Page 1 of 107

DOE Review Release 02/17/2021

Summary Site Profile for Sandia National 

Laboratories in Livermore, California

ORAUT-TKBS-0053 Rev. 02

Effective Date: 01/13/2021

Supersedes: Revision 01

Subject Expert(s): Lawrence A. Page, Jr.

Document Owner 

Approval: Signature on File

Lawrence A. Page, Jr., Document Owner

Approval Date: 01/09/2021

Concurrence: Wade C. Morris Signature on File for

John M. Byrne, Objective 1 Manager

Concurrence Date: 01/11/2021

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 01/10/2021

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 01/12/2021

Approval: Signature on File

Timothy D. Taulbee, Associate Director for Science

Approval Date: 01/13/2021

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 2 of 107

 

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.

04/28/2014 01 Revision initiated to update the document with new information. 

Modified Sections 1.0, 5.0, and 6.0 to incorporate the Special 

Exposure Cohort class through 1994, including references to SECER and HHS designation letter. Replaced reference to ORAUTTKBS-0048, Site Profile for the Brookhaven National Laboratory, with 

ORAUT-TKBS-0036-6, Argonne National Laboratory-East – External 

Dosimetry. Added information pertaining to chronic lymphocytic 

leukemia to Section 3.3 and added Table 3-6. Amended Section 5.2 

to eliminate the sentence that only tritium and uranium were of 

concern, and stating there were limited bioassay data for tritium and 

uranium. Changed Section 5.7 to reflect that unmonitored internal 

dose is only applicable post-SEC, to indicate applicable years based 

on facility decommissioning, to remove all references to maximum 

permissible concentrations, and to supply uranium intake information 

for all solubilities. Altered Table 5-5 to remove the row related to 

1959 and to change the “1990s” row to “post-1994.” Adjusted Table

5-8 for consistency with Table 5-2, to account for tritium intake 

through skin absorption, and to address uranium tritides, adding 

ORAUT-OTIB-0066 as a reference. Included an instruction for dose 

reconstructors to use minimum detectable activities as provided in 

the records when available. Incorporates formal internal and NIOSH 

review comments. Training required: As determined by the 

Objective Manager. Initiated by Lawrence A. Page, Jr. 

01/13/2021 02 Revision initiated to update the document with new information 

related to a revision of ORAUT-OTIB-0006, Dose Reconstruction 

from Occupational Medical X-Ray Procedures, and the 

implementation of ORAUT-OTIB-0088, External Dose 

Reconstruction. Included information for skin doses to various areas 

from posterior-anterior and lateral chest X-rays in Table 3-5, and 

from lumbar spine examinations in Table 3-7. X-ray doses and 

computational information in Tables 3-1, 3-2, and 3-4, which had 

previously been labeled “Pre-1970” were renamed to include the year 

1970. Information and doses in the tables previously labeled “1970–

1985” were changed to reflect the period “1971–1975.” Skin dose 

information in Table 3-4 for 1986 to present were changed to reflect 

the revision of the backscatter factor from ORAUT-OTIB-0006. 
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 3 of 107

 

EFFECTIVE

DATE

REVISION

NUMBER DESCRIPTION

01/13/2021 02

(continued)

Revision updated to upgrade the document with new and corrected 

environmental data, both internal and external. Previously 

unavailable reference documents from Lawrence Livermore National 

Laboratory were employed to reevaluate existing environmental data 

and to provide additional data for the period 2004 to 2018. Updated 

uncertainty information regarding medical dose with information from 

the International Vocabulary of Metrology (JCGM 2008). Included 

information and corrections regarding NTA film badges. Incorporates 

formal internal and NIOSH review comments. Constitutes a total 

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

Objective Manager. Initiated by Lawrence A. Page, Jr.
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 4 of 107

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations..................................................................................................................8

1.0 Introduction ...............................................................................................................................11

1.1 Purpose.........................................................................................................................12

1.2 Scope............................................................................................................................12

1.3 Special Exposure Cohort ..............................................................................................12

2.0 Site Description.........................................................................................................................12

2.1 Introduction ...................................................................................................................12

2.1.1 Purpose ..........................................................................................................12

2.1.2 Scope .............................................................................................................12

