NM) in Albuquerque. From 1956 to 1993, SNL/CA was managed and 

operated by American Telephone and Telegraph. In 1993, the contract was awarded to Martin 

Marietta Corporation, now known as Lockheed Martin Corporation (DOE 2003).

The SNL/CA site presently consists of approximately 70 buildings/facilities on 410 acres and is 

located just across East Avenue from the LLNL in Livermore, California. Figure 2-1 shows the 

SNL/CA facilities in relation to the boundaries; Figure 2-2 shows the individual buildings/facilities on 

the site. 

The SNL/CA site initially consisted of a long narrow strip of 50 acres stretching south from East 

Avenue. A personnel building was the first building to be completed in September 1957; the rest of 

the original buildings (warehouse, model shop, environmental test, and central steam plant, office and 

laboratory building) were completed in 1958. Design support for LLNL was originally provided by a 

small group of engineers and support staff. Initially, 14 SNL/NM employees worked with LLNL in 

LLNL facilities. In 1957, SNL/CA began using the LLNL Site 300 for explosive testing. By 1958, with 

over 800 employees, SNL/CA worked on the W38 warhead for Titan I and Atlas missiles. Support for 

LLNL also expanded to include effects test analyses and telemetry for LLNL nuclear tests of nuclear 

weapon designs. As part of the Plowshare Program, which ended in the early 1970s, SNL/CA also 

moved into evaluation of nuclear detonation in 1959 (Ullrich 2003).

The environmental test building for the new SNL/CA site (completed by the end of 1958) was used for 

some testing of new designs. During the 1960s, more support facilities were built, including storage 

and a maintenance shop. Additional test facilities (the centrifuge and Explosive Test Facility) were 

also constructed. In 1970, 86 acres were added to the site, providing an additional buffer area.

Although SNL/CA has never lost their core mission of nuclear ordnance design and testing, the 

laboratory moved further into scientific research in the late 1960s, bringing in scientists, 

mathematicians, and materials specialists to work in applied research. SNL/CA’s initial move into

research on 3

H grew out of its familiarity with, and use of, 3H in components. In 1974, the first 
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Figure 2-1. Boundaries of SNL/CA property (SNL/CA 2005). 

structure dedicated to 3

H research at SNL/CA, the Tritium Research Laboratory (TRL), was added to 

the site, with completion of the basic laboratory building during the summer of 1975 (Garcia and 

Gorman 1996). The TRL became operational in late 1978.

Ranging further from weapons engineering, SNL/CA also pursued combustion research and, during 

the 1973 to 1974 energy crisis, began conducting some research into alternative energy. The 

expanded purpose of SNL/CA resulted in an increased variety of facilities within the site. In addition 

to the TRL, a large complex for the Combustion Research Facility was completed in 1980. An 

additional 24 acres of land was added as a buffer zone on the east side near the Tritium Facility in 

1979. In 1986, an additional 228 acres was obtained, allowing an alternative exit route from the 

facility. Finally, in 1998, SNL/CA took part in a small land exchange to create a consistent buffer zone 

line along the western boundary, in which 2.82 acres were received in exchange for 5.41 acres. This 

brought the site’s area down to 410 acres where it remains today. In 2004, there were 1094 

employees at the site (SNL/CA 2005).
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Figure 2-2. Map of SNL/CA facilities (Ullrich 2003). 
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2.3 SITE ACTIVITIES

In support of the various missions identified in Section 2.2, a number of activities were undertaken at 

SNL/CA over the years, some of which involved handling and release of radioactive materials. A 

complete list of buildings along with known information about present and past uses and the presence 

of radioactive materials are given in Attachment A. A subset of this building list can be alternatively 

compiled to represent the major process complexes at the SNL/CA site that handled radioactivity in 

some manner over the years (as shown in Table 2-1) and the predominant radiologically related 

activities that took place in these complexes. 

Table 2-1. Area information and parameters.

Area Weapons Laboratory Facility Complex

Description Building 910, 912, 913, 914, 916, 918

Time Period 1958–1998

Activities Test/repair neutron detectors, wet machining of DU, radiography of weapons 

components, radiography for materials science studies, 3H storage studies, ion beam 

analysis of materials, Radiflo leak tests

Radionuclides DU, 3H, neutron generator, small accelerators, small sealed sources, small amounts of

85Kr

Area Radiography

Description Building 923

Time Period unknown–Early 1990s

Activities radiography using X-rays, gamma rays, neutrons, alpha and beta particles

Radionuclides 60Co, 192Ir, 252Cf, X-ray machines

Area Micro and Nano Technologies Laboratories

Description Buildings 941, 942, and 943

Time Period unknown–present

Activities radiography for materials science studies

Radionuclides In Building 941 only: X-ray, 238U, and beta sources (sealed)

Area Former Tritium Research Laboratories (Currently the Chemical and Radiological 

Detection Laboratory)

Description Buildings 967, 968, and 969

Time Period 1974–1996

Activities 3H research: >0.1 g 3H handled in glovebox, 0.0005–0.1 g in high velocity air hoods

Radionuclides 3H, DU (no radionuclides after decommissioning complete in 1996)

Area Explosives and Environmental Testing Complex

Description Buildings 955, 956, 966, 972, 974, 976, 977, 978, 979, 981, 983

Time Period 1958–present

Activities environmental testing of mock-up weapons and components

Radionuclides DU

Area Storage Facilities

Description Buildings 921, 927, 961, 982

Time Period unknown–present

Activities storage and packaging of waste materials

Radionuclides 3H, DU, natural Th, Trace 239Pu and mixed fission products

According to SNL/CA annual environmental reports dating back to 1983, the laboratory typically 

handled kilogram amounts of depleted uranium (DU), gram amounts of 3

H (when the TRL was 
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operational), and only microcurie quantities of other isotopes (SNLL 1983a). However, the 1992 

Environmental Impact Statement for LLNL and SNL/CA noted 100 Ci 192Ir and 60Co radiographic 

isotopes and many other smaller sealed sources with activity ranging from 1 μCi to 500 mCi, stored in 

a shielded radiography cell (DOE 1992).

DU is, and has been, largely in the form of alloyed metal components, often encapsulated. Wet 

machining of uranium metal did occur over the years (Adolphson 1972, SNL/CA 1989a, Wallace 

1988). Powdered DU sealed in storage containers for 3

H-storage studies (as the tritide) has also 

been present in approximately 1-kg amounts (SNL/CA 1991a). A classified activity involving 

powdered uranium hydride in gram amounts was conducted in gloveboxes in buildings 979, 916, and 

the TRL (Wright 2006). There were some thorium metal parts used on test systems as well (Wright

2006). 

Tritium was generally handled in the TRL in the form of a gas, although effluents could be in the form 

of a gas, liquid, or solid. Mixed solid and liquid tritiated waste was generated; the majority of which 

was in the form of scintillation cocktails, which were shipped offsite for incineration in Florida (DOE 

1992). Details regarding total quantities of 3

H present for each year of operation of the TRL and 

pertinent to types of research activities carried out in this facility are described in Section 2.3.1 below.

Most of the remaining radioactive materials were contained in sealed sources. However, a small 

amount of 85Kr was used in the Radiflo leak detection studies. The period of use of the Radiflo units is 

not known. 

In 1982, the Final Environmental Impact Statement for LLNL and SNL/CA (DOE 1982) indicated that 

the major activities taking place on the SNL/CA site were related to “tritium research; arming, fusing 

and firing systems; and aerodynamic and structural elements used in U. S. nuclear bombs and 

warheads.” At that time, it was found that operations within most facilities on site have no significant 

environmental impact and they handle radionuclides in small enough quantities such that special 

containment features and operating procedure are not required to assure that no radioactivity is 

released. The only exceptions to this finding of no potential significant impact were the SNL/CA 

radioactivity storage vaults, one of which was located in Building 927, and the other small vault within 

the TRL (Building 968), and the TRL itself. It was noted that the contents of the 927 vault varied 

according to work needs, but generally included depleted uranium-containing mock-ups of weapons, 

and small, sealed radioactive sources not in use. Tritium was not stored in the 927 vault; rather the 

small quantities present on site were kept in the laboratories where they were being used in Building 

968, or in the vault in that building. Other than the H-3 source due to vaults in Building 968 (the TRL), 

there were experiments conducted with up to 120 g (1.2 × 10 6 Ci) of 3

H per experiment in sealed 

gloveboxes. 

In 1992, the Final Environmental Impact Statement and Environmental Impact Report for Continued 

Operation of Lawrence Livermore National Laboratory and Sandia National Laboratories (DOE 1992) 

concluded that the only building for which a potentially significant radiological release could occur 

under accident conditions would be Building 968, which housed the TRL at that time. This was based 

on a review of all potential radiological sources and the likelihood that the source could become 

dispersed into the onsite or offsite environment. 

The 2003 Final Site-wide Environmental Assessment of the Sandia National Laboratories/California

(DOE 2003) addressed, in part, the potential environmental impact of continuing operations at the 

SNL/CA. The TRL operations were discontinued in 1996 (including the decontamination activities); 

thus, the radiological activities at this time involved radioactive material management of legacy 
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radioactive material inventories and current nuclear material inventories, including radioactive 

isotopes used in laboratory research and radiation monitoring activities. 

A Waste Management Site Plan from 1980 states: 

The radwastes comprise, primarily, tritium and depleted uranium (D-38) but no 

Transuranics, fission products or induced activity. Now that the Tritium Research 

Laboratory is operational, tritium contamination will account for more than 50% of the 

radwaste volume in FY80. Depleted uranium contaminated wastes will account for most 

of the remainder. D-38 wastes are mostly machine turnings from the Machine Shop in 

Bldg. 913 and components from the Test Assembly Group also in Bldg. 913. 

Contaminated paper and machine lubricants make up most of the actual D-38 waste 

volume. D-38 components may be classified or unclassified. (Wright 1981a) 

The information provided above suggest that activities involving DU and 3

H are the only two 

potentially-significant sources of internal exposure to workers based largely on the fact that these are 

the only two radioactive materials present in significant quantities throughout the history of SNL/CA. 

External exposures to radiation from radiation-generating devices (small accelerators, radiography 

sources, neutron generators, X-ray machines) are also of concern. The remainder of Section 2.3 will 

be devoted to aspects of the SNL/CA site pertinent to these sources of exposure.

2.3.1 Tritium Research Laboratory

The TRL was designed as a modern research and development facility to provide support to the DOE 

weapons complex (Garcia and Gorman 1996). The TRL was the first major 3

H research and 

development complex to use secondary containment coupled with a cleanup system as a means to 

control personnel exposure to levels as low as reasonably achievable (ALARA) and reduce 

environmental releases (Garcia and Gorman 1996). The TRL continued to operate until 1993 when 

all tritium and associated research was transferred to other DOE facilities. The TRL was 

decontaminated from January 1994 through October 1996 and now houses the Chemical and 

Radiation Detection Laboratory (CRDL).

Administrative controls limited the amount of tritium in the research laboratory to a maximum of 120 g 

at any time (excluding the vault) and to a maximum of 300 g at the facility in total at any time (Wall 

1981). Table 2-2 lists the historical inventory of elemental tritium (T2) at TRL (from Garcia and 

Gorman 1996).

The TRL was divided into 9 laboratories, 2 office areas, a 3

H storage room, a control room, a shop, a 

3

H gas purification equipment room, and 8 miscellaneous rooms. The Radioactive Materials Area was 

separated from the rest of the facility by two sets of double doors. The ventilation system directed 

airflow from clean areas to areas of increasing contamination potential (Wright 1981a). 

Operations in the TRL depended upon the types of research ongoing at the facility but were focused 

on the physical and chemical characterization of 3

H and its compounds, the fabrication of tritium 

compounds for use as engineering components (Wright 1981a), and examining the behavior of 

hydrogen isotopes and helium in metals to understand transport and structural properties (Garcia and 

Gorman 1996). Operations were performed inside sealed gloveboxes for any experiments involving 

more than 0.1 g 3

H (DOE 1992). 

Building 968 was not operated as a security exclusion area. Building access was controlled by a sign 

in/sign out procedure until a computer-based security system was put in place, which is known to 

have been in place in 1991 (SNL/CA 1991b). Radiological materials were also dispensed through 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 17 of 95

one authorized person and amounts were administratively controlled. The Nuclear Material 

Management Group had three Mass Balance Areas (MBAs) and a custodian for each area was 

Table 2-2. Historical T2 inventory at TRL.

Year T2 (grams) T2 (Bq)

1979 1.06 3.8E+14

1980 15.47 5.5E+15

1981 29.62 1.1E+16

1982 28.55 1.0E+16

1983 27.62 9.8E+15

1984 37.86 1.3E+16

1985 85.49 3.0E+16

1986 107.58 3.8E+16

1987 174.35 6.2E+16

1988 181.59 6.4E+16

1989 148.89 5.3E+16

1990 115.46 4.1E+16

1991 117.85 4.2E+16

1992 131.67 4.7E+16

1993 101.21 3.6E+16

1994 29.43 1.0E+16

1995 9.96 3.5E+15

1996 4.04 1.4E+15

1997 0 0.0E+00

responsible for issuing of nuclear materials to personnel in his MBA. Material was accounted for to

the nearest 0.01 gram, with amounts equal to or greater than 0.005 g rounded to the nearest 0.01 g.

Two central decontamination systems were used: (1) Gas Purification System (GPS) and (2) Vacuum 

Effluent Recovery System (VERS). The GPS was used to remove tritium, tritiated water, and tritiated 

hydrocarbons from the sealed glovebox atmosphere in the event of either a significant release or a 

slow buildup of background contamination. The VERS was used to remove tritium, tritiated water

vapor (HTO), and, tritiated hydrocarbons from the glovebox pressure control system and the gases 

exhausted from all of the vacuum pumps in the laboratory before venting this gas to the stack.

 

2.3.2 Depleted Uranium Machining

Machine shops in Buildings 913, 914, and in an annex of Building 918 have been used for machining 

of DU fairly infrequently at SNL/CA. A 1972 memorandum from Adolphson (1972) indicates that 

“machining of uranium alloy test specimens” began in 1971. No machining has been done since 

1998, when Building 913 was decommissioned. According to the Health Physicist at SNL/CA, this 

machining was done as a wet process; with the exception of one accident he was aware of where dry 

cutting occurred (Wright 2006). The 1989 Safe Operating Procedure (SOP) governing machining of 

uranium metal (SNL/CA 1989a) dictates that all machining operations shall be performed wet and that 

dry operations are prohibited. According to a 1985 DOE Headquarters appraisal of SNL/CA and 

response to findings (Wallace 1988), uranium machining occurs approximately once or twice a 

quarter. The appraisal/response report also indicates that the Hazards Control Division at SNL/CA, 

as of 1987, had 16 yr of air sampling data for this procedure, which “demonstrate that airborne 

contamination is not a problem during typical machining operations.” 
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2.3.3 Radiography, Accelerators, and Neutron Generators

Until the early 1990s, the Radiography Facility was a 3880 sq ft building (Building 923) which included 

X-ray machines, gamma ray source, neutron sources, and many alpha and beta sources (DOE 1992). 

The radiation sources were primarily used for the radioscopy and electron imaging of weapon and 

non-weapon components. The radiation sources included the radiation-producing machines with 

energies up to 420,000 volts; the 100 Ci 192Ir and 60Co radiographic isotopes; sealed 252Cf 

spontaneous fission neutron source located inside a massive neutron source shield; and many other 

smaller sealed neutron, alpha, beta, and gamma ray isotope sources of low activity (one μCi to 500 

mCi), which are stored in a shielded radiography cell. The building contained four shielded 

radiography cells. The sources were used infrequently for material characterization studies involving 

technologies such as radiation transmission gauging, backscatter measurements, X-ray fluorescence 

studies, and neutron activation analysis (DOE 1992). 

X-Ray diffraction equipment is reported as being present in building 913, room 115, as early as 1965 

(SNL/CA 1964). An incident that occurred with a diffractometer (see Section 2.4.1) was reported to 

have occurred in building 913, room 113 in 1979 (Lovell 1980). Radiography is currently conducted in 

Building 941, part of the Micro and Nanotechnologies Laboratory (DOE 2003). 

Two small accelerators have been, and remain, located in Building 916 (DOE 1992, 2003). These 

include a 1 MeV Tandem accelerator and a 700/200 keV positive ion accelerator (SNLL 1983b, 

SNL/CA 1991c, Wright 2006). 

Neutron generators used in the design, assembly, testing, calibration, and repair of neutron detectors 

at SNL/CA are located in Building 910 (SNL/CA 1990a). Neutron generator tests are also conducted 

in Building 974 (Attachment A). 

2.4 SITE PROCESSES 

2.4.1 Incidents

According to Garcia and Gorman (1996), a few incidents occurred with regard to the TRL which may 

have lead to 3

H exposures in excess of routine operations. Other sources (Lovell 1980, SNL/CA 

1960-1983) listed excess exposure to ionizing radiation, although only one was attributable to a workrelated exposure. These incidents are listed below:

Personnel Exposure to Ionizing Radiation, December 1979

During set-up of a high-temperature diffractometer attachment to port 2 of the X-ray unit tube tower, a 

worker was accidentally exposed to X-rays from port 1. This resulted in a dose assignment of 26 to 

34 rem to the basal cell layer of a skin area (less than 400 cm2

) of the chest and arm (Lovell 1980).

Personnel Exposure and Tritium Release, July 1984

During disassembly of an engineering experiment in Laboratory 115A, an environmental release of 

approximately 2.5 Ci of HTO occurred. Seven employees received doses ranging from a high of 

1,650 mrem down to 2 mrem.

Tritium Release, January 1986

A weld crack in a storage container caused a leak and release in the Decontamination Laboratory

115A causing an environmental release of 200 Ci of HTO. There were no resulting personnel 

exposures. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 19 of 95

Personnel Exposure and Tritium Release, August 1987

An 1100 Ci environmental release occurred when an operator disassembled a vessel outside a 

glovebox in Laboratory 115. This release of elemental tritium gas (T2) caused a building evacuation 

and the operator received a dose of 15 mrem.

Tritium Release, October 1988

An environmental release occurred during a GPS Regeneration operation in Laboratory 115A. The 

release was estimated at 124 Ci of HTO and no personnel exposure was involved.

Personnel Exposure and Tritium Release, March 1989

A personnel exposure of ~180 mrem occurred during VERS pump maintenance operations in 

Laboratory 120. An environmental release of approximately 11.5 Ci of HTO occurred. Subsequent 

operation required the flushing of pumps before pump maintenance operations.

Personnel Exposure and Tritium Release, October 1993

Six personnel received doses from 2 mrem to 5 mrem when a VERS pump failed and caused a tritium

release in Laboratory 115A. An environmental release of 2 Ci HTO was reported by Garcia and 

Gorman (1996), but was alternatively reported as a 30 Ci stack release by Garcia (994a). The

personnel doses came from both cleanup and pump replacement operations.

2.4.2 Effluents

The only radionuclides released due to normal operations from SNL/CA are small amounts of tritium 

and trace amounts of DU (SNLL 1982, 1983a, 1984; SNL/CA 1986 to 1988, 1989b, 1990b, 1991d, 

1992, 1993, 1994a, 1995 to 2005). The DU effluents are reported to be less than 10 μCi ([3.7 × 10-5 

Bq] for both liquid and airborne effluents collectively [SNLL 1982, 1983a, 1984; SNL/CA 1986]). In 

1990, DOE (1990) reported that emissions of radioactive particulates from SNL/CA activities that 

potentially generate such particulates are controlled with high-efficiency particulate air (HEPA) filters. 

A total of 33 HEPA filters, installed in four buildings (913, 916, 961, and 979) were in use at that time. 

Tritium from the TRL has been the only airborne effluent routinely reported, and is often declared to 

be the only detectable effluent (SNL/CA 1986 to 1988, 1989b, 1990b, 1991d, 1992, 1993, 1994a, 

1995 to 1997). Monitoring of tritium effluents from the TRL ceased in 1996 after the facility was 

transitioned for other uses (SNL/CA 1998).

Table 2-3 lists the measured effluents during the operational and decontamination periods of the TRL

(Garcia and Gorman 1996). These quantities include the incidental releases noted in Section 2.4.1 

above.

2.5 RADIOLOGICAL PROTECTION AND ACCESS CONTROL

Badging and bioassay programs were carried out throughout the history of operations at SNL/CA to 

provide information on exposure to workers. This information was used to limit annual exposures to 

workers to within exposure guidelines set forth by the U.S. Atomic Energy Commission (AEC), Energy 

Research and Development Administration, and DOE criteria (Kingsley 1968; SNL/CA 1994b, 

SNL/CA –undated a). When workers traveled to other locations, such as across the street to the 

LLNL site, the LLNL Site 300, or other non-SNL-operated sites, the host sites were to be requested to 

report exposures at those locations to the appropriate staff at SNL/CA or SNL/NM (SNL/CA - undated 

b). A response to the DOE site appraisal finding related to offsite dosimeters indicated that SNL/CA 

directed employees to “request internal and/or dosimetry, workplace monitoring, etc., if they must 

enter radiation controlled areas at the visited facility” in 1985 (Wallace 1988). 
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Dosimetry records for SNL/CA employees were transferred to the SNL/NM site in the late 1980s to 

the early 1990s (Hallman 1989, 1990; Perez-Romo 1994). In a 1994 memo (Perez-Romo 1994), it is 

indicated that dosimetry data on computer tapes from Reynolds Electric and Engineering Company 

(the contractor providing base support at the Tonopah Test Range where SNL/Tonopah employees 

worked) were also sent for storage at SNL/NM. It does not appear that the pre-1989 records for 

SNL/CA were incorporated into the currently accessible electronic database at SNL/NM for SNL 

employees. Dosimetry data covering the period 1973 – 1987 was recently recovered from SNL/NM, 

and data for 1988 was recovered from SNL/CA. Cumulative summary data covering the period prior

to 1973 has also been recovered from SNL/CA. However, annual data for all years prior to 1973 is 

currently being sought.

All SNL/CA workers were required to wear dosimetry badges between 1959 and 1969, according to a 

collection of annual Summary of Whole-Body Radiation Exposures to ionizing radiation reports

(SNL/CA 1958-1978). In 1958, the radiation exposure summary report indicates that 41 employees 

(of 537) were not monitored. However, a 1960 internal memo (SNL/CA 1958-1961) indicates that all 

employees were monitored from the time of establishment of the Sandia Livermore Corporation. It 

was noted in 1984 (Lovell 1984a, b) that Building 923 and Room 100 of Building 916 had sources 

and/or equipment capable of producing radiation levels that could exceed 500 mrem/yr (which was 

10% of the Radiation Protection Standards for workers at that time). Thus, it was stated (Lovell 

1984a,b) that radiation dosimeters would be required for access to these buildings during radiographic 

procedures and during accelerator operations. In 1965, DeSelm (1965) wrote a memorandum 

indicating that Buildings 911 (Medical), 913, 914, 916, 921, 9143, and Areas 8 and 9 could produce 

levels exceeding AEC’ Manual Chapter 0524 (AEC 1963) criteria of 6 mR/hr for radiation exposure to 

the whole body or critical organs for wearing dosimeters. 

Table 2-3. Effluents of 3

H to air and sewer during operations and cleanup of TRL. 

Year

Total tritium in stack 

discharges to air (Bq) Estimated HTO to air (Bq)

Total tritium (as HTO) in 

wastewater discharges to 

sewer (Bq)

1979 2.2E+11 Not estimated 1.5E+08

1980 9.3E+11 Not estimated 1.5E+08

1981 1.6E+12 Not estimated 5.7E+09

1982 7.5E+12 Not estimated 1.1E+10

1983 3.5E+12 2.7E+12 1.4E+10

1984 6.1E+12 5.4E+12 1.4E+10

1985 1.9E+13 1.4E+13 7.4E+10

1986 2.7E+13 2.3E+13 9.3E+08

1987 6.8E+13 2.1E+13 8.7E+09

1988 5.8E+13 3.8E+13 1.7E+10

1989 3.1E+13 2.4E+13 1.1E+10

1990 1.1E+13 9.0E+12 7.4E+09

1991 1.7E+13 1.3E+13 4.6E+09

1992 9.8E+12 5.0E+12 2.3E+09

1993 7.0E+12 4.9E+12 2.5E+09

1994 3.5E+12 3.4E+12 2.2E+09

1995 2.7E+12 2.7E+12 8.9E+08

1996 2.9E+09 Not estimated 0.0E+00

Other than requiring dosimeters for access to certain facilities at SNL/CA, there are other forms of 

access control in place. The 700/200 keV accelerator (Room 104, Building 916) and 1 MeV tandem 

accelerator were fitted with interlocks to prevent entry during operation (SNL/CA 1991c, SNLL 1983b). 

Access was also restricted to the machine shop (Room 119) in Building 913 during uranium 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 21 of 95

machining operations by roping off the area and having supervisors present during all such operations 

(SNL/CA 1989a). For the TRL (Building 968), building access was controlled by a sign in/ sign out 

procedure until a computer-based security system was put in place, as noted in Section 2.3.1.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 22 of 95

3.0 OCCUPATIONAL MEDICAL DOSE

As part of the requirements for employment at SNL/CA starting in 1956, some employees received 

periodic physical examinations. These could include annual radiographic examinations of the chest, 

as well as lumbar spine X-rays and lateral (LAT) chest X-rays at hire. Because these examinations 

were required for employment, the External Dose Reconstruction Implementation Guidelines (NIOSH

2002) and 42 C.F.R. pt. 82, require the X-ray doses to be part of the occupational radiation exposure. 

This section of this SPD discusses medical screening X-rays required as a condition of employment; it 

does not include diagnostic and therapeutic exposures that were not required for employment. 

The following sections describe the methodology used to estimate absorbed dose from X-ray

exposure for SNL/CA workers. Section 3.1 describes X-ray examination frequency at SNL/CA as 

reconstructed from claimant files. Section 3.2 provides information on equipment and techniques 

used at SNL/CA, including assumptions necessitated by lack of protocol, measurement, or records 

data. Section 3.3 provides organ dose estimates by calendar year and type of X-ray. Section 3.4

documents uncertainties.

3.1 EXAMINATION FREQUENCY 

Protocol for the frequency and type of X-ray examinations for SNL/CA workers from 1956 through the 

present time has not been located. A protocol for the frequency of chest X-ray examinations as a 

function of job category has also not been located and equipment type is not known. However, 

claimant files available at the time of this documents preparation generally indicated that a single 

posterior-anterior (PA) chest X-ray examination was performed at hire, annually, and possibly at 

termination from 1956 through the 1980s. They also showed that anterior-posterior (AP) and LAT 

lumbar spine X-rays were at least sometimes performed at hire as late as 1971. LAT chest X-rays 

were rarely taken; only two files (of over 50 files available at the time) indicated that LAT chest X-rays 

had been taken at hire. LAT chest X-rays should not be included in the default estimates for SNL/CA 

employees unless it is noted in their medical files that they were taken.

According to the health physics department, worker X-rays ceased in the 1980s, but no specific date 

is known. The X-ray machine equipment was removed from the site in approximately 1990 (Wright 

2006). No evidence of the use of photofluorography has been found. It is recommended that annual 

chest X-rays be assumed through 1989 until documentation of the date of cessation of X-rays is 

found.

3.2 EQUIPMENT AND TECHNIQUES

The analysis assumed that radiological practices followed standards of medical practice to minimize 

dose to the worker; however, the type of equipment, technique factors, and some machine 

calibrations are not known. Medical records did record the beam current (mA), applied kilovoltage 

(kVp), and distance (presumably from machine to image) for PA chest examinations and occasionally 

for lumbar spine X-rays. The notation of mA in many of the records is likely the exposure (mAs), as 

the setting recorded in claim files (5 mA) is too low to have produced chest radiographs [1]. 

Additionally, the distance is assumed to be the source to image distance (SID) as it does not vary in 

the worker files reviewed (it is consistently 72 inches). However, exposure time, filtration, and 

entrance skin exposure (ESE) were not noted, nor was the use of screens or grids if applicable. A

medical X-ray unit was present on the site; personnel report that the X-ray machine was removed in 

about 1990 and was not replaced. In the 1950s, the pre-employment set of X-rays may have been 

performed at SNL/NM or off the site in California. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 23 of 95

This section of the SPD provides organ dose estimates from occupational X-ray examinations 

administered at SNL/CA from 1956 through 1969, 1970 to 1985, and post-1985 using calculated sitespecific estimates with input from Technical Information Bulletin [TIB]: Dose Reconstruction from 

Occupationally Related Diagnostic X-ray Procedures (ORAUT 2005a). For the years before 1970, the 

default values from ORAUT 2005a assume minimal beam collimation and a half-value layer (HVL) of 

2.5 mm Al. For 1970 to 1985, the default values assume that the beams were collimated and the HVL 

was 2.5 mm Al. For post-1985, the default values assume collimation and an HVL of 4.0 mm Al. 

These HVLs were used in the site-specific calculations as site-specific information was not available.

For all periods, the analysis should assume that a single PA chest X-ray occurred at hire, at each 

annual physical examination, and at termination of employment, as evidenced by available medical 

records. AP and LAT lumbar spine X-rays were taken once, at hire, for the years 1956 to 1971 [2]. 

Most workers are assumed to have received these X-rays. LAT chest X-rays were rarely taken, but 

have been included here for those energy employees with LAT chest X-rays (taken at hire or at a 

regular physical examination) noted in their medical files (two claim files with this X-ray view have 

been found). Dose reconstructors should assign dose from the X-ray procedures listed on the preemployment X-ray record form. If the claim file is missing the pre-employment X-ray record, the dose 

reconstructor should assign dose from a PA chest and AP and LAT lumbar spine as the default for the 

pre-employment X-rays.

Efforts will continue to locate related SNL/CA X-ray protocol information. Until that is located, 

information from claim files related to SNL/CA have been reviewed and notations regarding machine 

settings (specifically mAs, kVp, and SID) have been used to calculate site-specific estimates of dose 

received from occupationally required X-ray examinations. The X-ray machine has been assumed to 

be single phase [3]. 

3.3 ORGAN DOSE ESTIMATES

This section discusses organ dose estimates. Section 3.3.1 describes the methodology used to 

estimate these doses and Section 3.3.2 discusses results.

3.3.1 Parameters and Estimation Method

ICRP (1982) guidance uses the following parameters to estimate air kerma and absorbed dose:

• Source to image distance (SID) in centimeters (cm)

• Total filtration (millimeters of aluminum, mmAl)

• Estimate of tissue thickness (AP and LAT)

• Machine settings (mA, exposure time [seconds], kVp, film size, and machine type)

If measured air kerma are available, these should be used. For SNL/CA, air kerma was estimated 

from Figure 3-1 (ICRP 1982) assuming a single-phase machine was used at SNL/CA. Assumptions 

(from medical records within claim files) were used to estimate air kerma for chest X-rays for pre1970, 1970 to 1985, and post-1985 time periods. For the pre-1970 and 1970 to 1985 periods, 80 

kVp, 5 mAs, a total filtration of 2.5 mmAl, and a source to subject distance (SSD) of 153 cm results in 

an air kerma estimate of 0.06 mGy per mAs at 1 meter. Total filtration of 2.5 mmAl is a reasonable 

estimate for those time periods (ORAUT 2005a). The entrance kerma is then calculated to be 0.13 

mGy, as shown in Table 3-1. The LAT chest entrance kerma is 2.5 times the PA chest entrance 

kerma (ORAUT 2005a).
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 24 of 95

For post-1985, 110 kVp, 300 mA and 1/30 second exposure time, and SID of 183 cm (from claim 

files), with a total filtration estimate of 2.5 mmAl and HVL of 4.0 (ORAUT 2005a) result in an estimated 

air kerma of 0.10 mGy/mAs and calculated entrance kerma of 0.43 mGy for PA chest X-rays (Table 3-

1). The LAT chest X-ray entrance kerma is 2.5 times the PA chest entrance kerma (ORAUT 2005a).

For lumbar spine X-rays, an applied kilovoltage of 75 kVp and beam current of 75 mAs were assumed 

for AP lumbar spine X-rays and 85 kVp and 150 mAs for LAT lumber spine X-rays based on 

information in claim files. A total filtration of 2.5 mmAl and HVL of 2.5 mmAl were assumed (ORAUT 

2005a), resulting in an estimated air kerma of 0.045 mGy/mAs for AP lumbar spine X-rays and 0.070 

mGy/mAs for LAT lumbar spine X-rays. A SSD of 63 cm for AP and 52 cm for LAT X-rays was used, 

resulting in entrance kermas of 8.5 mGy and 38.8 mGy for AP and LAT X-rays, respectively.

Entrance kerma for PA and LAT chest X-rays and for AP and LAT lumbar spine X-rays are presented 

in Table 3-1. Dose conversion factors (DCFs) are listed in Table 3-2.

Table 3-1. Entrance kerma by procedure and period.

Period

PA chest 

entrance kerma 

(cGy) 

LAT chest 

entrance kerma, 

(cGY)

AP lumbar spine 

entrance kerma 

(cGY)

LAT lumbar spine

entrance kerma

(cGy)

Pre-1970 0.013a 0.0325a 0.85b 3.9c

1970–1985a 0.013 0.0325 NA NA

Post-1985d 0.043 0.1075 NA NA

a. Based on 80 kVp and 5 mAs, as observed in SNL/CA claim files and 2.5 mmAl total filtration (ORAUT 2005a). LAT 

chest entrance kerma is 2.5 times PA chest entrance kerma. PA entrance kerma = (0.06 mGy/mAs) x (5 mAs) x 

(100cm/153cm)2 = 0.13 mGy. 

b. Assumes 75 kVp and 75 mAs, as observed in SNL/CA claim files, and total filtration of 2.5 mmAl (ORAUT 2005a). 

Entrance kerma = (0.045 mGy/mAs) x (75 mAs) x (100cm/63cm)2 = 8.5 mGy.

c. Assumes 85 kVp and 150 mAs, as observed in SNL/CA claim files, and total filtration of 2.5 mmAl (ORAUT 2005a). 

Entrance kerma = (0.070 mGy/mAs) x (150 mAs) x (100cm/52cm)2 = 38.8 mGy.

d. Based on 110 kVp, 300 mA and 1/30 second exposure time (noted in one claim file) and total filtration of 2.5 mmAl 

(ORAUT 2005a). LAT chest entrance kerma is 2.5 times PA chest entrance kerma. PA entrance kerma = (0.10 

mGy/mAs) x (10 mAs) x (100cm/153cm)2 = 0.43 mGy. 

The International Commission on Radiological Protection (ICRP) tables used to estimate absorbed 

dose (ICRP 1982) do not include all the organs included in the Interactive RadioEpidemiological 

Program (IREP) computer program. For organs in IREP but not identified in the ICRP tables, the 

dose conversion coefficient that is anatomically closest to the IREP-specified organs can usually be 

used to estimate dose. For example, the factor for lung can be applied to all other organs in the 

thoracic cavity, such as the esophagus and bone surface. For abdominal organs (bladder, colon), the 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 25 of 95

Table 3-2. DCFs (mGy per Gy air kerma); absorbed dose (1 mGy) for organs at 

various Al HVL for radiography (ORAUT 2005a; ICRP 1982).a

Organ

Lumbar spine DCFs Chest DCFs

Pre-1970 

(2.5-mmAl HVL)

Pre-1970 

(2.5-mmAl HVL)

Pre-1970 

(2.5-mm Al HVL)

1970–1985

(2.5-mm Al HVL)

Post-1985

(4.0-mm Al HVL)

LAT AP LAT PA LAT PA LAT PA

Thyroid 0.01 0.3 137 174b 115 32 164 78

Eye/brain 0.01 0.3 137 32 115 32 164 78

Ovaries N/Ac N/Ac N/A N/A 0.6 1 2.5 5.2

Liver/gall bladder/spleen 14d 79d 220 451 220 451 351 674

Urinary bladder N/Ad N/Ad N/A N/A 0.6 1 2.5 5.2

Colon/rectum N/Ad N/Ad N/A N/A 0.6 1 2.5 5.2

Testes N/Ac N/Ac N/A N/A 0.1 0.01 0.1 0.01

Lungs (male) 14 79 193 419 193 419 313 628

Lungs (female) 14 79 220 451 220 451 351 674

Thymusd 14 79 220 451 220 451 351 674

Esophagusd 14 79 220 451 220 451 351 674

Stomachd 14 79 220 451 220 451 351 674

Bone surfacesd 14 79 220 451 220 451 351 674

Remainderd 14 79 220 451 220 451 351 674

Female breast 13e 25e 255 49 255 49 343 116

Uterus 31 287 N/A N/A 0.6 1.3 2.1 5.2

Bone marrow (male) 22 37 37 92 37 92 76 178

Bone marrow (female) 22 37 29 86 29 86 59 172

Skinf 1.35f 1.35f 1.35f 1.35f 1.35f 1.35f 1.40f 1.40f

a. DCFs for lumbar spine are from ICRP 1982 and DCFs for chest are from ORAUT 2005a.

b. Per ORAUT (2005a), DCF for AP cervical spine corrected for depth by 0.2. 

c. N/A = not applicable; organ dose values for the testes and ovaries for lumbar spine reflect actual measurements reported in Lincoln and Gupton 

(1958).

d. Using analogs listed in Table 3-4.

e. DCFs for lumbar spine examination not given in ICRP (1982). Values for the respective upper gastrointestinal examinations were used instead.

f. Backscatter factor from NCRP 102 (1989, Table B-3) (ORAUT 2005a); see Table 3-5. 

dose coefficient for ovaries is used. This approach should be either favorable to the claimant or 

neutral. Table 3-3 lists analogs for IREP organs, as originally presented in ORAUT (2005a). 

Table 3-3. Analogs for IREP organs not specified in ICRP (1982). 

Anatomical 

location

ICRP #34 

reference organ IREP organ analogsa

Thoracic cavity Lung Thymus, esophagus, stomach, bone surface, liver/ 

gall bladder, remainder organs

Abdominal cavity Ovaries Urinary bladder, colon/rectum

Head and neck Thyroid Eye/brain

a. ORAUT 2005a 

3.3.2 Organ Dose Estimates

Table 3-4 lists calculated organ dose estimates from PA and LAT chest X-ray examinations for each 

period (pre-1970, 1970 to 1985, and post-1985), although LAT chest X-rays were discontinued after 

1970. The estimates for exposure from chest X-rays for these periods have been calculated using 

information taken from claim files in conjunction with assumptions from ORAUT 2005a. 

Table 3-5 presents dose from AP and LAT lumbar spine X-rays that should be used for years 1956 to 

1971. These X-rays were often taken at hire.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 26 of 95

Table 3-4. Organ dose estimates for chest X-rays (rem). 

Organ

Pre-1970

estimated dosea,b,c

HVL = 2.5 mm Al 

(uncollimated)

1970–1985

estimated dosea,b,c

HVL = 2.5 mm Al 

(collimated)

Post-1985

estimated dosea,b

HVL = 4.0 mm Al 

(collimated)

LAT PA LAT PA LAT PA

Thyroid 4.38E-03 2.26E-03 3.68E-03 4.16E-04 1.75E-02 3.35E-03

Eye/brain 4.38E-03 4.16E-04 3.68E-03 4.16E-04 1.75E-02 3.35E-03

Ovaries 1.3E-02c 2.5E-02c 1.92E-05 1.30E-05 2.68E-04 2.24E-04

Liver/gall bladder/

spleen

7.04E-03 5.86E-03 7.04E-03 5.86E-03 3.76E-02 2.90E-02

Urinary bladder 1.3E-02c 2.5E-02c 1.92E-05 1.30E-05 2.68E-04 2.24E-04

Colon/rectum 1.3E-02c 2.5E-02c 1.92E-05 1.30E-05 2.68E-04 2.24E-04

Testes 2.5E-03c 5.00E-03c 3.20E-06 1.30E-07 1.07E-05 4.30E-07

Lungs 7.04E-03 5.86E-03 7.04E-03 5.86E-03 3.76E-02 2.90E-02

Thymus 7.04E-03 5.86E-03 7.04E-03 5.86E-03 3.76E-02 2.90E-02

Esophagus 7.04E-03 5.86E-03 7.04E-03 5.86E-03 3.76E-02 2.90E-02

Stomach 7.04E-03 5.86E-03 7.04E-03 5.86E-03 3.76E-02 2.90E-02

Bone surfaces 7.04E-03 5.86E-03 7.04E-03 5.86E-03 3.76E-02 2.90E-02

Remainder 7.04E-03 5.86E-03 7.04E-03 5.86E-03 3.76E-02 2.90E-02

Female breast 8.16E-03 6.37E-04 8.16E-03 6.37E-04 3.67E-02 4.99E-03

Uterus 1.3E-02c 2.5E-02c 1.92E-05 1.69E-05 2.25E-04 2.24E-04

Bone marrow 1.18E-03 1.20E-03 1.18E-03 1.20E-03 8.13E-03 7.65E-03

Skind,e 4.32E-02 1.76E-02 4.32E-02 1.76E-02 1.50E-01 6.02E-02

a. SID = 183 cm.

b. Image receptor size 35.6 cm by 43.2 cm.

c. Modified from Webster and Merrill (1957) as presented in ORAUT (2005a).

d. Calculated using backscatter factor of 1.35 from NCRP 102 (1989, Table B-3) for skin dose estimates through 1985, 

consistent with ORAUT (2005a). 

e. Calculated using backscatter factor of 1.40 from NCRP 102 (1989, Table B-3) for skin dose estimates after 1985, 

consistent with ORAUT (2005a).

3.4 UNCERTAINTIES

As stated in ORAUT (2005a), error is defined as deviation from the correct, true, or conventionally 

accepted value of a quantity, and uncertainty is defined in terms of the potential range of a stated, 

measured, or assumed or otherwise determined value of a quantity. Error and uncertainty provide an 

indication of confidence in the dose estimates. Uncertainty, expressed in terms of a confidence level, 

is a more appropriate term than error, which implies that the actual value is known. Uncertainty, 

stated as a probability of falling within a stated range, includes precision and reproducibility of the 

measurement as well as accuracy (i.e., how close the estimate comes to the actual value).

Although many factors can introduce uncertainty and error into X-ray exposures, five factors 

contribute the most uncertainty to the dose estimate: (1) measurement error, (2) variation in applied 

kilovoltage, (3) variation in beam current, (4) variation in exposure time, and (5) SSD. Film speed, the 

use of screens, or the use of grids would not affect the beam output intensity. The lack of historical 

records for some of these measurements introduces uncertainty into the dose estimates that cannot 

be readily quantified, although there is no apparent reason to believe that practices at SNL/CA or its 

medical subcontractors were different from those at other facilities or from recommended standards of 

the medical community at the time. The following estimates of uncertainty associated 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 27 of 95

 Table 3-5. Organ doses from lumbar spine X-ray, 1956 to 1971. 

Organ

Estimated dosea,b

HVL = 2.5 mm Al (collimated)

LAT (rem) AP (rem)

Thyroid 3.88E-05 2.55E-04

Eye/brain 3.88E-05 2.55E-04

Ovaries 1.52E+00c 1.12E+00c

Liver/gall bladder/spleen 5.43E-02 6.72E-02

Urinary bladder 1.52E+00c 1.12E+00c

Colon/rectum 1.52E+00c 1.12E+00c

Testes 1.12E+01c 5.40E-02c

Lungs 5.43E-02 6.72E-02

Thymus 5.43E-02 6.72E-02

Esophagus 5.43E-02 6.72E-02

Stomach 5.43E-02 6.72E-02

Bone surfaces 5.43E-02 6.72E-02

Remainder 5.43E-02 6.72E-02

Female breast 5.04E-02 2.13E-02

Uterus 1.20E-01 2.44E-01

Bone marrow 8.54E-02 3.15E-02

Skind 5.39E+00 1.18E+00

a. SSD = 63 cm for AP X-ray and 52 cm for LAT X-ray.

b. Image receptor size 35.6 cm by 43.2 cm.

c. Organ dose values for the testes and ovaries (and analogs) for 

lumbar spine reflect actual measurements reported in Lincoln and 

Gupton (1958). 

d. Skin dose values include backscatter factors of 1.39 from 

Table B.8 of NCRP 102 (1989).

with X-ray exposures are from ORAUT (2005a), which this analysis relied on for default information 

when site-specific records of X-ray machine settings could not be found (specifically, for filtration and 

HVL values). Other values were taken from available notes in claim files and were selected to be 

favorable to claimants whenever a range of values was found in the claim files.

ORAUT (2005a) reports that X-ray doses are derived largely from actual measurements of X-ray

machine output with R-meters or similar ionization chamber devices. Reportedly, these typically had 

an uncertainty of +2% for photon energies below 400 keV if properly calibrated and used. Although 

more current machinery could have a smaller uncertainty, +2% is assumed to be conservative.

Variation in applied voltage generally falls within +5% of the machine setting. Beam intensity is 

approximately proportional to the 1.7 power of the kilovoltage, resulting in an uncertainty of 

approximately +9% in relation to beam intensity for voltages in the 110- to 120-kVp range. Variations 

in tube current are normal and generally small. As the tube current drops, beam intensity falls in 

direct proportion. Large decreases in beam output would be readily detectable and would indicate the 

need for machine maintenance or, as a temporary measure, an increase in the current or voltage to 

provide the necessary intensity for proper radiography. ORAUT (2005a) estimates the variation in 

tube current to be approximately + 5% for this parameter.

Exposure time can significantly affect the dose received from radiography (exposure times are a 

fraction of second). Even a small variation in exposure time due to timer error can significantly 

change beam output. Because early X-ray machine timers are known to have been inaccurate, 

ORAUT (2005a) assume uncertainty in beam output due to timers to be +25%. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 28 of 95

SSD can contribute to variability because the ESE is determined by this distance. Variations result 

from accuracy of positioning as well as worker size (thickness). As expressed in ORAUT (2005a), this 

is generally thought to vary by no more than a few centimeters, with an upper limit of 7.5 cm (±10%). 

A potential source of uncertainty for SNL/CA is the number and type of X-rays taken. As noted above, 

reports indicate the performance of only an annual PA chest X-ray examination, but no official 

protocol has been found that would rule out the possibility of other X-ray views or more frequent chest 

examinations. At this time, dose reconstructors should assume a single annual PA chest X-ray for 

1956 through 1989 [4]. Dose reconstructors should assign dose from the X-ray procedures actually 

listed on the pre-employment X-ray record form. If the claim file is missing the pre-employment X-ray 

record, the dose reconstructor should assign dose from a PA chest and AP and LAT lumbar spine as 

the default for the pre-employment X-rays. 

Another source of uncertainty is the lack of site-specific values for filtration and exposure time. While 

the claim files show a consistent notation of kVp and SID, interpretation was required to determine the 

beam current exposure before 1985. For all time periods, Figure 3-1 was required to estimate air 

kerma. Filtration and HVL were not noted in claimant files so the estimated values used in this SPD 

were favorable to claimants and consistent with practices during each time period.

Consistent with ORAUT (2005a), this analysis relies on the statistical root mean square to estimate 

total uncertainty. The root mean square is the square root of the sum of the squares of the individual 

uncertainty values and equals 28.9% assuming all variation is in the positive direction. An estimate of 

30% uncertainty is favorable to claimants.

Figure 3-1. Kerma in air at 1 m from X-ray source 

as a function of total filtration for various values of 

tube potential (ICRP 1982).
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 29 of 95

4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

4.1 PURPOSE AND SCOPE

This section addresses the occupational environmental dose applicable to the SNL/CA site from the 

beginning of operations (1956) to the present time. The term occupational environmental dose refers 

to the radiation dose received outside of buildings, but on the SNL/CA site, as a result of ambient 

airborne radionuclides or ionizing radiation. 

4.2 INTRODUCTION

The SNL/CA site is located directly to the south of LLNL. The southern perimeter fence of LLNL is 

directly across the street from the northern perimeter fence of SNL/CA. Thus, environmental dose is 

addressed in terms of potential exposures that may occur on the SNL/CA site as a result of both 

SNL/CA and LLNL operations. Aside from operations at the TRL, the internal exposures on the 

SNL/CA are assumed to be due to LLNL operations alone. According to SNL/CA (1998), no other 

measurable effluents of radionuclides from SNL/CA have existed. The external dose rate at the LLNL 

south perimeter, measured by TLDs, has been reported from 1967 to the present. These 

measurements are also considered in estimating external dose for SNL/CA. 

Internal and external exposures to radionuclides in the outdoor environment are considered 

separately in this section. Section 4.3 presents information necessary to estimate internal 

environmental dose; radionuclides of concern are first identified. The estimated source terms (release 

rates) for radionuclides considered potentially significant to internal environmental dose and internal 

exposure (yearly intake) are also addressed. Annual environmental reports for SNL/CA and LLNL 

form the basis of these estimates (LLL 1972 to 1982, 1984; LLNL 1986 to 2005; LRL 1961a,b, 1963 to 

1971; SNL/CA 1986 to 1988, 1989b, 1990b, 1991d, 1992, 1993, 1994a, 1995 to 1998, 2005; SNLL 

1982, 1983a, 1984). 

Section 4.4 contains information necessary for estimating external environmental dose. Ambient 

external dose rates, reported in annual environmental reports for LLNL and SNL/CA, are used to 

estimate dose rates for the operational period of the SNL/CA. 

Section 4.5 considers uncertainties in the information provided for estimating occupational 

environmental dose. The discussion addresses sources of uncertainty and provides quantitative 

information where possible. 

4.3 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

4.3.1 Radionuclides of Concern

The only detectable effluents from the SNL/CA site occurred during operations of the TRL, when 3H 

was released (Sect. 2.4.2). Releases of DU from the SNL/CA site are not considered a significant 

source of environmental exposure at the SNL/CA. Such releases would only occur during machining 

operations, but according to SNL/CA (1990b), all operations at SNL/CA using DU were equipped with 

exhaust air systems with absolute filters at that time. Furthermore, machining operations, which 

began in 1971 (Adolphson 1972), were monitored closely for airborne contamination; operations were 

immediately stopped if airborne contamination was detected (SNL/CA 1989a). 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 30 of 95

However, the presence of measurable concentrations of tritium and airborne particulate radionuclides 

in the LLNL south perimeter area, nearly adjacent to the north perimeter area of SNL/CA, requires 

that these radionuclides be considered as potentially significant contributors to environmental dose on 

the SNL/CA site. In addition to 3

H, LLNL has processed and handled a number of radionuclides, 

including uranium and transuranic elements, mixed fission products, and accelerator-produced 

isotopes. Tritium, 239Pu, and isotopes of uranium have been identified as radionuclides of significance 

for the ORAUT Environmental Occupational Dose Technical Basis Document (TBD) for LLNL 

(ORAUT 2005b), consistent with the list of radionuclides identified by the LLNL environmental 

monitoring program as representing more than 90% of the LLNL radioactive materials inventory.

Before considering any or all of the particulate radionuclides from the LLNL site as isotopes of 

significance at SNL/CA with respect to environmental dose, the concentrations at the LLNL perimeter 

nearest to SNL/CA and the potential associated doses were first evaluated. Between 1961 and 1970, 

concentrations of gross alpha and gross beta were measured at one or more perimeter locations near 

the south side of the LLNL site, as well at several offsite locations in the Livermore valley (LRL 1961a, 

1963 to 1970). Isotope-specific data were not reported during this period. The data reported included 

contributions from naturally occurring alpha- and beta-emitting radionuclides. By comparing average 

off-site measured concentrations of gross alpha- and beta-emitting airborne particulate radionuclides 

with those measured at the southern perimeter, net concentrations of these particulates were 

calculated. The maximum net beta concentration during this period was 3 × 10-2 Bq/m3

; the maximum 

net alpha concentration was 5.9 × 10-5 Bq/m3

. When not zero (i.e., when the concentrations at the 

perimeter were not less than average measured off-site concentrations), committed organ doses 

associated with the alpha- and beta-emitting particulate radionuclides were calculated, assuming an 

annual inhalation rate of 2400 m3

, and that the alpha-emitting radionuclides were comprised wholly of 

234U, and the beta-emitting radionuclides were comprised wholly of 90Sr (a fission product). Dose 

factors were taken from ICRP (2001) and were selected on the basis of which organ received the 

highest dose (i.e., the maximum organ dose was calculated). The representative radionuclides were

selected based on their relatively higher organ dose factors compared to other possible representative 

radionuclides, such that the analysis is not likely to overlook potentially significant contributors to 

internal dose. A comparison of the gross-alpha and gross-beta dose calculations indicated that the 

organ dose associated with beta-emitting radionuclides generally contribute less than 5% of the 

committed dose, but always contribute less than 1 mrem to the yearly committed organ dose 

associated with environmental exposures for this time period. The committed doses associated with 

gross alpha for a 1-year intake ranged from 0 to 140 mrem, the latter based on the assumption that 

the “gross alpha” intake could be represented by 234U.

From 1971 to the present, concentrations of particulate 239/240Pu, 235U, 238U, and gross beta-emitters 

have been measured at a south LLNL perimeter location directly across the street from SNL/CA (LLL 

1972 to 1982, 1984; LLNL 1986 to 2004). Tritium was measured by LLNL at this location since 1973. 

At this location (labeled CAFÉ in LLNL annual environmental reports), the maximum net beta-emitter 

inhalation dose is estimated to be 0.03 mrem/yr, assuming that 90Sr is the representative radionuclide. 

The maximum concentration of 239

Pu between 1971 and 2004 was 2.6 × 10-6 Bq/m3, which 

corresponds to an inhalation intake of 6.3 × 10-3 Bq/yr, assuming a yearly inhalation rate of 2400 

m3

/yr. The maximum committed organ dose from this annual intake is estimated as 1.1 mrem. At the 

same location, the maximum concentrations of 235U and 238U were 1.6 × 10-12 g/m3 and 2.2 × 10-10 

g/m3, respectively. These concentrations correspond to annual intakes of 3.1 × 10-4 Bq/yr and 6.4 × 

10-3 Bq/yr of 235U and 238U, respectively. Making the favorable to claimant assumption that these 

intakes are both of 234U; the corresponding committed maximum organ dose for a one-year intake is 

6.7 mrem.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 31 of 95

These measured concentration data and dose calculations indicate that particulate airborne betaemitting radionuclides originating from LLNL are not significant contributors to environmental 

occupational dose on the SNL/CA site. This is consistent with the finding in LLNL environmental 

reports, indicating that the gross beta activity is due to global fallout, and fluctuates in a manner 

typical of that source (LLL 1973 to 1982, 1984; LLNL 1986 to 2005). The alpha-emitting radionuclides 

234U and 239Pu (represented as gross alpha for pre-1971), from LLNL, are retained as potentially 

significant contributors to internal environmental dose at SNL/CA. Inhalation intakes for these 

radionuclides are addressed in Section 4.3.3 below.

Tritium releases from the LLNL operations (Peterson 2005) and from the TRL are also considered as 

potentially significant contributors to intakes. Intakes are estimated according to the methods 

described in Section 4.3.3. 

4.3.2 Source Terms for Internal Dose

No onsite environmental sources or outdoor concentrations of radionuclides have been reported for 

SNL/CA before the beginning of operation of the TRL in 1979. Release of radionuclides from 

activities at LLNL may have affected intakes by workers at SNL/CA, as noted in Section 4.3.1 above. 

The assumption that the nearest perimeter concentrations at the LLNL site represent the onsite 

concentrations at SNL/CA before operation of the TRL is favorable to the claimant, as a decrease in 

air concentration with distance from the LLNL perimeter would likely occur. From April through 

September the prevailing winds are from the west and southwest and variable throughout the 

remainder of the year (LLNL 1988); thus, this assumption is most favorable to the claimant during 

these months.

4.3.3 Annual Intake of Radioactivity

Inhalation intakes of airborne particulate 234U and 239Pu and of airborne 3H are estimated in this 

section. Ingestion of drinking water contaminated by 3

H is also considered. 

In order to calculate inhalation intake, it was necessary to consider the onsite air concentrations of 

234U, 239Pu, and 3

H due to the presence of these isotopes near the LLNL south boundary (closest to 

the SNL/CA). Between 1979 and 1996, it was also necessary to consider on site 3

H concentrations 

resulting from TRL emissions.

Intakes were calculated by multiplying the relevant concentrations by an assumed inhalation rate of 

2400 m3

/yr. For 3H, the intakes also reflect skin absorption, as the inhalation intake was multiplied by 

a factor of 1.5. For airborne particulates (i.e., 234U and 239Pu), no particle size information was 

available, and thus the default ICRP 66 (ICRP 1994) value of 5 μm activity median aerodynamic 

diameter is recommended. Furthermore, no solubility information is available for airborne uranium or 

plutonium particulates, and thus assumed solubility should be selected based on what is most 

favorable to the claimant in light of the organ of interest. Tritium should be assumed to be associated 

with HTO, as this will give the highest dose associated with the two forms (T2 and HTO) known to be 

released from both LLNL and SNL/CA (Garcia and Gorman 1996; Peterson 2005).
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 32 of 95

4.3.3.1 Inhalation Intakes

The following methodological information is summarized in Table 4-1. Calculated intakes are 

provided in Tables 4-2 and 4-3. 

Table 4-1. Summary of methodology for estimating intakes of 3

H, 234U, and 239Pu).

Radionuclide Applicable period Method used

3

H 1956–1972 Multiply LLNL 3H release (Peterson 2005) by average ratio 

of south perimeter 3H concentration to release, derived 

from 1973-2003 data, to obtain estimate of maximum 

SNL/CA concentration (Bq/m3

). Multiply maximum 

concentration by 2400 m3

/yr inhalation intake rate, and by 

1.5 to account for skin absorption.

1973–1978 and 

1997–2004

Reported south LLNL perimeter 3H concentrations are 

used directly to estimate maximum SNL/CA concentration 

(Bq/m3

). Multiply maximum concentration by 2400 m3/yr 

inhalation intake rate, and by 1.5 to account for skin 

absorption.

1979–1996 Reported south LLNL perimeter 3H concentrations are 

used directly to estimate LLNL contribution to maximum 

SNL/CA concentration (Bq/m3

). TRL contribution (onsite) 

derived by multiplying reported TRL releases by the 

maximum calculated ratio of near-TRL air concentrations 

to TRL releases for the years 1994 and 1995. Multiply 

both perimeter and near-TRL concentrations by 2400 

m3

/yr inhalation intake rate, and by 1.5 to account for skin 

absorption. Add intakes at LLNL perimeter to intakes near 

TRL to account for both SNL/CA and LLNL contributions.

234U 1956–1960 No date available for this time period.

1961–1970 Reported south LLNL perimeter gross alpha concentration 

measurements were corrected for average Livermore 

valley background contributions, and multiplied by an 

inhalation intake rate of 2400 m3

/yr to estimate maximum 

SNL/CA intakes (Bq/yr).

1971–2004 Reported south LLNL perimeter 235U and 238U activity 

concentrations (Bq/m3) were multiplied by an inhalation 

intake rate of 2400 m3

/yr to estimate maximum SNL/CA 

intakes (Bq/yr). The unreported 234U concentrations were 

addressed by assuming 234U intake was equivalent to 238U. 

Activity intakes for all radionuclides are added, and 

assumed to be 234U. 239Pu 1956–1960 No data available for this time period.

1961–1970 Reported south LLNL perimeter gross alpha concentration 

measurements were corrected for average Livermore 

valley background contributions, and multiplied by an 

inhalation intake rate of 2400 m3

/yr to estimate maximum 

SNL/CA intakes (Bq/yr).

1971–2004 Reported south LLNL perimeter 239Pu activity 

concentrations (Bq/m3) were multiplied by an inhalation 

intake rate of 2400 m3

/yr to estimate maximum SNL/CA 

intakes (Bq/yr).

3

H Intakes

Measured concentrations of 3

H at the south perimeter of the LLNL are available in LLNL annual 

reports from 1973 to 2004 (LLL 1973 to 1982, 1984; LLNL 1986 to 2004). These values are used to 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 33 of 95

estimate environmental intakes of 3

H for SNL/CA workers during these years without considering 

additional dilution in transport to the SNL/CA site, thus providing an estimate that is favorable to the 

claimant. Two LLNL annual environmental reports were not located (for 1982 and 1984); for these 2 

yr, the 3

H concentrations were estimated to be the average of the preceding and following year’s 

concentrations.

 

Table 4-2. Maximum site-wide annual median inhalation intakes 

(Bq/yr), 1956 to 1970.

a 

Year 3Hb,c 234U or 239Pud

1956 1.2E+04 (e)

1957 4.0E+04 (e)

1958 2.1E+04 (e)

1959 1.2E+04 (e)

1960 1.1E+04 (e)

1961 3.9E+03 1.2E-01

1962 1.5E+04 1.4E-01

1963 2.1E+04 2.7E-02

1964 9.3E+04 6.2E-02

1965 1.3E+06 0.00f

1966 6.0E+04 0.00f

1967 2.3E+04 0.00f

1968 2.4E+04 0.00f

1969 2.6E+04 8.9E-03

1970 1.0E+06 0.00f

a. Assumes an inhalation rate of 2,400 m3/yr.

b. Assumes total 3H intake is the sum of inhalation and skin absorption, estimated 

by multiplying the inhalation intake by a factor of 1.5. 

c. Intake of 3

H is based on historical LLNL source terms from Peterson (2005); 

see Section 4.3.3.1 and Table 4-1.

d. Gross alpha is reported for 1961 to 1970; recommend assuming the isotope of 

either 234U or 239Pu that gives highest dose to the organ of interest.

e. No measurements reported for these years. 

f. A zero value indicates the measured concentration was less than or equal to 

the off-site background gross alpha concentration.

Before 1973, 3H concentrations at the LLNL perimeter were not reported in annual reports. For these 

years (1956 to 1972), concentrations are estimated by multiplying the average ratio of perimeter air 

concentration of 3

H to reported annual LLNL 3H released for the years 1973 through 2003 (found to be 

2.7 × 10-14) multiplied by the reported 3

H release for each year from 1956 to 1972. Tritium releases 

from LLNL are estimated back through 1953 by Peterson (2005). Tritium intakes from 1956 through 

1970 are reported in Table 4-2. Intakes derived using this methodology for 1971 and 1972 are 

included in Table 4-3. After 1972, but for years when the TRL was not operating (1973 to 1978 and 

1996 to 2004), the intakes were calculated based on the LLNL south perimeter concentrations and 

are given in Table 4-3. 

Between 1979 and 1996, the environmental 3

H concentrations on the SNL/CA site were due to LLNL 

releases as well as the TRL releases. Measurements on the SNL/CA site near the TRL (where 3

H 

concentrations were expected to be highest) were not reported before 1994, due to the inability of 

SNL/CA monitors to detect environmental levels of 3

H at that time (SNL/CA 1994a). In 1994 and 

1995, 3

H releases from the TRL were between 70 and 80% of the LLNL releases. Thus, although 

most of the measured 3

H near the TRL was likely to be due to TRL releases during these 2 yr, some 

of it would have been due to LLNL releases. Thus, the ratio of the measured 3

H concentration in air 

near the TRL to the 3

H release rate from the TRL in 1994 and 1995 is an overestimate of the true 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 34 of 95

ratio. Despite this, the maximum ratio derived for these 2 yr was 6.9 × 10-13, and this multiplier was 

used to estimate air concentrations near the TRL for the years where measurements were not 

available for this location (1979 to 1993). 

Table 4-3. Maximum site-wide annual median inhalation intakes (Bq/yr), 1971 to 2004.

a 

Year 3Hb,c 234Uc 239Puc

1971 4.9E+03d 3.5E-03 6.3E-03

1972 4.8E+03d 6.0E-03 2.8E-03

1973 8.3E+03 5.5E-03 1.5E-03

1974 1.1E+04 4.8E-03 2.9E-03

1975 1.2E+04 4.6E-03 2.0E-03

1976 1.9E+04 5.4E-03 8.0E-04

1977 1.6E+04 7.3E-03 2.1E-03

1978 1.7E+04 9.8E-03 2.7E-03

1979 8.7E+03e 5.2E-03 1.5E-03

1980 9.6E+03e 3.6E-03 4.4E-04

1981 1.4E+04e 7.1E-03 1.4E-03

1982 2.6E+04e 8.0E-03 1.4E-03

1983 1.5E+04e 8.9E-03 2.1E-03

1984 2.0E+04e 7.1E-03 1.2E-03

1985 5.2E+04e 5.3E-03 1.8E-04

1986 7.2E+04e 4.7E-03 1.2E-04

1987 1.8E+05e 5.6E-03 5.3E-05

1988 1.5E+05e 1.3E-02 8.0E-05

1989 8.2E+04e 9.4E-03 1.0E-04

1990 2.9E+04e 5.7E-03 1.2E-04

1991 4.5E+04e 5.9E-03 1.1E-04

1992 2.7E+04e 5.1E-03 8.9E-05

1993 1.8E+04e 4.2E-03 8.6E-05

1994 9.0E+03 3.3E-03 8.2E-05

1995 2.8E+03 2.7E-03 5.9E-05

1996 5.8E+02 2.7E-03 5.8E-05

1997 4.7E+02 2.9E-03 1.6E-05

1998 3.3E+02 2.0E-03 1.4E-05

1999 2.3E+02 3.9E-03 1.3E-05

2000 1.5E+02 0.0E+00f 2.2E-05

2001 1.5E+02 0.0E+00f 9.2E-06

2002 1.7E+02 6.9E-04 7.6E-06

2003 2.0E+02 1.4E-03 6.0E-06

2004 9.2E+01 1.3E-03 7.4E-06

a. Assumes an inhalation rate of 2,400 m3/yr.

b. Assumes total intake is the sum of inhalation and skin absorption, estimated by multiplying the inhalation 

intake by a factor of 1.5.

c. Intakes based on LLNL south perimeter concentrations; see Table 4-1.

d. Intake of 3

H is based on historical LLNL source terms from Peterson (2005); see Table 4-1.

e. Intakes include both LLNL south perimeter concentrations and TRL-generated concentrations; see 

Table 4-1.

f. Zero values indicate the analytical background exceeded the measured concentration of uranium 

isotopes.

In order to consider both LLNL sources and the TRL, the intakes estimated for the south LLNL 

perimeter were added to the intakes estimated near the TRL. Due to the varying relationship 

between the amount of LLNL 3

H releases and the TRL releases, it was not possible to assume 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 35 of 95

that the method of estimating concentrations near the TRL adequately considered LLNL 

contributions, especially for years when LLNL 3

H releases were 10 to 100 times greater than the 

TRL releases. Thus, intakes due to the LLNL south perimeter concentrations and the intakes 

due to the onsite concentrations near the TRL were added.

Particulate intakes (239Pu and 234U)

Measured concentrations of 239Pu, 235U, and 238U at the south perimeter of the LLNL, near the SNL/CA 

site, were reported in annual environmental reports from 1971 through 2004 (LLL 1972 to 1982, 1984; 

LLNL 1986-2005). Intakes of 239Pu for these years (Table 4-3) were calculated assuming the SNL/CA 

site concentration could be approximated by this south perimeter concentration. Intakes of uranium

for these years (Table 4-3) were calculated by summing the intakes of 235U and 238U (Bq/yr) and an 

estimated 234U intake associated with the perimeter concentrations. The 234U intake was estimated by 

assuming it is equal to that of 238U, which is approximately the case with natural uranium. The 

235U:238U ratios on the LLNL main site perimeter have been reported as representative of natural 

uranium (LLNL 1994). The total activity intake of uranium is then assumed to be represented by 234U, 

which is favorable to the claimant because most of the uranium activity (99%) is associated with 238U 

and 234U, and dose factors for 234U are higher.

Before 1971, only gross alpha measurements were available (LRL 1961a, 1963 to 1971). From 1961 

through 1970, the net alpha measurements, calculated by subtracting the average offsite gross alpha 

measurements (i.e., background values) from the reported values, were used to derive intakes of 

either 234U or 239Pu (Table 4-3). Net concentrations calculated to be less than zero were assumed to 

be zero. The assumed radionuclide should be the one which gives the highest dose to the organ of 

interest. For the five years of operation before 1961, there are no measurements available with which 

to estimate the intakes of 234U and 239Pu. 

4.3.3.2 Ingestion Intakes

Ingestion of 3

H from drinking water available to workers was evaluated by considering the maximum 

site-wide intakes estimated for LLNL workers in ORAUT (2005b). It was noted in ORAUT (2005b) 

that several sources of drinking water for the LLNL site were sampled, along with water from the LLNL 

onsite swimming pool, which is close to the main sources of tritium at LLNL. The median activity in 

the drinking water sources was found to be below the limits of detection. However, concentrations in 

the pool were reported to range from 0.8 to 200 Bq/L; thus, the pool water was adopted as the sitewide maximum source of drinking water for LLNL workers (ORAUT 2005b). There is no reason to 

believe that SNL/CA workers would be exposed to drinking water at a higher concentration than the 

pool water at LLNL, so it was assumed here that the maximum concentration of 3

H in drinking water at 

any time might be represented by 200 Bq/L. This concentration would lead to an annual ingestion 

dose of less than 1 mrem/yr, regardless of whether the assumed form of 3

H is tritiated water or 

organically bound 3

H. This calculation assumes a 3 L/day ingestion rate and a 250-day work year 

(i.e., 750 L/yr ingestion of drinking water). Based on this calculation, ingestion of 3

H in drinking water 

is not a significant route of worker exposure.

4.4 EXTERNAL EXPOSURE TO ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

4.4.1 Locations of Concern

Radionuclides present on the SNL/CA site with the potential to cause elevated ambient external 

exposures above background levels are restricted to those associated with the Radiography facility 

(SNL/CA 1993). An aerial survey of the LLNL and SNL/CA, conducted in 1975 (Tipton 1977), found 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 36 of 95

gamma-emitting radioactivity in excess of background levels in the vicinity of Building 9143 (the 

Radiography Building at the time) and to a much lesser extent, in the vicinity of a waste holding area 

known to contain DU and thorium (Building 9122), and a material storage vault known to contain DU, 

thorium, and small amounts of shielded 60Co and 133Ba, and 235U (Building 921, incorrectly referred to 

as B-291 in the Tipton 1977 report). The highest estimated exposure rate at the center of the 

Radiography facility (as seen in the flyover) in excess of background gamma was 40 to 70 μR/hr, or 

100-175 mrem/yr, for a 2,500 hr/yr occupational exposure. The highest values for the holding area 

and storage vaults were 2.0 to 8 μR/hr, or 5-20 mrem/yr for a 2,500 hr/yr occupational exposure. It 

was indicated in the 1977 annual monitoring report (LLL 1978) that none of the elevated areas posed 

a radiation hazard to workers. Furthermore, the survey indicated that the sources were restricted to 

work areas where access was limited. Thus, from the standpoint of environmental exposure, there 

were no areas identified in the survey containing radionuclides that posed a source of elevated 

ambient external exposure to workers.

As with the internal exposures, consideration must be given to external exposures arising from 

activities at the LLNL site in addition to those at the SNL/CA site. As early as 1964, LLNL measured 

perimeter external radiation (LRL 1965). Fluoroglass dosimeters were used at that time, with a limit of 

detection of 50 mrem. The reported dose rates were reported to be less than 0.01 mR/hr, which 

corresponds to less than 88 mrem/yr for continuous exposure. The use of TLDs apparently began in 

1967 (LRL 1968). In 1971, a few perimeter locations were identified at which exposure rates were 

considered elevated above background (LLL 1972). One of these locations (Location 5) was adjacent 

to the LLNL cyclotron building and is at the south perimeter of the LLNL site and thus, directly across 

the street from the north perimeter of SNL/CA. Therefore, this LLNL perimeter location (south 

location) will be considered in evaluating external exposures. Environmental neutron measurements

were also made at this location and reported in annual reports beginning in 1973. These are 

discussed in Section 4.4.3 below.

4.4.2 Gamma-Emitting Radionuclides

The use of TLDs to measure environmental radiation exposure at the SNL/CA site perimeter 

apparently began around 1989, as this was the first time that the five SNL/CA perimeter 

Thermoluminescent Dosimeters (TLDs) (Figure 4-1) are mentioned in annual environmental reports 

(SNL/CA 1990b). Before that time, the site relied on LLNL perimeter and off-site measurements, as 

the latter encompassed the SNL/CA perimeter (Figures 4-2 and 4-3). The estimated average and 

maximum dose external dose rates are listed in Table 4-4 and are based on both LLNL and SNL/CA 

perimeter measurements. The values in this table are applicable to an exposure duration of 2,500 

hr/yr (50 hours per week, 50 weeks per year).
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 37 of 95

Figure 4-1. Thermoluminescent dosimeter locations since 1989.

Figure 4-2. Location of LLNL perimeter gamma and 

neutron dosimeters (LLNL 1986). 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 38 of 95

Figure 4-3. Location of LLNL offsite gamma 

dosimeters (LLNL 1988). 

The data preference for this table was to include SNL/CA measurements when available, but this only 

occurred from 1990 to 1994 (SNL/CA 1991d, 1992, 1993, 1994a, 1995). Before 1990, south LLNL 

perimeter measurements were used to estimate SNL/CA exposure rates. This provides a maximum 

estimate of the contribution of LLNL exposure rates to the SNL/CA rates. This south perimeter value 

is also a reasonable estimate of the environmental exposure rates expected for the SNL/CA. The 

1975 aerial survey indicated this measurement point coincided with the LLNL accelerator and thus, 

had a slightly elevated exposure rate on the order of that seen for the storage vaults and holding area 

at SNL/CA (Tipton 1977). Since most of the SNL/CA area surveyed in 1975 did not show exposure 

rates elevated above background, this is an assumption that is favorable to the claimant.
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Table 4-4. External gamma radiation dose based on 2,500 hr/yr exposure duration (mrem/yr).

Year

Average 

perimetera Errorb 

Maximum 

dose ratec

1956 120d 36 156

1957 120d 36 156

1958 120d 36 156

1959 120d 36 156

1960 120d 36 156

1961 120d 36 156

1962 120d 36 156

1963 120d 36 156

1964 120d 36 156

1965 120d 36 156

1966 120d 36 156

1967 11 3 15

1968 120 36 156

1969 24 7 31

1970 40 12 52

1971 25 7 32

1972 35 10 45

1973 27 8 36

1974 39 12 51

1975 90 27 118

1976 84 25 109

1977 65 19 84

1978 27 8 35

1979 27 8 36

1980 23 7 30

Year

Average 

perimetera Errorb 

Maximum 

dose ratec

1981 17 5 23

1982 19e 6 25

1983 21 6 27

1984 22e 7 28

1985 23 7 30

1986 17 5 23

1987 17 5 22

1988 16 5 21

1989 15 5 20

1990 15 4 19

1991 16 5 20

1992 16 5 20

1993 15 5 20

1994 16 5 21

1995 16f 5 20

1996 16f 5 20

1997 16f 5 20

1998 16f 5 20

1999 16f 5 20

2000 16f 5 20

2001 16f 5 20

2002 16f 5 20

2003 16f 5 20

2004 16f 5 20

a. Until 1990, unless otherwise noted, values are the south perimeter LLNL dose rate; after 1989, values are the 

average of SNL/CA perimeter dosimeters.

b. Error assumed to be ± 30% of the higher of the average perimeter (ORAUT 2006a). 

c. Maximum dose rate is the average perimeter value plus error term.

d. Assumed maximum dose rate estimated for the average perimeter – no measurements reported for the LLNL south 

perimeter or for SNL/CA from 1956 through 1966.

e. Assumed the average of the year prior and year following – missing annual environmental reports for 1982 and 

1984.

f. Assumed average of SNL/CA reported average perimeter values for 1990 to 1994.

There were no reliable measurements for either SNL/CA or the south perimeter LLNL location before 

1967. The 1967 to 1969 values in Table 4-4 assume that the “experimental physics facility” 

mentioned in the corresponding annual reports (LRL 1968 to 1970) represents the south perimeter 

location that is labeled Location 5 from 1970 to 1980 (LRL 1971, LLL 1972 to 1981), but is relabeled 

Location 11 after 1980 (LLL 1982, 1984; LLNL 1986 to 2004). This is a reasonable assumption, since 

this location at that time corresponded to the location of the LLNL cyclotron.

The Table 4-4 exposure estimates for 1970 through 1980 correspond to measurements at LLNL’s 

south perimeter TLD location 5 and, from 1981 through 1989, to TLD locations 11 and 12, with the 

exception of the 1982 and 1984 values, which were estimated by averaging 1981, 1983, and 1985 

values due to the inability to locate annual reports for 1982 and 1984. Finally, SNL/CA TLD 

measurements are included for the years they are reported: 1990 through 1994. For these years, the 

“average perimeter” dose rate in Table 4-4 represents the average of the SNL/CA measurements

reported. After 1994, the average perimeter dose rate is estimated by averaging the 5 yr worth of 

SNL/CA data. 
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Background radiation measurements for the Livermore valley have been reported in LLNL annual 

reports since 1971 (LLL 1972 to 1982, 1984; LLNL 1986 to 2004). Figure 4-4 is a summary of the 

quarterly results from 1988 through 1996, comparing LLNL average perimeter measurements and 

average LLNL Site 300 measurements to the off-site measurements. From Figure 4-4, it is evident 

that there have been only minor quarterly fluctuations around the value of 14 mrem/quarter, but no 

significant long-term trends of either the background measurements or the LLNL perimeter 

measurements. The average yearly background rate, calculated from the yearly rate for these 9 yr

(LLNL 1997), is 57 mrem/yr for continuous exposure or 16 mrem/yr for a 2,500-hr/yr occupational 

exposure.

Figure 4-4. Livermore valley radiation background measurements in comparison with

annual average measurements at LLNL site perimeter and LLNL Site 300 (LLNL 1997). 

4.4.3 Neutron Exposure

The Insulating Core Transformer (ICT) (also known as Rotating Target Neutron Source I) accelerator 

began operation in 1966 in Building 212 at LLNL (Peterson 2005). This facility was known to be 

responsible for elevated levels of environmental neutrons at the LLNL south perimeter, across the 

street from the north perimeter of SNL/CA (LLL 1974 to 1982, 1984; LLNL 1986 to 1988). The 

increased neutron dose rate was attributed to operation of the ICT, also referred to as a 14-MeV 

neutron generator in LLNL annual reports. Environmental neutron monitoring began at LLNL in 1973, 

but was discontinued in 1994 since the responsible operations have been discontinued (LLNL 1996). 

Measurements were made at Location 5 (Figure 4-2) initially (LLL 1974), but a monitor at Location 3 

(originally called 5a) was added in 1975 (LLL 1976). By 1987, the dose rates were at background 

neutron levels, which is approximately 4 mrem/yr for continuous exposure (LLNL 1995).

Although the dose rates were fairly elevated at the measurement location for several years, the dose 

rate at the SNL/CA perimeter would be considerably less due largely to dilution as the beam spread 

increases with distance from the source. The distance from the source to the monitoring locations on 

the LLNL perimeter was approximately 17 m (56 feet) (Willhoite 1979). Assuming that this is also the 

approximate distance to the nearest SNL/CA perimeter (which currently is across a 4-lane street, with 

shoulders on either side), the neutron dose rate at the SNL/CA perimeter would be approximately 

one-fourth the LLNL perimeter dose rate according to the inverse square law. The resulting estimated 

dose rates at the SNL/CA north perimeter are also corrected for an occupational exposure duration of 

2,500 hr/yr and background contributions and are given in Table 4-5. These values are favorable to 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 41 of 95

the claimant, as the dose rate drops off fairly rapidly throughout other areas of the SNL/CA site. Due 

to the lack of information in years before 1973 regarding the operation of the ICT, it was assumed that 

the 1973 value was representative of all previous years of operation (i.e., 1966 to 1972). 

Table 4-5 Elevated neutron dose due to LLNL perimeter neutron source, based on 2,500 

hr/yr exposure duration (mrem/yr)a

. 

Year

North perimeter 

doseb Errorc Maximum dosed

1966 18e 5 23

1967 18e 5 23

1968 18e 5 23

1969 18e 5 23

1970 18e 5 23

1971 18e 5 23

1972 18e 5 23

1973 18 5 23

1974 26 8 34

1975 50 15 65

1976 43 13 55

1977 39 12 51

1978 9 3 12

1979 13 4 16

1980 6 2 8

1981 2 1 3

1982 8f 2 10

1983 8 2 10

1984 8f 2 10

1985 <1 <0.3 <1

1986 2 1 3

1987 <1 <0.3 <1

a. Reported for years neutron dose rate was believed, or unknown, to have exceeded background levels. 

b. Value for south LLNL perimeter dose rate at Location 3 or 5, corrected for distance to SNL/CA, background, and 

assumed exposure time of 2,500 hr/yr. 

c. Error assumed to be ± 30% of the average perimeter dose (ORAUT 2006a). 

d. Maximum dose rate is the average perimeter value plus error term.

e. Assumed value reported for 1973 applied to previous years back to 1966.

f. Assumed the 1983 value – missing annual environmental reports for 1982 and 1984.
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5.0 INTERNAL DOSIMETRY

5.1 SUMMARY OF POTENTIAL INTERNAL RADIONUCLIDE EXPOSURES

Work with radionuclides that created a potential for internal exposure at SNL/CA included limited 

uranium operations primarily (but not exclusively) DU machining, and research conducted in the 

Tritium Research Laboratory (TRL). Radiation workers were monitored using external dosimetry and

urine bioassay. A “Radiation Worker” was defined as an “employee who receives or could potentially 

receive radiation exposure from his job in excess of 10% of the applicable standards for internal or 

external exposure” (Wright 1979a). For the purpose of internal dose, individuals working with 

radioactive materials that present a potential internal exposure were included as radiation workers 

(Wright 1979a). The numbers of radiation workers based on internal exposure varied over the years. 

Twenty-one individuals were defined as radiation workers for 1980 based on potential tritium 

exposure. 

5.1.1 Tritium Research Laboratory

The TRL was established in 1976 to perform research and development functions for the DOE Office 

of Defense Programs to support weapons development. The TRL at its peak operation employed 

approximately 35 experimenters and staff support personnel (Garcia and Gorman 1996). 

The building was divided into two zones, (1) an office area and (2) a radioactive materials area. The 

zones were separated by two sets of double doors. The room air in the radioactive materials area 

was continuously monitored for tritium. The monitoring systems were set to alarm at specific action 

levels but there is no indication that the tritium monitoring data were ever used to estimate worker 

intakes. The TRL operations were generally concerned with the physical and chemical 

characterization of tritium and its compounds. Fabrication of tritium compounds for use as 

engineering components was also part of the mission of the TRL. All operations involving gram 

quantities of tritium were conducted inside gloveboxes. The building had a control room where data 

from the various monitoring devices were stored. This provided for real time monitoring of facility 

conditions (Wright 1981a).

Tritium in quantities greater than 0.1 grams was doubly contained in gloveboxes, special Sandiadesigned containers, or Department of Transportation-approved containers. The Safety Analysis 

Report (Wright 1981a) anticipated that 50 grams of tritium might be handled in one doubly contained 

system and that the total tritium inventory in the facility would be approximately 300 grams. 

Tritium handling operations were terminated in 1992. Cleanup activities were conducted at TRL from 

January 1992 to December 1995. After 1996, the TRL complex was converted to the CRDL (Garcia 

and Gorman 1996). 

5.1.2 Uranium Alloy Machining

Uranium alloy machining was performed at the SNL/CA facility starting before 1972. Several memos 

indicate that machining and testing of uranium alloys had previously been performed at the Union 

Carbide Y-12 plant and elsewhere. However, requirements for SNL/CA were such that they needed 

“between 50 and 100 specimens of various shapes and sizes per month,” (Adolphson 1972), thus 

requiring initiation of an on-site program.

The radiation safety requirements for the machining operation are described in SOP Machining 

Depleted Uranium Metal (SNL/CA 1989a). The depleted uranium is designated in the SOP as D-38. 
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The radiation safety requirements include air sampling and urine bioassay (SNL/CA 1989a). Daily air 

samples were required during all machining operations even though air-sampling data had shown that 

neither wet machining nor “burning (D-38) chips” (SNL/CA 1989a) resulted in airborne particulates. 

Urine bioassay was required semi-annually.

In addition to inhalation exposures, machining of D-38 posed a potential hazard from cuts and 

splinters. All such incidents were to be reported to the Medical Department and wound counts made 

on any puncture wounds. No data on wound counts were found in the records available to the 

authors.

The Uranium Tritide Bed installed at the TRL in 1991 also presented a potential for exposure to 

uranium powder [TRL, SOP No. 757 (SNL/CA 1991a)]. The operating procedure did not specify 

bioassay requirements. In addition, parts contaminated with UO2 dust were received by SNL/CA 

(Lovell 1982) posing a potential airborne dust hazard.

5.2 BIOASSAY PROGRAMS

The only radionuclides of concern for internal dosimetry at SNL/CA appear to have been tritium and 

uranium. Tritium exposure occurred primarily in the TRL. However, tritium is ubiquitous in the natural 

environment as it is a cosmogenic naturally occurring radionuclide and was produced by atomic 

weapons testing (Turner, 1995). Tritium was also present in the environment due to activities at the 

LLNL facility adjacent to SNL/CA. 

The primary detection method for intakes of 3

H at all SNL facilities has been urine bioassay (Potter ca. 

1997). There is no evidence that urine bioassay samples were analyzed for any radionuclides except 

H-3 and natural uranium (U-nat), or that other types of bioassay, i.e., fecal analyses or in vivo 

counting, were employed at SNL/CA. The H-3 bioassay was limited to TRL experimenters and staff. 

The uranium bioassay was performed on individuals involved in machining DU as well as others 

involved with handling uranium powders or those in areas where air concentrations potentially 

exceeded 10% of the air concentration guidelines (Wright 1979a). Records indicate that tritium 

bioassay was performed at SNL/CA only after the TRL became operational. 

5.2.1 Tritium Bioassay Programs

According to (Garcia and Gorman 1996), tritium bioassays were performed weekly for individuals 

involved in experimental work at the TRL from 1979 through 1995 when the TRL was 

decommissioned. Individuals working in the Waste Handling Facility may also have participated in the 

tritium bioassay program, at least during 1991 (Garcia 1991a, 1991b). Bioassays were also required 

for all personnel inside the TRL when an evacuation alarm occurred (SNL/CA 1991b). Additional 

samples were required in some cases by Safe Work Permits (SWPs) or at the discretion of Health 

Physics. 

The Tritium Research Laboratory Safety Analysis Report (Wright 1981a) specifies weekly bioassay 

with samples analyzed by liquid scintillation counting. Laboratory analyses were performed, in house, 

by the Health Physics Division. Doses were calculated and reported on a monthly basis. 

A 1993 memo from Donn Wright to Lydia Perez (Wright 1993) describes the methods by which tritium 

bioassay data were managed between 1977 and 1993. According to Wright, bioassay results were 

originally maintained in a VAX text file. The VAX files were transferred into REFLEX, a database 

management program in 1988. In 1990, the files were sent to SNL/NM for entry into personnel 
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dosimetry histories. As of 1993, the date of the memo, the bioassay data were collected by SNL/CA 

and a hard copy sent to SNL/NM monthly or quarterly. 

The TRL Health Physics Quarterly Summaries from 1988 through 1995 confirm that urine specimens 

were collected on a weekly basis and tritium concentrations determined by liquid scintillation counting. 

Initially absorbed doses were calculated for all individuals whose urine bioassay results indicated that 

they might have received a dose greater than 10 mrem per calendar quarter based on the fact that 10 

mrem was the reported sensitivity of the TLDs used for external dosimetry. From 1991 through 1992, 

absorbed doses were calculated for individuals whose bioassay results indicated a dose greater than 

2 mrem per calendar quarter (Garcia 1991b). From 1993 through 1995, absorbed doses were 

calculated for individuals whose bioassay results indicated a dose greater than 1 mrem per calendar 

month (Garcia 1993a). 

5.2.2 Uranium Bioassay Programs

Uranium bioassays were required for SNL/CA workers involved in DU machining and other operations 

where airborne uranium might have been encountered. The urine bioassay criteria for DU were 

described in Wright (1979a). The criteria for minimum routine (semi-annual) uranium bioassay were 

as follows:

• When air sampling results show concentrations at or greater than 10% of the concentration 

guide of 1E-10 μCi/cm3

. 

• For routine handling of uranium hydrides, solutions of uranium compounds, and uranium 

powders (>4 times per quarter). 

• Machining of uranium. 

• During any operation that the Hazards Control Division deemed hazardous, or requires air 

sampling in a SOP or SWP. 

Nonroutine bioassays were performed in the following situations:

• Cut or lesion during handling or machining of uranium. 

• An individual in close proximity or exposed to a uranium metal fire.

• Skin contact with a solution of uranium. 

DU alloy machining was performed under SOP 1066 (SNL/CA 1989a). The SOP required that all 

machining operations be performed wet, reducing the risk of fire and generation of airborne dust. 

Revision C, dated October 1, 1989 (SNL/CA 1989a), required semi-annual urine bioassay for uranium 

but noted that “more frequent urine samples are usually collected.” 

Air sampling was also required during machining. The machine operators were charged with the 

responsibility for turning on and off the air samplers. No air concentration data were available to the 

authors.
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5.3 BIOASSAY LABORATORY PROCEDURES AND MINIMUM DETECTABLE INTAKES 

The bioassay measurements were performed by various different entities during the period of 

operation of SNL/CA. No specific laboratory procedure manuals were available for either tritium or 

uranium analyses at SNL/CA. 

5.3.1 Tritium

Tritium bioassay was performed at SNL/CA during the entire history of the TRL by liquid scintillation 

counting. Guidelines for the bioassay program are contained in Health Physics Operating Procedure 

471583. Note: That document was not available to the authors as of the time this SPD was 

submitted. 

Revision 3 of the TBD for internal dosimetry at SNL/CA and SNL/NM states that workers were 

instructed to fill the entire 1500 mL urinalysis container (Potter ca. 1997). However, a single void 

urine sample was prescribed for tritium. It should be noted that Revision 3 of the TBD was produced 

after the TRL operations ceased at SNL/CA. (The earliest date for the document would have been 

1997 based on the dates of the references listed.) 

The frequency and minimum detectable activities (MDAs) for tritium bioassay, gleaned from the 

available quarterly Health Physics Reports and Standard Operating Procedures are given in 

Table 5-1. 

Table 5-1. MDAs for and frequency of tritium bioassay. 

Year MDA Frequency Source

1979–

1983

Approximately 10 

nCi/L 

(background 

+/-2 std. dev.)

Weekly TRL Health Physics Summary, 1979-1983 (Lovell, Wright, and 

Hafner 1984)

1980 10 nCi/L, Weekly TRL Health Physics Summary, First Quarter 1980

(Hafner 1980)

1980 20 nCi/L Calculated from liquid scintillation counter efficiency and 

background information in 3H dose worksheets for 1980 

1981 Approximately 10 

nCi/L 

(background +/- 2 

std. dev.)

Weekly TRL Health Physics Summary, First Quarter, 1981

(Wright 1981a) 

1986 Approximately 20 

nCi/L

Calculated based on computer printout for positive exposures 

for 1986

1988 None given Weekly Tritium Research Laboratory Health Physics Quarterly 

Summaries (October) (HCD 1988)

1989 None given Tritium Research Laboratory Health Physics Quarterly 

Summaries (HCD 1989a,b,c,d) 

1990 None given Weekly Tritium Research Laboratory Health Physics Quarterly 

Summaries (Garcia 1990a,b,c,d)

1991 None given Weekly Tritium Research Laboratory Health Physics Quarterly 

Summaries (Garcia 1991c,d,e,f)

1991 None given. 

Reporting limit of 

1 μCi/L specified. 

Minimum frequency –

monthly: weekly for 

individuals working 

primarily in the TRL

SOP No. 709, Tritium Research Laboratory Building 968

1992 None given Weekly Tritium Research Laboratory Health Physics Quarterly 

Summaries (Garcia 1992a,b,c,d)

1993 None given Weekly Tritium Research Laboratory Health Physics Quarterly 

Summaries (Garcia 1993a,b,c,d)

1994 None given Weekly Tritium Research Laboratory Health Physics Quarterly 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 46 of 95

Summaries (Garcia 1994a,b,c,d)

1995 None given Weekly Tritium Research Laboratory Health Physics Quarterly 

Summary for First Quarter (Garcia 1995)

1998 5.5 nCi/L Wright 1998 Memo to Debbie Miller DOE/AL

Special bioassay

According to Hafner (2006), tritium bioassay samples were counted in a liquid scintillation counter 

under a standard protocol that required a 10-minute count. The volume of urine to be used in any 

sample was not specified but is assumed to be a minimum of 0.5 mL based on information in 

available dose calculation forms. The background count in the tritium channel was reported by Hafner 

to be approximately 22 cpm. Calibration forms for 1995 give background counts in the tritium channel 

as approximately 15 cpm with counting efficiencies of approximately 0.67. Daily performance checks 

from 1993 to 1995 showed efficiencies in the range of 0.65 and background counts from 17 to 21 

cpm. Based on these parameter values, the minimum detectable concentration (MDC) for tritium in 

urine would be approximately 13 nCi/L.

MDA = (4.66Sb + 2.71)/(E*A*R*V) [Vosburg 1993]

where:

Sb = blank standard deviation

E = efficiency

A = decays per disintegration = 1 for tritium

R = chemical recovery (not applicable for liquid scintillation counting for tritium)

V = volume of the sample = 0.5 mL

MDA = 13.5 dpm

Assuming a 0.5 mL aliquot, the MDC would be 29,000 dpm/L or 13 nCi/L. This is consistent with the 

MDA of 10 nCi/L reported in the 1980 and 1981 TRL Health Physics Summaries (Hafner 1980; Wright 

and Hafner 1981). However, a printout of positive exposures for 1986 shows efficiencies between 0.3 

and 0.4 and background approximately 20 cpm. While the volume of the aliquot used is not specified, 

a review of the data in the printout supports the assumption that aliquots of 0.5 mL were used in the 

analyses. This would lead to an MDC of approximately 20 nCi/L. 

Tritium monitoring results were generally reported in dose rather than urine concentration or 

estimated intake. However, bioassay appointment cards used during the period April 1982 through 

October 1982 provided the results of the bioassay and the calculated dose.

The SNL/CA personnel records reviewed generally report calculated dose for tritium rather than urine 

bioassay data. Chronic Dose Worksheets for 1980 indicate that tritium doses for SNL/CA workers 

were calculated based on bioassay results as follows: 

Dose = (0.286)(Δt)(μCi/L)

The origin of the constant, 0.286, is not defined, and Δt is assumed to be the period of time for which

dose is calculated. The dose formula is included here to allow back-calculation of urinary tritium 

concentrations and tritium intake from dose data.
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5.3.2 Uranium 

Uranium bioassay samples were sent to SNL/NM for analysis either in house or by a contract 

laboratory (SNL/CA 1965-1990, Potter 1994). In general, bioassay procedures at SNL/NM required 

24-hour urine sample collection. However, urine bioassay questionnaires that were completed by 

monitored workers at SNL/CA in 1989 and 1990 noted that approximately 50 mL were needed for 

analysis (SNL/CA 1965-1990). The form contained very brief instructions for sample collection. In 

contrast, the SNL/NM bioassay kit for 1993 included detailed instructions on how to obtain and deliver 

the sample (SNL/NM 1993a). Presumably since the uranium bioassays were processed by SNL/NM, 

the SNL/CA kits would have been the same for that period. 

The analyses of uranium bioassays were performed by various different laboratories over the course 

of uranium processes at SNL/CA. For most of the time, the SNL/NM Industrial Hygiene Laboratory 

performed the uranium bioassay by either fluorimetry or by inductively coupled plasma mass 

spectrometry (ICP-MS). For several years during the late 1980s, Thermo-Analytical Incorporated/EAL 

Corporation (TMA/EAL) performed the uranium bioassay. Controls for Environmental Pollution (CEP) 

performed analyses during the early 1990s (SNL/NM 1993a, 1993b). However, subsequent intercomparison studies indicated that the data from CEP were not reliable (Ball et al. 1995). Therefore, 

the DR should not use urine bioassay data from CEP analyses in the dose reconstruction. The 

timeline for uranium bioassays of SNL/CA workers is given in Table 5-2. The information in the table 

was gleaned from individual employee bioassay reports and Wright (1979a).

The results of urine bioassay were generally reported in mass concentration units. There is no 

information to demonstrate that appreciable amounts of either natural or enriched uranium were used 

at SNL/CA. Therefore, the DR should assume a specific activity for DU when converting mass 

concentration to activity concentration. The most reasonable specific activity for depleted uranium is 

4.38E-7 Ci/g as stated by Wright (1979a). 

Wright (1979a) states that the minimum detectable limit by the fluorimetric analysis method used by 

SNL/NM was 1E-2 μg/L or 4E-3 μCi/L. These two values are inconsistent, assuming the specific 

activity of 4.38E-7 Ci/g for depleted uranium also stated by Wright. A mass concentration of 1E-2 

μg/L would be equivalent to an activity concentration of 4.38E-9 μCi/L. A 1975 analytical report from 

the Livermore Medical Department shows a detection limit for uranium of 0.005 μg in a 50 mL sample 

by an unspecified method, indicating a MDC of 0.1 μg/L (SNL/CA 1975-1977). Other sample data 

sheets show minimum detectable amounts of 0.01 μg (SNL/CA 1975-1977). With an aliquot 

presumed to be approximately 15 mL (the record was nearly unreadable), this would indicate an MDC 

of approximately 0.7 μg/L. 

Table 5-2. Preliminary time line uranium in urine analyses. 

Period Laboratory Method MDA Basis and comments

<1968 No information No information No information No information

1968–

unknown

Radiation Detection 

Company (RDC)

Fluorimetry 5 μg/L Reported error (+/- 2σ)

1973–1986 SNL/NM Industrial 

Hygiene Services

Fluorimetry

Methods SAND 

75-0014 also 

reported on the 

data sheets. 

Based on lowest 

reported levels, 

and the prior 

MDA , the MDA 

applicable to this 

period would 

most likely be in 

the range of 5 

μg/L

Results reported as 0 in most 

cases with the lowest reported 

non-zero values of 1.1 μg/L to 

3.2 μg/L appearing occasionally 

in the records. Retests were 

requested for positive results

1986 SNL/NM Industrial ICP listed as No information; It is unlikely that the method was 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 48 of 95

Hygiene Services Procedure No. all results 

reported as 0

inductively coupled plasma mass 

spectrometry. It did not come 

into general use until the early 

90s. ICP alone is not sensitive.

1986 to 1987 TMA/EAL Not specified but 

probably 

fluorimetry

“<” values 

ranged from 3 to 

5 μg/L

“<” values provided in the reports

1987–1990 SNL/NM Industrial 

Hygiene Services

Fluorimetry 

Method SAND88-

1149 also 

referenced

“<” values 

ranged from 10 

to 12 μg/L 

Notes on the reports

1990–1993 CEP No MDC given Audit memos show that CEP 

analyses were problematic.

1993-present SNL/NM Industrial 

Hygiene Services

Not defined but 

probably ICP-MS

0.1 μg/L Reasonable MDAs for ICP-MS

5.4 BIOASSAY RESULTS

Garcia and Gorman (1996) provide data on doses calculated from urine bioassay of TRL 

experimenters and staff. However, he provides no information on how the urine tritium concentrations 

or intakes were estimated. For the early years of operation (1979 to 1981) the only data available are 

the number of workers with doses within specified ranges as shown in Table 5-3. 

Table 5-3. Range of tritium doses based on bioassay, 1979 to 1981 (Garcia and 

Gorman 1996). 

Year

Number of workers receiving:

10–100 mrem/yr 101–500 mrem/yr

1979 4 1

1980 4 1

1981 6 0

For the years from 1982 through 1995, (Garcia and Gorman 1996) shows maximum and average 

doses from intake of H-3 at TRL as well as total person-mrem (Table 5-4). The number of workers 

monitored can be inferred by dividing the total person-mrem by the average dose in mrem.

Table 5-4 Estimated tritium doses based on bioassay, 1982 to 1995 (Garcia and Gorman 1996). 

Year

Maximum Dose

(mrem/y)

Average Dose

(mrem/y)

Total Person-Dose

(mrem/y)

No. of monitored 

Individuals (inferred)

1982 70 30 183 6

1983 79 49 148 3

1984 1620 152 3040 20

1984* 234 75 1420 19

1985 347 65 2270 35

1986 229 67 1330 32

1987 178 42 580 14

1988 218 63 1652 26

1989 232 46 2465 54

1990 262 30 1056 35

1991 111 11 458 42

1992 53 11 222 20

1993 63 17 257 15

1994 69 15 190 13

1995 42 15 134 9

*Maximum, average, and total person-dose excluding maximum individual single event
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5.5 METHOD OF CALCULATING DOSE FROM URINE BIOASSAY DATA

In most cases, the information available in the records generally provides only the end result of the 

calculation, not the actual urine bioassay concentration data. The methods of calculating doses would 

have varied over time as more metabolic information and data on retention and urinary excretions of 

the radionuclides of interest became available. 

The SNL/CA personnel records reviewed generally report-calculated dose for tritium rather than urine 

bioassay data. Chronic Dose Worksheets for 1980 indicate that tritium doses for SNL/CA workers 

were calculated based on bioassay results as follows: 

Dose = (0.286) (Δt) (μCi/L)

Again, the origin of the constant, 0.286 is not defined. 

5.6 AIRBORNE RADIONUCLIDE CONCENTRATIONS

Work in the TRL was conducted in sealed gloveboxes and/or in gloveboxes used in high-velocity air 

hood (HVAH) mode. The TRL used two systems for decontaminating glovebox air before release to 

the environment. The GPS removed tritium from sealed gloveboxes. The Vacuum Effluent Recovery 

System (VERS). The TRL high flow ventilation provided 6 to 10 room air changes per hour (SNL/CA 

1991b). 

Tritium monitors were installed in TRL. The monitors were used to detect the release of tritium into 

room air, gloveboxes, the stack, and to monitor the performance of the GPS/VERS (SNL/CA 1991b). 

The monitors activated audible/visible alarms at the monitoring point and in the TRL control room. At 

least one operating tritium monitor was required in each room. 

According to the Quarterly Reports, the tritium monitoring system was continually upgraded during the 

1988 through 1995 period. A planned new tritium monitoring system was cancelled in 1992 (Garcia 

1992a). Monitors were installed in the hallways of the TRL. Double monitoring was conducted in 

each laboratory. As of 1988, sixty tritium monitors were in use (Hazards Control Division 1989a,

1989b, 1989c, 1989d).

The tritium monitors had Room Air Low and Room Air High Alarms. The Room Air Low alarm was 

triggered at a concentration of 30 μCi/m3. The High Alarm was set at 1 mCi/m3. The maximum 

permissible concentration in air (MPCa) and the more recent derived air concentration (DAC) 

remained constant at 2E-5 μCi/mL (20 μCi/m3) throughout the operational history of the TRL. The 

alarms were checked on a routine basis. If a Room Air Low alarm was triggered, workers were 

required to mitigate the release within a few minutes and exit the room if the air concentrations did not 

decrease. A Room Air High alarm required evacuation from the room within one minute (SNL/CA 

1991b). 

Airborne uranium dust concentrations were measured in the areas where DU was machined. No data 

on measured concentrations or action levels were available to the authors.

5.7 UNMONITORED DOSE ESTIMATES

The vast majority of SNL/CA workers were employed in areas with little or no potential for intake of 

radionuclides generated at their work areas (SNL/CA 1979, Bryson 1972). Therefore, only a small 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 50 of 95

proportion of workers were required to participate in the bioassay program. Individuals who worked in 

the laboratory areas as experimenters or support personnel may have been exposed to airborne 

tritium (SNL/CA 1991a, 1991b). Machinists and others who worked in areas where DU was machined 

or other uranium operations were conducted may have been exposed to airborne uranium dust

(SNL/CA 1989a). 

If no bioassay records are available for claimants who are on record as having worked in the TRL or 

uranium machining areas, potential unmonitored doses or intakes can be calculated based on 10 

percent of the airborne radionuclide concentrations limits since procedures required monitoring for 

individuals who could have potentially received internal doses in excess of 10 percent of the dose limit 

(Wright 1993) or when air sampling results demonstrated that air concentrations could reach or 

exceed 10 percent of the concentration guides for air (Wright 1979b). The applicable MPCas in air, by 

year, are given in Table 5-5. 

Table 5-5. MPCas for tritium and uranium. 

Date MPCa or DAC (40-hr week) Source

3

H (water) 238U

1959 2 E-5 μCi/mL 1E-10 NBS Handbook 69 (NBS 1959)

1990s 2 E-5 μCi/mL 2E-11 10 CFR 835, Appendix A

The calculated annual intake for a worker inhaling tritium at an MPCa of 2E-5 μCi/ml would be 5E4 

μCi. The calculated annual intake for a worker inhaling uranium at the MPCa of 1E-10 μCi/ml would 

be 0.24 μCi; at 2E-11μCi/ml, the annual intake would be 0.05 μCi assuming a breathing rate of 10 

cubic meters per day for 240 days per year (2.4 × 109 ml/year). A concentration of uranium in air of 

10 × 10-10 μCi/ml multiplied by the breathing rate would result in a total annual intake of 0.24 μCi. 

Because SNL/CA is located in close proximity to LLNL, it is not the only source of airborne 

radionuclides. Intakes for workers outside those facilities can, if necessary, be calculated based on 

environmental air concentrations.

5.8 RADIATION DOSIMETRY REPORTS

The AEC required annual dose report summaries for workers (Burke 1969). Records of annual dose 

reports for 1958 through 1967 were reviewed. The reports from 1964 through 1968 and 1973 showed 

no internal body depositions as determined by bioassays or exposure to airborne materials that

resulted in internal body deposition. The reports for 1962 and 1963 specified no internal depositions 

in excess of one-half the body burden. Between 1958 and 1962, the reports state that there were no 

exposure resulting in internal body deposition of radioactive materials. The reporting form changed in 

1976; subsequent reports for 77 and 78 did not show internal exposures. 

The Personnel Monitoring and Laboratory Services department at SNL/NM issued monthly ALARA 

Radiation Dosimetry Reports that list all employees whose annual dose equivalent exceeds specific 

action levels of 2%, 6%, and 10% of the applicable 10 C.F.R. pt. 835 limiting values (Personal 

Monitoring and Laboratory Services 1997). The reports include external doses as measured by TLD 

badges and internal doses as committed effective dose equivalent (CEDE). The column for CEDE 

remained blank in most cases, presumably where no bioassays were performed. It is not clear 

whether these reports included SNL/CA employees.

Pre-1973 external doses were documented by SNL/CA. Tritium doses acquired at sites other than 

SNL were noted occasionally in the records (SNL/CA undated-c). The notation showed values less 
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than 3 μCi/L presumably in urine. One record in this undated document showed 3 MC/L; however, 

the entries in this record were all in capital letters, and most likely were most likely meant to be 3 

μCi/L. 

Termination Occupational Exposure Reports were also filed for SNL/CA employees. These reports 

were primarily focused on external exposures but did include a section for internal exposure (SNL/CA 

1980-1984). 

5.9 SUMMARY TABLES

The following summary tables for MDAs incorporate the best general information available. However, 

it should be noted that some individual data in the records reviewed showed lower MDAs or reported 

urine bioassay values less than the MDA listed.

Table 5-6. Internal dose control program. 

Routine Monitoring Type Period Frequency

Urine – tritium 1977–1995 Weekly

Urine – uranium 1968–present Semi-annual or as required by the 

Hazards Control Department

Table 5-7. Detection limits for urine bioassay 

for 3

H with liquid scintillation counting. 

Period MDA (pCi/L)

1977–1983 10,000 (reported)

1980 - 1985 20,000 (calculated for 

1980 based on 

background and 

efficiency for the LSC)

1986 20,000 (calculated)

1986–1997 10,000 (estimated 

based on previous 

reported MDAs)

1998 – present 5,500

Note: Reporting limits were based on dose not 

bioassay results.

Table 5-8. Detection limits for total uranium bioassaya. 

Method Period MDA (μg/L)

No data <1968 5 (based on state of 

technology)b 

Fluorimetry 1968 5 (based on reported 

error)

Fluorimetry 1969 –1973 5 (estimated based on 

1968 data)

Fluorimetry 1973 –1986 1.1 (based on lowest 

reported value in 

individual employee 

bioassay reports)

Fluorimetry 1975 0.7 (based on 

individual employee 

data

Fluorimetry 1986 –1987 3

Fluorimetry 1987–1990 10
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CEP 1991–1992 12 (highest reported 

in previous years from 

Table 5-2)c

ICP-MS 1993 – Present 0.1

a. No reporting limits were available in the documents provided to the authors. 

b. MDA of 5 μg/L is consistent with the value reported for bioassay conducted at nearby LLNL 

site (ORAUT 2005c) and the Hanford site (ORAUT 2004a). 

c. Significantly higher than the years immediately following.

Table 5-9. Solubility type, fraction activity and particle size. 

Facility Nuclide Solubility type Particle size

TRL (Building 968) 3H Tritiated water Not applicable

Building 913 

Machine Shop

DU Metallic form – unspecified 

solubility type but likely Type

S

No data

Building 968 Uranium tritide Metallic Type S; no data to 

show tritide is more soluble 

than metal

No data

Building 913 UO2 Type S No data
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 53 of 95

6.0 OCCUPATIONAL EXTERNAL DOSE

6.1 PURPOSE

The purpose of this section is to detail historical external dosimetry programs, systems, and practices 

at SNL/CA. This information may be used by dose reconstructors as needed to evaluate external 

occupational doses for EEOICPA claimants in terms of supplementing individual dose records for 

monitored SNL/CA workers, estimating respective missed doses, or estimating doses for unmonitored 

workers. 

6.2 SCOPE

Historical documentation related to radiological protection programs at SNL/CA indicates that external 

dosimetry monitoring for workers and visitors at SNL/CA has been performed throughout the site’s 

history. The information provided in this section draws on review of currently available records. The 

reivew for years prior to 1991 contains less information due to fewer available supporting documents. 

In the event further relevant documents are found for any period during the site’s history, this section 

of the SPD will be revised accordingly. 

6.3 INTRODUCTION

Historically, tritium exposure has been the primary radiological concern at SNL/CA (Ullrich 2003, DOE 

2006). However, tritium exposure is not significant in terms of external dose. As noted in Section 2.3, 

the laboratory typically handled kg amounts of DU, typically in the form of alloyed metal components, 

g amounts of 3

H, and μCi quantities of other isotopes. The Radiography Facility (Building 923) also 

contained 100 Ci 192Ir and 60Co sources, a sealed 252Cf source, and many other smaller sealed 

sources of activity ranging from one Ci to 500 mCi (see Section 2.3). According to a DOE Web site, 

uranium exposure potential at SNL/CA is extremely small and there is no potential for plutonium 

exposure (DOE 2006). Radioactive waste at SNL/CA is categorized as low-level. High-level or 

transuranic waste is not generated or stored at SNL/CA (SNL/NM / ES&H 1992). Table 2-1 

summarizes the SNL/CA buildings of potential radiological interest. 

With respect to radiation protection at SNL/CA, from 1956 to 1959 Lawrence Livermore National 

Laboratories (LLNL) performed external film dosimetry monitoring services for SNL/CA (SNL/CA 

1958–1961). After that, a commercial vendor called Radiation Detection Company (RDC) provided 

film dosimetry services until the early 1970s (SNL/CA 1958–1961, SNL/CA 1961–1962, SNL/CA 

1962, DeSelm 1965, SNL/CA 1964, Lovell 1966, RDC 1969). In 1962, there was a brief transfer of 

contracted film processing services from RDC to a similar provider called Tracerlab (SNL/CA 1962),

but Tracerlab was quickly dropped due to poor performance and the contract with RDC was reestablished. From about 1972 to 1988, SNL/CA external dosimetry was outsourced to the DOE 

Radiological and Environmental Sciences Laboratory (RESL) in Idaho Falls, ID (Wright 1993, Wallace 

1988, Ormond 1986). In 1989, dosimetry services for SNL/CA were transferred to SNL headquarters

in Albuquerque, New Mexico (SNL/NM).

In 1991, the dosimetry processing laboratory at SNL/NM became DOE Laboratory Accreditation 

Program (DOELAP)-accredited (Loesch 1991) as part of the overall plan to provide centralized, 

unified, and permanent dosimetry services for SNL/CA and other Sandia sites (Stanley 1991; SNL/NM 

/ES&H 1992; Ward et al. 1994). Most of the dosimetry information presented, described, and 

evaluated in this section of the SPD for years after 1991 was obtained at SNL/NM by the ORAU Team 

for the purpose of developing a SPD for that site (those records are directly applicable to the 

dosimetry program at SNL/CA from 1991 forward). Although the records currently available for this 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 54 of 95

SPD do not fully describe many technical details of dosimetry programs before DOELAP accreditation 

in 1991, they do provide an indication of the types of dosimeters used, exchange periods, and in most 

cases, types of radiation dose quantities that were measured and recorded.

6.4 DOSE RECONSTRUCTION PARAMETERS

6.4.1 Site Historical Administrative Practices

6.4.1.1 Administrative Practices Before 1989

Documents from early in SNL/CA’s history indicate a policy of maintaining permanent dosimetry 

records (SNL/CA 1963). Between 1956 and 1959, most SNL/CA employees, contractors and visitors 

were required to wear dosimeter badges (SNL/CA 1958-1961, SNL/CA 1958-978). During this time,

LLNL was performing dosimetry services for SNL/CA, with respective records management

implemented manually by SNL/CA personnel. Records management continued to be administered by 

SNL/CA (division 8242-2) for the subsequent period in which RDC provided dosimetry services (1959 

to about 1971). By 1965, discussions among SNL/NM management were taking place with respect to 

the need to badge all personnel on site versus badging only personnel with access to “exclusion 

zones” (areas where radiation fields were present). Eventually, the policy of badging all personnel 

was abandoned in favor of badging only individuals with potential to exceed certain exposure limits, 

which appears to have occurred in 1970, according to summarized annual reports of external 

exposures (SNL/CA 1958-1978). Through about 1972, dosimetry results were recorded manually on 

4 by 6-in. cards or were stored as text file hard copies, all of which were transferred to SNL/NM in 

1993 (Wright 1993). Electronic external dosimetry records for years between 1973 and 1988 were 

transferred to SNL/NM in 1989 (Wright 1993), apparently to become part of a master electronic 

records database management program called SANDOS. Since 1989, dosimetry records for all 

Sandia sites have been retained and managed by SNL/NM (Ward et al. 1994). 

6.4.1.2 Administrative Practices After 1989

In 1989, the dosimetry program at SNL/CA was transferred to SNL/NM and has since been directed 

and managed by SNL/NM’s Personnel Dosimetry Department 7715, with on-site oversight and 

implementation conducted by SNL/CA’s Personnel Dosimetry Department 8541 (Ward et al. 1994). 

Dosimetry records generated by Department 7715 are entered into SANDOS. Some example 

personnel dose history output (SNL/NM 1995a) is shown in Attachment B (Figure B-1). Although 

dose units are not stated in SANDOS output, review of other documents (SNL/NM 1995a) suggests 

that they are reported in rem. There are specific protocols and training required for dosimetry data 

entry and records management (Ward et al. 1994). 

Consistent with DOE Order 5480.11, Section 9.g. (DOE 1989), administrative external dose limits for 

SNL/CA personnel are as follows:

• 100 mrem (0.001 sievert) annual effective whole body dose equivalent 

• 5 rem (0.05 sievert) annual dose equivalent to the skin

• 5 rem (0.05 sievert) annual dose equivalent to any extremity

• 1.5 rem (0.015 sievert) annual dose equivalent to the lens of the eye
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 55 of 95

Any site personnel with potential to receive annual doses exceeding these limits are required to wear 

personal dosimeters (Thompson et al. 1991; Ward et al. 1994). Monitored workers with interim 

dosimetry results indicating that annual doses could approach or exceed these limits may have work 

restrictions imposed. In 1997, a memo was issued (SNL/NM 1997) announcing an apparent change 

in dose limit reporting protocols under SNL’s ALARA policy (Figure 6-1). These limits are similar to 

those listed above, but it is unclear whether they represent a combined total of both internal and 

external doses. 

Figure 6-1. ALARA reporting action levels.

There are two dosimeter categories for monitoring personnel on site: routine field dosimeters and 

nonroutine field dosimeters. Routine field dosimeters are required (as specified above) for regular 

employees or contractors working at the site for extended periods, and results are reported on a 

quarterly basis. Nonroutine field dosimeters are issued to short-term temporary workers, site visitors, 

or personnel performing special radiation work outside the scope of their normal activities (i.e.

performing nonroutine work tasks where the potential for additional or special types of exposures may 

exist). 

While not made explicitly clear in Section 8.4.1 of SNL’s External Dosimetry Program Manual (Ward 

et al. 1994), routine field dosimeter exchanges appear to occur on a monthly or quarterly basis, while 

nonroutine exchanges occur on a bi-weekly basis (for special radiation work cases, exchange periods 

may vary). 

Before doses are calculated from gross field TLD readings, the amount of measured 

thermoluminescence due to background radiation is subtracted to obtain a net result due only to 

radiation from occupational exposure (Bradley et al. 1994, 1995; Walker 1996). Background 

thermoluminescence is the summation of “system background” and “environmental background.” 

System background consists of thermoluminescence inherent in the instrumentation (e.g. noise due to 

photomultiplier tubes), and is determined from re-read values for daily measurements of calibration 

cards. Environmental background consists of accumulated exposures from cosmic and terrestrial 
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background sources, and is determined from trip control dosimeters that accompany each batch of 

field dosimeters to and from each SNL site. In addition to monitoring exposures during shipping, trip

control dosimeters remain stored at daily check-in and check-out locations while at a given SNL site 

so that only site-specific background radiation is accumulated. The locations of badge storage areas 

for SNL/CA have not been identified. 

Routine field dosimeters are assigned and issued by Department Managers and are subject to 

specific protocols for on site use, as well as subsequent handling and shipment to the Personnel 

Dosimetry Division in the New Mexico office for processing. Personnel issued routine field dosimeters 

are required to have documented training with respect to dosimeter use, as well as general radiation 

safety training. 

Dosimeter shipping and handling quality control protocols include the use of special zippered 

envelopes for transport, dosimeter issue and return lists, chain-of-custody control documentation, trip 

control cards, express (overnight) shipping, and upon arrival at the NM Processing Center all field 

dosimeters are inspected for physical damage and card assignments are verified (Ward et al. 1994). 

Nonroutine field dosimeters for temporary employees and visitors are issued by the Security Patrol 

Division at a specified site check in locations. The Security Inspector is responsible for insuring 

proper documentation of personnel being issued nonroutine field dosimeters, and for instruction in 

their use while on site. With the apparent exception of dosimeter exchange periods, protocols for 

nonroutine field dosimeter handling and shipment to the NM Processing Center are the same as those 

employed for routine field dosimeters. 

Nonroutine field dosimeters for special radiation work cases are requested from Division 7715 and 

issued by Department Managers for any personnel involved in such work (Walker 1995). Respective 

examples of reporting for this type of nonroutine evaluation are provided in Attachment B (Figures B-2 

and B-3). 

Dosimetry results greater than 1 rem, questionable results, suspected misuse, processing difficulties, 

and non-returned field dosimeters will result in a formal investigation to resolve the issue as well as 

document and justify the dose equivalent assigned to the respective individual’s dosimetry record. 

Radiation work restrictions may be imposed pending resolution of an investigation. Example 

documentation of a non-returned dosimeter investigation and assignment of missed dose are 

provided in Attachment B (Figures B-4 to B-6). 

The overall performance of SNL’s Personnel Radiation Dosimetry Program (PRDP) is monitored 

quarterly by a blind audit (Ward et al. 1994), but it is unclear as to who performs these audits. 

In addition to personnel radiation dosimetry, SNL/CA conducts a Special Dosimetry Program, which 

includes work area monitoring (Ward et al. 1992; SNL/NM 1995b). These dosimeters are also 

considered “nonroutine” and respective exchange periods can vary depending on the intended 

monitoring purpose. Request forms and example results report for work area dosimeters are provided 

in Attachment B (Figures B-7 and B-8). Shipping and handling procedures are identical to those for 

personnel field dosimeters and the dose equivalent results are also recorded in SANDOS (SNL/NM

1995b). 

6.4.2 Site Dosimetry Technology

The historical progression of dosimeter technologies used for external dose monitoring at SNL/CA is 

shown in Table 6-1. As broken down and discussed in the following sub-sections, site dosimetry 
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technologies have been partitioned into two historical categories (before 1989 and after 1989). This is 

because for the purposes of this SPD, more complete and detailed information is available to describe 

the dosimetry program after 1989.

Table 6-1. Documented dosimetry technologies and dose quantities. 

6.4.2.1 Dosimetry Technology Before 1989

As suggested by certain documents cited in Table 6-1, dosimeters worn by SNL/CA personnel 

between 1956 and the early 1970s consisted of two-element Du Pont Type 554 beta/photon film along 

with a special nuclear track emulsion, type A film. Film badges consisted of four windows with one 

open and the others filtered with various densities of Pb, Cd, and Al (RDC 1963). Apparently, neutron 

film was exchanged on a monthly basis, while a quarterly exchange period was generally 

implemented for beta/photon film. There is a discrepancy in the available record as to exactly when 

SNL/CA discontinued using film dosimeters and began using the thermoluminescent dosimeter (TLD). 

Dosimetry History 1993 (Wright 1993) indicates that this switch occurred in about 1966. Lovell 1966 

indicates continued use of film in 1966, with plans to continue that use. Lovell (1983) estimates the 

use of TLDs began in about 1968. Radiation Detection Company (1969) indicates that they were still 

processing badges for SNL/CA as late as 1970, but it is not clear whether TLD technology was in use 

at the time. It is possible that the switch to TLD’s coincided with a switch of dosimetry service 

providers from RDC to RESL in Idaho Falls, Idaho (Wright 1993), or that there was some overlap in 

usage of film and TLDs. Due to the uncertainty, it is favorable to the claimant, in cases where missed 

dose is estimated, to assume that film dosimeters were used at SNL/CA until about 1972. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 57 of 95

technologies have been partitioned into two historical categories (before 1989 and after 1989). This is 

because for the purposes of this SPD, more complete and detailed information is available to describe 

the dosimetry program after 1989.

Table 6-1. Documented dosimetry technologies and dose quantities.

6.4.2.1 Dosimetry Technology Before 1989

As suggested by certain documents cited in Table 6-1, dosimeters worn by SNL/CA personnel 

between 1956 and the early 1970s consisted of two-element Du Pont Type 554 beta/photon film along 

with a special nuclear track emulsion, type A film. Film badges consisted of four windows with one 

open and the others filtered with various densities of Pb, Cd, and Al (RDC 1963). Apparently, neutron 

film was exchanged on a monthly basis, while a quarterly exchange period was generally 

implemented for beta/photon film. There is a discrepancy in the available record as to exactly when 

SNL/CA discontinued using film dosimeters and began using the thermoluminescent dosimeter (TLD).

Dosimetry History 1993 (Wright 1993) indicates that this switch occurred in about 1966. Lovell 1966 

indicates continued use of film in 1966, with plans to continue that use. Lovell (1983) estimates the 

use of TLDs began in about 1968. Radiation Detection Company (1969) indicates that they were still 

processing badges for SNL/CA as late as 1970, but it is not clear whether TLD technology was in use 

at the time. It is possible that the switch to TLD’s coincided with a switch of dosimetry service 

providers from RDC to RESL in Idaho Falls, Idaho (Wright 1993), or that there was some overlap in 

usage of film and TLDs. Due to the uncertainty, it is favorable to the claimant, in cases where missed 

dose is estimated, to assume that film dosimeters were used at SNL/CA until about 1972.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 57 of 95

technologies have been partitioned into two historical categories (before 1989 and after 1989). This is 

because for the purposes of this SPD, more complete and detailed information is available to describe 

the dosimetry program after 1989.

Table 6-1. Documented dosimetry technologies and dose quantities.

6.4.2.1 Dosimetry Technology Before 1989

As suggested by certain documents cited in Table 6-1, dosimeters worn by SNL/CA personnel 

between 1956 and the early 1970s consisted of two-element Du Pont Type 554 beta/photon film along 

with a special nuclear track emulsion, type A film. Film badges consisted of four windows with one 

open and the others filtered with various densities of Pb, Cd, and Al (RDC 1963). Apparently, neutron 

film was exchanged on a monthly basis, while a quarterly exchange period was generally 

implemented for beta/photon film. There is a discrepancy in the available record as to exactly when 

SNL/CA discontinued using film dosimeters and began using the thermoluminescent dosimeter (TLD).

Dosimetry History 1993 (Wright 1993) indicates that this switch occurred in about 1966. Lovell 1966 

indicates continued use of film in 1966, with plans to continue that use. Lovell (1983) estimates the 

use of TLDs began in about 1968. Radiation Detection Company (1969) indicates that they were still 

processing badges for SNL/CA as late as 1970, but it is not clear whether TLD technology was in use 

at the time. It is possible that the switch to TLD’s coincided with a switch of dosimetry service 

providers from RDC to RESL in Idaho Falls, Idaho (Wright 1993), or that there was some overlap in 

usage of film and TLDs. Due to the uncertainty, it is favorable to the claimant, in cases where missed 

dose is estimated, to assume that film dosimeters were used at SNL/CA until about 1972.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 58 of 95

As reported in Wright (1993), up until 1982, SNL/CA’s TLD technology consisted of a 2-chip badge 

with only one filter type [30 mg/cm2 total filtration according to Wright (1981b)] and as such, could not 

discriminate between different radiation types or energies. However, the actual dosimetry records for 

this period report doses in penetrating and nonpenetrating dose categories, including beta, photon, 

and neutron radiations. If the TLD’s during this period could not directly measure these different dose 

quantities, it is not clear if (or how) adjustments to recorded dose were made. Wright (1981b) 

suggests that nonpenetrating dose may have also been assigned as penetrating dose, but the 

document does not provide any indication of how different radiations, including neutron doses, were 

assessed. In 1982, SNL/CA switched to a three-element TLD from Eberline that apparently could 

better discriminate between different radiation types and provide directly measurable indicators of 

these various dose quantities. SNL/CA continued using the Eberline dosimeter until about 1989.

6.4.2.2 Dosimetry Technology After 1989

Since 1989, radiation dose monitoring at SNL/CA has been based primarily on the use of Harshaw 

TLD’s. The Harshaw 8800 series TLD systems were the first to be DOELAP accredited at SNL/NM, 

and are the predominant systems used at all Sandia sites. Harshaw 4000 series TLD’s are

apparently utilized as well (Stanley et al. 1987; SNL/NM 1996a), though little information is currently 

available on the specific application and extent of use of this series. There is some evidence that the 

4000 series may have been used, at least until 1996, for extremity dose monitoring (Walker 1997a). 

In 1997, SNL began using Harshaw/Bicron EXTRAD dosimeters for extremity monitoring. There is 

some indication that in addition to routine dosimeters for regular workers in the radiography facility, 

pocket dosimeters were used for nonroutine personnel accessing that facility (Lovell 1984a, 1984b). 

Because the Harshaw 8800 system is a primary method for external personnel dosimetry monitoring 

at SNL/CA, it is pertinent to describe some basic technologies involved. Model 8801 and 8802 TLD 

cards consist of four thermoluminescence elements encapsulated between two thin sheets of Teflon, 

all of which is sandwiched between aluminum jackets. The jackets have four holes positioned over 

the thermoluminescence elements so that they can be heated by hot N2 gas in the card reader. For 

operational efficiency, SNL color-codes TLD card edges according to use (e.g. calibration cards have 

green edges, quality control cards have red edges, field cards for worker monitoring have no added 

coloring). 

The assembled dosimeter consists of the TLD card positioned inside a Model 8812 card holder. The 

cardholder front has radiation modifying filters to evaluate radiation and dose equivalent quantities as 

shown in Figure 6-2. For simplicity, the manufacturer refers to the entire assembly as a Model 8812 

dosimeter, while SNL refers to this assembly as the SNL dosimeter. 

The Harshaw 8800 card reader is an automated system in which up to 1400 cards can be loaded at a 

time and read automatically. Barcode ID information printed on each card is automatically recorded 

by the card reader before heating. Precisely controlled heating causes the TLD elements to give off 

light in proportion to the amount radiation they have received. The light produced is converted to an 

electrical signal by a photomultiplier tube. The relative strength of the electrical signal is measured in 

units of charge (nanocoulombs) creating “glow curves” that are then analyzed with respect to card 

reader calibration parameters. Non-contact heating of the cards with hot nitrogen gas improves 

dosimeter re-use, durability, and improves glow-curve reproducibility. Raw data from the card reader 

is acquired, analyzed, and stored with TLD Radiation Evaluation and Management System software 

resident on a desktop computer. Card reader results are converted into dose equivalent by a complex 

algorithm developed specifically for the Harshaw 8800/8812 system and SNL sites. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 59 of 95

Figure 6-2. Harshaw Model 8812 dosimeter assembly, element 

specifications, and respective radiation, and dose equivalent quantities 

measured by Rhea and Bradley (1990). 

Along with explicit protocols for card reader calibration, DOELAP accreditation is an on-going process 

that includes repeated algorithm validation and blind audits as part of an overall data quality 

assurance program. Calibration and algorithm protocols are described in more detail in subsequent 

sections of this report. 

6.4.3 Calibration

6.4.3.1 Calibration for Dosimeter Technologies Before 1989

Calibrations for film dosimeters processed by RDC in the 1960s involved exposing film in SNL/CA 

badges to known Co-60 gamma and X-ray fields (effective X-ray energies were 35 keV and 90 keV) 

based on prior calibration data for RDC badges (RDC 1963). SNL/CA badges had Al, Cd, and Pb 

filters, each of which was cross-calibrated against similarly filtered RDC badges in parallel runs. 

Currently, no information is available regarding the calibration of TLD badges used by SNL/CA 

between 1971 and 1988 (the period in which RESL performed dosimetry services).

6.4.3.2 Calibration for Dosimeter Technologies After 1989

Harshaw Model 8800 TLD Card readers are setup and maintained in accordance with manufacturer 

recommendations. Figure 6-3 shows a calibration procedure flowchart (Bradley et al. 1995) along 

with an example calibration checklist, glow curves, and calibration results output (SNL/NM 1996b).
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 60 of 95

Figure 6-3. Example Harshaw 8800 card reader calibration procedures, forms, analysis, and output. 

Specific details of the Harshaw 8800/8812 system calibration procedures can be found in Bradley et 

al. (1993) and Rhea and Bradley (1990). Procedures appear different for calibration of the Harshaw 

4000 system (SNL/NM 1996a) and an example calibration form is shown in Figure 6-4. No official 

manuals or procedural documents have been found concerning Harshaw 4000 dosimetry system 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 61 of 95

calibration or system applications, nor has any similar documentation been found regarding the 

Harshaw/Bicron EXTRAD system.

Figure 6-4. Example calibration form for the Harshaw 

4000 dosimetry system.

6.5 DOSE CALCULATION AND REPORTING

6.5.1 Dose Calculation and Reporting Before 1989

As shown in Table 6-1, reported doses for SNL/CA employees, contractors, and visitors before 1989 

include penetrating and nonpenetrating categories for beta, photon, and neutron radiations. For the 

period in which film was used for measuring these quantities, it is assumed that the degree of film 

darkening under the various windows (filtered to different degrees) was used to estimate doses within 

these various reporting categories by comparison against calibration curves developed for each batch 

of film. No information is currently available with respect to minimum detection limits (MDLs), 

adjustments to recorded dose, uncertainty or bias, or adjustments to recorded dose during this period 

of film dosimeter use. 

As mentioned previously, the 2-chip, single filter TLD dosimeters in use from about 1972 to 1982 

apparently could not directly measure each of the various dose quantities reported in the record for 

this period. No definitive information has been located with respect to adjustments to recorded dose, 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 62 of 95

nor has any information been found regarding dose calculation algorithms, system MDLs, or 

uncertainty/bias doses during this period.

Like the earlier period in which 2-chip TLD’s were used at SNL/CA and processed by RESL in Idaho, 

there is no documentation regarding dose calculation algorithms, MDLs, adjustments to recorded 

dose, or uncertainty/bias for the 3-element Eberline TLD badges used from 1982 to 1988. 

6.5.2 Dose Calculation and Reporting After 1989

Doses to personnel and work area TLD cards are calculated per SNL/NM procedure PRDP-14-01

(SNL/NM 1991). Algorithms used for calculating dose equivalents are specific to each particular 

dosimeter system used. Both shallow (0.007 cm) and deep (1.0 cm) dose equivalents are measured, 

calculated, and reported. Although dose to lens of the eye is not measured directly, it can be 

calculated indirectly from routine dosimetry results. Due to the complexity of this calculation, 

however, SNL does not routinely calculate dose to lens of the eye, but instead uses an annual 

cumulative shallow dose equivalent of 1.5 rem as a benchmark value that triggers respective manual 

calculation and reporting of results (Walker 1997b). No information has been found on extremity 

calculation and reporting. 

There are specific circumstances and respective protocols for making adjustments to an individual’s 

recorded dose or estimating doses when reliable measurements are not available (Potter et al. 1993; 

SNL/NM 1995a). Some primary examples are shown in Table 6-2. Calculation or database

programming errors are also included, but don’t necessarily involve an adjustment. For example, in 

1997 a SANDOS programming oversight was discovered that had resulted in incorrect background 

subtractions since 1992, resulting in slightly overestimated respective doses (Walker 1997c). The 

programming correction was made when discovered, but no adjustments to the affected prior records 

were implemented since the slight overestimations were considered conservative. 

Table 6-2. Circumstances and protocols for dose adjustment or estimation. 

Calculated Dose Level Dose Assignment

Calculated dose* < 10 mrem for deep or 

neutron dose equivalent. Report 0 rem

Calculated dose* < 35 mrem for deep or 

neutron dose equivalent.

Set shallow dose equivalent equal to 

deep dose equivalent

Calculated neutron dose* > 100 mrem 

(not applicable to work area dosimeters)

Contact user to verify neutron code, 

reassess and adjust doses if justified

Missed Dose (non-returned or lost 

dosimeter)

Estimate dose based on prior history, 

dose to other personnel, or reconstruct 

based on work area dosimetry results 

Suspected problem with analysis 

(damaged card, processing difficulties, 

etc.)

Estimate dose based on prior history, 

dose to other personnel, or reconstruct 

based on work area dosimetry results 

Calculation or related software 

programming mistakes

Correct calculation or programming 

errors and adjust records as warranted 

and justified

*Effective dose equivalent calculated per PRDP-14-01
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 63 of 95

The angular dependence of the Harshaw 8812 TLD was studied by SNL as part of the DOELAP 

accreditation process (Friedman et al. 1991). A batch of 8812 TLD’s was sent to Battelle Pacific 

Northwest Laboratory (PNL), which maintains an National Institute of Standards and Technology 

(NIST)-traceable irradiation and measurement quality assurance program, to assess the angular 

dependence of respective dosimeter readings for various exposure geometries. The results (Figure 

6-5) led to a conclusion that the horizontal and vertical angular dependence of the TLD system used 

by SNL are acceptable and “comparable to or better than results for modern dosimetry systems.”

Figure 6-5. Angular dependence testing results for the Harshaw 

8800/8812 dosimetry system. Results in each category are normalized 

(apparently to the average result from control cards exposed under the 

normal, perpendicular geometry).

The lower limits of detection (LLDs) for the Harshaw Model 8802 dosimeter were evaluated by SNL in 

1997. Both monthly and quarterly exchange periods were tested within various DOELAP exposure 

categories and the results are shown in Table 6-3. An earlier LLD study, conducted in 1990 for the 

Model 8801 dosimeter, showed shallow dose equivalent LLD of about 31 mrem. The only difference 

between the Model 8802 and 8801 dosimeters is that the 8802 has a slightly thicker shallow chip. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 64 of 95

Because the shallow dose LLD for the newer 8802 cards is 10 mrem or less across all exposure 

categories, SNL uses 10-mrem LLD value in SANDOS for shallow dose equivalent. 

Table 6-3. Monthly and quarterly LLDs for the Harshaw Model 

8802 dosimeter card (adapted from Walker et al. 1997). 

Exposure Radiation Test Monthly LLD Quarterly LLD

Category Type(s) Depth (rem) (rem)

I Accident X-rays Deep 0.003 0.004

II Accident Gammas Deep 0.003 0.005

IIIA General X-rays Shallow 0.004 0.008

Deep 0.004 0.006

IV Cs-137 Gammas Shallow 0.003 0.007

Deep 0.003 0.004

VA Beta Particles Shallow 0.003 0.007

VI Moderated Neutrons Deep 0.003 0.005

VII (mixtures) 

III + IV Shallow 0.005 0.010

Deep 0.004 0.006

III + VA Shallow 0.004 0.008

Deep 0.005 0.008

III + VI Deep 0.003 0.004

IV + VA Shallow 0.004 0.008

Deep 0.004 0.006

IV + VI Deep 0.003 0.004

6.5.3 Exposure Energy Spectra

Dose reconstruction under the NIOSH program (NIOSH 2002) requires estimates of exposure 

percentages within specific energy bands for each type of radiation as follows:

Photons

• < 30 keV

• 30 – 250 keV

• > 250 keV

Betas

• < 15 keV

• > 15 keV

Neutrons

• < 10 keV

• 10 – 100 keV

• 100 keV – 2 MeV

• 2 – 20 MeV

• > 20 MeV

For external exposures, betas with energies < 15 keV are not applicable. Table 6-4 provides 

estimates of percentages of radiation energies within each relevant category, broken down by process 

within each facility. Most of these estimates were obtained from SNL/CA site personnel. Other 

sources and assumptions are provided as applicable in the table footnotes. The Tritium Research 

Laboratory is not listed because the primary radionuclide was tritium with beta energies < 15 keV. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 65 of 95

Table 6-4. Selection of radiation energies and percentages. Process description Radiation type Energy selection (keV) Percentagea

Weapons Laboratory Facility Complex 

(buildings 910, 912, 913, 914, 916, 918)

Years: 1958 - 1998 

Test/repair of neutron and X-ray 

detectors (neutron and X-ray 

generators)

Photon

Neutron

>250

10-100

100-2,000

2,000-20,000

100

5 

5 

90

Wet machining of DU Beta 

Photon

>15

30-250

>250

100

50b

50b

Radiography for weapons 

components

Beta 

Photon

>15

30-250

>250

100

50b,c

50b,c

Radiography for materials science 

studies (X-ray diffraction operations 

later moved to building 941) 

Beta

Photon

>15

<30

30-250

>250

100

40

30

30

3H storage studies Photon 30-250

>250

50b,d

50b,d

Ion beam analysis of materials Photon 30-250

>250

30

70

Radiflo leak tests

 

Beta 

Photon

>15

30-250

>250

100

10

90

Radiography 

(building 923) 

Years: unknown – early 1990’s

Radiography using X-rays, gamma 

rays, neutrons, alpha and beta 

particles

Beta

Photon

Neutron

>15

<30

30-250

>250

10-100

100-2,000

2,000-20,000

100

 

5 

45

50

5 

70

25

Micro and Nano Technologies Laboratories

(buildings 941, 942, and 943)

Years: unknown – current

Radiography for materials science 

studies (radiological materials in 

building 941 only)

Beta

Photon

>15

<30

30-250

>250

100

40

30

30

Explosives and Environmental Testing Complex

(buildings 955, 956, 966, 972, 974, 976, 977, 978, 979, 981, 983)

Years: 1958 – current

Environmental testing of mock-up 

weapons and components (DU)

Beta

Photon

>15

30-250

>250

100

50

50

Storage Facilities

(buildings 921, 927, 961, 982)

Years: unknown – current

Storage and packaging of waste 

materials

Beta

Photon

>15

30-250

>250

100

70

30

a. Estimated primarily by site personnel. 

b. Based on favorable to claimant default assumptions provided for DU in ORAUT 2006b. 

c. Assumes radiography primarily associated with weapons mock-ups and DU. 

d. Assumes only external exposures would be associated with depleted uranium beds in tritium storage facility. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 66 of 95

6.5.3.1 Neutron Dose Conversion Factors

As described in ORAUT-OTIB-0055 (ORAUT 2006c), adjustments to neutron dose are necessary to 

account for changes in quality factors between historical and current scientific guidance. Using the 

methodology indicated in ORAUT-OTIB-0055, adjustment factors were determined for the various 

energy groups at the two SNL/CA facilities in which neutron exposures were possible. Table 6-5 

shows multiplier values to use for dose reconstruction.

Table 6-5. Neutron dose conversion factors for applicable SNL/CA facilities. 

Facility

Neutron 

energy 

intervals

NCRP 38 

Quality 

factor

ICRP 60 

weighting factor Dose fractiona Corrected dose 

equivalent multiplierb

Weapons 

laboratory 

complex

10 - 100 keV 5.38 10 0.05 0.093

0.1 - 2.0 MeV 10.49 20 0.05 0.095

2.0 - 20.0 MeV 7.56 10 0.90 1.190

Radiography 10 - 100 keV 5.38 10 0.05 0.093

0.1 - 2.0 MeV 10.49 20 0.70 1.335

2.0 - 20.0 MeV 7.56 10 0.25 0.331

a. From Table 6-4. 

b. Multiply reported dose by these factors to determine the corrected neutron dose equivalent for each 

applicable neutron energy interval. 

6.5.4 Missed Dose 

Given the lack of documentation regarding MDLs for dosimeter systems for the period prior to 1989, it 

is necessary to estimate dosimeter system MDLs and missed dose. During this early period, it is 

reasonable to assume that MDLs for the various systems were similar to contemporary technologies 

used at other AEC or DOE facilities during corresponding periods. Maximum annual missed doses 

are estimated using exchange periods and MDL/2 as recommended by NIOSH (NIOSH 2002). After 

1989, MDL information presented in the previous section is used to estimate missed dose using 

exchange periods and MDL/2. Table 6-6 lists documented or estimated MDLs and maximum missed 

doses for each dosimetry technology used by SNL/CA. 

Table 6-6. Estimated maximum annual missed photon, beta, and neutron dose. 

Period of use Dosimeter

MDL 

(mrem) Exchange frequency

Maximum 

annual missed 

dose (mrem)a

1956 - 1959 Two-element Beta/Photon film 30b Monthly; Quarterly 60; 20

1959 - 1971 Two-element Beta/Photon film 30b Monthly; Quarterly 60; 20

1956 - 1971 Neutrons (NTA film) 50b Monthly; Quarterly 60; 20

1972 - 1982 2-chip TLD 20c Semi-annual 15

1982 - 1988 3-chip Eberline TLD 20c Annual; Semi-annual 8; 15

1972 - 1988 Neutrons (2-chip & 3-chip TLD 

systems)d 20c Annual; Semi-annual 8; 15

1989 - 1990 Multi-element Harshaw TLD 10 Quarterly 20

1991-current Multi-element Harshaw TLD 10 Quarterly 20

1989-current Neutrons (Harshaw TLD 

systems) 5 Quarterly 10

a. Maximum annual missed dose calculated using N x MDL/2 from NIOSH (2002). 

b. Estimated MDL based on contemporary film systems of similar sensitivity (ORAUT 2004b and ORAUT

 2005d). 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 67 of 95

c. Estimated MDL based on contemporary TLD systems of the period (ORAUT 2005d, ORAUT 2006d, and 

 ORAUT 2006e). 

d. Although neutron doses are reported for the 2-chip system, no documentation of the neutron dosimetry

 methodology employed has been found. 

 

6.5.5 Organ Dose Conversion Factors 

During the period in which film dosimeters were used at SNL/CA, dosimeters were calibrated in units 

of roentgen (R). No information is available regarding calibration of TLDs for the period between 1972 

and 1988, but it is favorable to claimants to assume TLDs were calibrated in units of exposure (R). 

After dosimetry services were transferred to SNL/NM to gain DOELAP accreditation and consolidate 

all SNL dosimetry programs, the Personal Dose Equivalent [Hp(10)] should be used. Table 6-7 

shows dose units to use for organ dose conversion factors. 

 Table 6-7. Dose units for organ dose conversion factors. 

Years

Photon dose units for 

use with organ dose 

conversion factors

1956 - 1959 R

1959 - 1971 R

1972 - 1982 R

1982 - 1988 R

1989 - 1990 Hp(10)

1991-current Hp(10)

6.5.6 Uncertainty

Given the lack of specific technical information obtained with respect to dosimetry systems for much 

of SNL/CA’s history, it is necessary to estimate respective measurement uncertainty based on 

reported values for contemporary systems in use at other facilities. Table 6-8 shows estimates based 

on the technical basis document for the Hanford Site (ORAUT 2006d), with additional values for NTA 

film adapted from ORAUT 2006e. Some general analogies can be drawn between Hanford and 

SNL/CA in terms of dosimetry technologies employed during various time periods of interest.

Table 6-8. Bias and uncertainty estimates (adopted from ORAUT 2006d).

Dosimeter Period of use Bias magnitude and range Uncertainty factors

Overall biasa Range in bias Systematicb Randomc

Two-element film 1956 - 1971 1.02 0.86 – 1.12 1.1 1.4

Two-chip TLD 1972 – 1982 1.12 1.04 – 1.16 1.05 1.2

Three-chip TLD 1982 - 1988 1.01 0.95 – 1.05 1.05 1.2

Multi-element TLD 1989 - current 1.00 0.95 – 1.05 1.05 1.2

NTA filmd 1956 - 1971 1.5 0.5 – 1.5 1.5

a. Divide recorded dose by bias value to determine deep dose. 

b. Systematic uncertainty related to lack of knowledge concerning energy distributions and geometries. 

c. Random uncertainty due to variation across workers, energy levels, and geometries. 

d. The most favorable to claimant estimates from the various energy ranges in ORAUT 2006e are applied to all 

 applicable energy ranges. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 68 of 95

7.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this SPD, 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 provided to 

identify the source and justifications for each associated item. Conventional references are listed in 

the next section, linking data, quotations, and other information to documents available for review on 

the NIOSH Project computer network.

[1] Thomas, Elyse. ORAU Team. Principal Medical Dosimetrist. October 2006. The X-ray 

record form included in claim files has a place for the technician to record the projection, 

mA, kVp, distance, and time. For PA chests, the mA is almost always listed as 5, the 

distance is listed as 72", and the kVp is usually around 70. Exposure times are not usually 

listed. It is assumed that the person completing this part of the X-ray record form was 

recording the mAs, not the mA, primarily since radiographic machines are not designed to 

operate at such low mA settings, and also because time settings were not recorded. Five 

mAs would not be unusual for a PA chest exposure.

[2] Lopez, Theresa. ORAU Team. Senior Toxicologist. September 2006. 

Lumber spine X-rays were performed at hire from 1956 through 1971 

as evidenced by medical records in claim files.

[3] Lopez Theresa. ORAU Team. Senior Toxicologist. September 2006. 

Per review of X-ray records and at the direction of ORAU in comments dated October 

2006, the X-ray machines have been assumed to be single-phase.

[4] Lopez Theresa. ORAU Team. Senior Toxicologist. September 2006.

Lumbar spine X-rays were performed at hire from 1956 through 1971 

as evidenced by medical records in claim files.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 69 of 95

REFERENCES

Adolphson, D. R., 1972, “Machining Test Specimens in Uranium Alloys”, Memo to J. F. Bryson, April 

7, 1972, Sandia Laboratories, Livermore, California. 

AEC (U.S. Atomic Energy Commission), 1963, "Standards for Radiation Protection," Chapter 0524, 

AEC Manual, Washington D.C., August 12. [SRDB Ref ID: 13037]

Ball, M. D. Stephanie, G. Murphy, M. J. Patton, K. O. Reil, R. L. Schwoebel, 1995. Review Radiation 

Protection Program and Invalid Bioassays by CEP, Sandia National Laboratories, 

Albuquerque, NM. (as cited in ORAUT-TKBS-0037-5). 

Bradley, E.W.; Loudermilk, F.R.; Ward, D.C.; Stanley, A.L. 1993. Source normalization and 

traceability of the HF-870 90Sr/90Y source for the Harshaw Model 8800 TLD reader system. 

Sandia National Laboratories, Personnel Radiation Dosimetry Program. Albuquerque, NM.

Bradley, E.W; Loudermilk, F.R.; Ward, D.C.; Stanley, A.L. 1994. Harshaw Dose Calculation 

Algorithm. Sandia National Laboratories, Personnel Radiation Dosimetry Program, Health 

Instrumentation Division, 7715. Albuquerque, NM. 

Bradley, E.W.; Loudermilk, F.R.; Zamora, J.A.; Walker, S.W. 1995. Harshaw 8800 TLD Card Reader 

– Daily Calibration (Procedure RPXD-13-06 Manual). 

Bryson, J. F. 1972. Machining Test Specimens in Uranium Alloys, internal memo to W. T. Ryan, April 

12, 1972, Sandia Laboratories, Livermore, California.

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SNLL (Sandia National Laboratories Livermore), 1983b, Pelletron 1MV Tandem Accelerator, SOP 

NO. 851, May 10, Sandia National Laboratories, Livermore, California.

SNLL (Sandia National Laboratories Livermore), 1984, Environmental Summary, Calendar Year 1984. 

Sandia National Laboratories, Livermore, California (no publication date given).

SNL/NM / ES&H. 1992. Performance Indicator Report, 1st quarter CY 1992. (First indication of 

SNL/CA being evaluated in conjunction with SNL/NM personnel).

SNL/NM, 1991, PRDP-14-01. Reference to be inserted.

SNL/NM, 1993a, Chain of Custody Records (Collection of chain of custody records for 1993 for 

samples sent to Controls for Environmental Pollution.)

SNL/NM, 1993b, Miscellaneous Uranium Bioassay Documents (Collection of handwritten calculations 

and bioassay result letters to bioassay subjects).

SNL/NM. 1995a. Example Documentation of Non-returned Dosimeters. (Example investigation 

reports, questionnaire documents, and dose histories related to a non-returned dosimeter 

investigation). 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 80 of 95

SNL/NM. 1995b. Work Area Dosimetry Processing. (Procedures and request forms for work area 

dosimetry at SNL sites). 

SNL/NM. 1996a. Example TLD Calibration Form. (Example calibrations and respective forms for 

Harshaw 4000 series dosimetry system). 

SNL/NM. 1996b. Card Reader Calibration Records. (Example calibrations and respective forms for 

Harshaw 8800 series dosimetry system). 

SNL/NM. 1997. Monthly ALARA Radiation Dosimetry Report 12-31-97.

SNL (Unknown). 1994. Internal Dosimetry Questions Answered for Sandia National Laboratory. QSNL. WP Predecisional Draft. December 9.

Stanley, A. 1991. Basis for Wearing One Dosimeter. (Memo discouraging use of anything other than 

DOELAP approved SNL dosimeters at all Sandia sites). 

Stanley, A.L. et al. 1987. Third Quarter Progress Report, July-September, 1987 (Series of progress 

report memos in SNL/NM TBD documentation files entitled “10-27-87 Dosimetry Equipment”). 

Thompson, B.C. et al. 1991. Radiation Protection Overview Program. (Various sections of the 

overall radiation protection program for Sandia National Laboratories, circa 1991). 

Tipton, W. J., 1977, An Aerial Radiological Survey of the Lawrence Livermore Laboratory (Livermore, 

California), EGG-1183-1693, EG&G/Energy Measurements, Las Vegas, NV, October.

Turner, J. E., 1995, Atoms, Radiation and Radiation Protection, 2nd Edition, John Wiley & Sons, New 

York, NY.

Ullrich, R. A., 2003, Cold War Context Statement, Sandia National Laboratories, California Site, 

SAND2003-0112, Sandia National Laboratories, Albuquerque, New Mexico and Livermore, 

California, January.

Vosburg, S. K. 1993. Sandia National Laboratories, Statement of Work for Bioassay Radiochemical 

Analysis Laboratory. Cover letter date July 16. 

Walker, S. 1995. Example of Nonroutine Personnel Dosimeter Evaluations, 1995. (Various examples 

of nonroutine personnel dosimeter evaluations, including special radiation work cases).

Walker, S. 1996. Environmental background accumulation rates for SNL/NM, SNL/CA, and SNL/NV. 

(Internal SNL memo containing reported background accumulation rate values for subtraction 

from TLD values). 

Walker, S. 1997a. Interim Processing Steps for EXTRAD and Model 8807 Dosimeters.

Walker, S. 1997b. Lens of the Eye Dose Equivalent Monitoring at SNL. (Memo documenting 

justification for shallow dose benchmark approach to calculating doses to lens of the eye). 

Walker, S. 1997c. SANDOS Background Subtraction Problem 4-8-97. (Memo documenting a 

discovered programming error affecting dose history records since 1992, and justification for 

not adjusting those records). 

Walker, S.W.; Loudermilk, F.R.; Thompson, D.J. 1997. Lower Limit of Detection (LLD) for the 

Harshaw 8802 TLD. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 81 of 95

Wall, W.R., 1981, Tritium Control and Accountability Instructions. SAND80-8057. Sandia National 

Laboratories, Livermore, California. March.

Wallace, S. A., 1988, Current Status of 1985 Headquarters (HQ) Nuclear Facilities Appraisal (Ref. 

Audit E431), Memorandum to W. V. Ormand with attached 1985 HQ appraisal of Sandia 

National Laboratories, Livermore, Sandia National Laboratories, Albuquerque, New Mexico, 

September 9, 1988. 

Ward, D.C.; Loudermilk, F.R.; Stanley, A.L. 1994. External Dosimetry Program. Sandia National 

Laboratories, Albuquerque, NM. (Manual of official procedures for TLD program circa 1994 for 

SNL sites including NM, CA, and NV). 

Ward, D.C.; Loudermilk, F.R.; Thompson, D.J.; Stanley, A.L. 1992, Special Dosimetry Program.

(Special Dosimetry monitoring procedures for nuclear accident, environmental radiation, and 

work area dosimetry). 

Webster, E. W., and O. E. Merrill, 1957, “Measurements of Gonadal Dose in Radiographic 

Examinations,” New England Journal of Medicine, 257:811-819.

Willhoite, S. B., 1979, Determination of Gamma and Neutron Yearly Doses at the Building 212 BusStop, Lawrence Livermore Laboratory, Livermore, California, December.

Wright, D. A. 1979a. Sandia Livermore’s Bioassay Program from (sic) Depleted Uranium. May 14. 

Wright, D. A. 1979b. Memo to H. M. Patterson – 8214. Radiation Workers at SLL. December 20.

Wright, D.A., 1981a, Tritium Research Laboratory Safety Analysis Report. SAND81-8234, Sandia 

National Laboratories, California, August.

Wright, D. A. 1981b. Radiation Dosimetry at Sandia National Laboratories Livermore, Memo to P. K. 

Lovell dated November 6, 1981, Sandia National Laboratories, California.

Wright D. A. and R. S. Hafner. 1981. TRL Health Physics Summary, First Quarter, 1981. Sandia 

National Laboratories. Albuquerque, New Mexico, Livermore, California.

Wright, D. 1993, History of Radiation Dosimetry at SNL/CA as Remembered by D.A. Wright, 8641, 

Memo to Lydia Perez, 7715, June 1. 

Wright, D. A. 1998 Memo to Debbie Miller DOE/AL. February 19. 

Wright, D. A., 2006, Notes from Visit to SNL/CA on 4/12/06, Interview by L. McDowell-Boyer, April 12, 

2006. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 82 of 95

GLOSSARY

alpha particles

Positively charged particles of discrete energies emitted by certain radioactive 

materials; alpha particles usually expend their energy in short distances and will not 

usually penetrate the outer layer of skin; they are a significant hazard only when taken 

into the body where their energy is absorbed by tissues.

becquerel (Bq)

A special unit of activity. One becquerel equals one nuclear transition per second. 

beta particles

Negatively charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Most (if not all) of the 

direct fission products emit beta particles. Physically, the beta particle is identical to 

an electron moving at high velocity.

curie

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

depleted uranium (DU)

A very dense by-product of the enrichment of natural uranium, which contains about 

99.8% 238U, 0.2% 235U and 0.001% 234U by mass. 

dosimeter

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

exposure 

As used in the technical sense, exposure refers to a measure expressed in roentgens 

(R) of the ionization produced by photons (i.e., gamma and X-rays) in air. 

film

In general, a "film packet" that contains one or more pieces of film in a light-tight 

wrapping. When developed, the film has an image caused by radiation that can be 

measured using an optical densitometer.

film dosimeter

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

fission

The splitting of a heavy atomic nucleus, accompanied by the release of energy.

fissionable

Material capable of undergoing fission.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 83 of 95

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. 

isotope

Elements having the same atomic number but different atomic weights; identical 

chemically but having different physical and nuclear properties. 

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen 

atom.

neutron film dosimeter

A film dosimeter that contains a Neutron Track Emulsion, type A, film packet.

operating area

Designation of major onsite operational work areas.

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.

radionuclide

A radioactive isotope of an element, distinguished by atomic number, atomic weight, 

and energy state. 

rem

A unit of dose equivalent equal to the product of the number of rad absorbed and the 

quality factor.

thermoluminescence

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

thermoluminescent dosimeter (TLD)

A holder containing solid chips of material that when heated will release stored energy 

as light. The measurement of this light provides a measurement of absorbed dose. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 84 of 95

ATTACHMENT A

SUMMARY OF ACTIVITIES AT SITE BUILDINGS

Page 1 of 3 

Building

Number

Building Use Radiological

Activities

Dates of 

Use

Nuclides Comments

904 Auditorium None 1980s–

present

905 Offices None 1980s–

present

906 Combustion Laboratory 

Research

None 1980s–

present

907 Equipment room None 1980s–

present

910 Weapons Laboratory 

Facility

Neutron 

Detection

Used to design, 

assemble, test, 

calibrate and repair 

neutron detectors

911 Administration None

912 Administration, 

computers 

None 1958–

present

913 Test assembly, machine

shop, plating, 

MetallographyTritium 

HVAHs

Tritium storage 

studies, U 

machining

1976–

1998

H-3, U-238

914 Test laboratories Radiflow, NDT, U 

machining, 

Kr-85, beta 

sources, 

U-238

No radiological work 

for ~15 yr

915 New building, office 

space

None

916 Mostly light chemistry 

laboratories, Previously 

a warehouse, Ar 

gloveboxes

e-microscopes, 

accelerators, 

various small 

sources, H-2 

storage tests

700 keV, 

1 MeV, 

U-238

Sealed sources

917 Cylinder storage

918 Raw stock U-machining U-238 Deconstructed in the 

late 1980s

920 Office space None 1977–

present

921 Decommissioned in 

1980

Radiological 

material storage 

and 

decontamination

of NTS test units

U-238, trace

Pu-239 and

MFPs

922 Office space None

923 Radiography Laboratory Radiography 

operations; X-ray

machines 

Co-60, 

Ir-192 X-ray 

machines. 

Cf-252, 

Converted to records 

storage in early 

1990s

924 Mo Mobile; Counting 

Laboratory

HP laboratory moved 

to Building. 973

925 Health services None
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ATTACHMENT A

SUMMARY OF ACTIVITIES AT SITE BUILDINGS

Page 2 of 3

Building

Number

Building Use Radiological

Activities

Dates of 

Use

Nuclides Comments

927 Radiological Material 

Storage

Warehouse and 

Vault

U-238, 

natural 

thorium

928 Shipping & receiving None

929 Office space None

940 Office space None

941 Light mechanical, 

electrical, chemical 

laboratories

Radiography X-ray U-238, 

beta sources

Sealed sources

942 Light mechanical, 

electrical, chemical 

laboratories

None

943 Plating operations

955 Environmental Test 

Facility

Env. Testing of 

various Test 

units

U-238 metal

956 Vibration Test Facility Environmental 

Testing of 

various Test units

U-238 metal

960 Offices

961 Radiological and Mixed 

Waste Storage and 

packaging

Storage & 

packaging

H-3, U-238

9611 Chemical waste storage None

963 Maintenance Facilities None

964 Security None

966 High Pressure Gas 

Dynamics Test Facility

None

967 Office None

968 TRL Tritium research 1976–

1989

H-3, U-238 

beds

Decommissioned in 

1998, converted to 

biotech

969 Former TRL waste 

storage 

Radiological 

waste storage for 

site

1976–

1989

H-3, U-238 Converted to shop 

and storage when 

TRL decommissioned

970 Welding shop

972 Centrifuge Test package 

centrifuge, 

Penetrator 

studies

U-238

973 ES&H laboratories

974 Explosives test tanks Neutron 

generator tests

H-3

976 Four test cells for High 

Press work

H-2 storage tests U-238

978 Flight test unit testing Mass properties 

tests

U-238
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 86 of 95

ATTACHMENT A

SUMMARY OF ACTIVITIES AT SITE BUILDINGS

Page 3 of 3

Building

Number

Building Use Radiological

Activities

Dates of 

Use

Nuclides Comments

979 H-2 storage, research 

and development

U-238 Three Ar gloveboxes,

two contaminated 

machines

983 Test cells; Flight Test 

Assembly

U-238

ESA Magazines; Explosives

Storage

H-3

Sources: DOE (1982); Wright (1981, 2006). 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 87 of 95

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

Page 1 of 9 

LIST OF FIGURES

FIGURE TITLE PAGE

B-1 Example external dosimetry history records (output from SANDOS) ................................ 88

B-2 Example nonroutine dosimeter evaluation form for special radiation work case 

(potential neutron exposures)........................................................................................... 89

B-3 Example nonroutine dosimeter evaluation form for special radiation work case 

(potential neutron exposures)........................................................................................... 90

B-4 Example unreturned dosimeter investigation report.......................................................... 91

B-5 Example unreturned dosimeter questionnaire follow-up letter........................................... 92

B-6 Example portion of unreturned dosimeter questionnaire showing various codes 

used by the Personnel Dosimetry Division........................................................................ 93

B-7 Work Area Radiation Dosimetry Request Form ................................................................ 94

B-8 Example work area radiation dosimetry results report. ..................................................... 95
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ATTACHMENT B 

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

Page 2 of 9 

Figure B-1. Example external dosimetry history records (output from SANDOS).
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ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES 

Page 3 of 9 

Figure B-2. Example nonroutine dosimeter evaluation form for special radiation work case 

(potential neutron exposures).
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ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES 

Page 4 of 9 

Figure B-3. Example nonroutine dosimeter evaluation form for special radiation work case 

(potential neutron exposures). 
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ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES 

Page 5 of 9 

Figure B-4. Example unreturned dosimeter investigation report. 
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ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

Page 6 of 9 

Figure B-5. Example unreturned dosimeter questionnaire follow-up letter. 
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ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

Page 7 of 9 

Figure B-6. Example portion of unreturned dosimeter questionnaire showing various codes 

used by the Personnel Dosimetry Division. 
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ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

Page 8 of 9 

Figure B-7. Work Area Radiation Dosimetry Request Form. 
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 95 of 95

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

Page 9 of 9 

Figure B-8. Example work area radiation dosimetry results report. 
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ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

Page 6 of 9

Figure B-5. Example unreturned dosimeter questionnaire follow-up letter.
Document No. ORAUT-TKBS-0053 Revision No. 00 Effective Date: 05/01/2007 Page 92 of 95

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

Page 6 of 9

Figure B-5. Example unreturned dosimeter questionnaire follow-up letter.