2.1.3 Background ....................................................................................................13

2.2 Site Activities.................................................................................................................14

2.2.1 Tritium Research Laboratory..........................................................................18

2.2.2 Depleted Uranium Machining .........................................................................19

2.2.3 Radiography, Accelerators, and Neutron Generators ....................................19

2.3 Site Processes ..............................................................................................................20

2.3.1 Incidents .........................................................................................................20

2.3.2 Effluents .........................................................................................................21

2.4 Radiological Protection and Access Control .................................................................21

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

3.1 Examination Frequency ................................................................................................23

3.2 Equipment and Techniques ..........................................................................................23

3.3 Organ Dose Estimates ..................................................................................................24

3.3.1 Parameters and Estimation Method ...............................................................24

3.3.2 Organ Dose Estimates ...................................................................................27

3.4 Uncertainties .................................................................................................................27

4.0 Occupational Environmental Dose............................................................................................32

4.1 Introduction ...................................................................................................................32

4.1.1 Purpose ..........................................................................................................32

4.1.2 Scope .............................................................................................................33

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations .........................33

4.2.1 Radionuclides of Concern ..............................................................................33

4.2.2 Source Terms for Internal Dose .....................................................................35

4.2.3 Annual Intake of Radioactivity ........................................................................35

4.2.3.1 Inhalation Intakes ........................................................................35

4.3 External Exposure to Onsite Atmospheric Radionuclide Concentrations .....................38

4.3.1 Locations of Concern .....................................................................................38

4.3.2 Gamma-Emitting Radionuclides.....................................................................39

4.3.3 Neutron Exposure ..........................................................................................42

5.0 Internal Dosimetry.....................................................................................................................44

5.1 Introduction ...................................................................................................................44

5.1.1 Purpose ..........................................................................................................44

5.1.2 Scope .............................................................................................................44

5.1.3 Special Exposure Cohort................................................................................44

5.2 Summary of Potential Internal Radionuclide Exposures ...............................................45
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5.2.1 Tritium Research Laboratory..........................................................................45

5.2.2 Uranium Alloy Machining................................................................................45

5.3 Bioassay Programs .......................................................................................................46

5.3.1 Tritium Bioassay Programs ............................................................................46

5.3.2 Uranium Bioassay Programs..........................................................................47

5.4 Bioassay Laboratory Procedures and Minimum Detectable Intakes.............................47

5.4.1 Tritium ............................................................................................................47

5.4.2 Uranium..........................................................................................................49

5.5 Bioassay Results...........................................................................................................50

5.6 Method of Calculating Dose from Urine Bioassay Data ................................................50

5.7 Airborne Radionuclide Concentrations..........................................................................51

5.8 Unmonitored Dose Estimates .......................................................................................52

5.9 Radiation Dosimetry Reports ........................................................................................52

5.10 Summary Tables ...........................................................................................................53

6.0 Occupational External Dose .....................................................................................................54

6.1 Introduction ...................................................................................................................54

6.1.1 Purpose ..........................................................................................................54

6.1.2 Scope .............................................................................................................54

6.1.3 Special Exposure Cohort................................................................................54

6.2 Background...................................................................................................................54

6.3 Dose Reconstruction Parameters .................................................................................55

6.3.1 Site Historical Administrative Practices ..........................................................55

6.3.1.1 Administrative Practices Before 1989 .........................................55

6.3.1.2 Administrative Practices After 1989 ............................................56

6.3.2 Site Dosimetry Technology ............................................................................58

6.3.2.1 Dosimetry Technology, 1956 to 1988..........................................58

6.3.2.2 Dosimetry Technology, 1989 to Present .....................................59

6.3.3 Calibration ......................................................................................................60

6.3.3.1 Calibration for Dosimeter Technologies, 1965 to 1988 ...............60

6.3.3.2 Calibration for Dosimeter Technologies, 1989 to Present...........61

6.4 Dose Calculation and Reporting ...................................................................................61

6.4.1 Dose Calculation and Reporting, 1956 to 1988..............................................61

6.4.2 Dose Calculation and Reporting, 1989 to Present .........................................62

6.4.3 Exposure Energy Spectra ..............................................................................65

6.4.3.1 Neutron Dose Conversion Factors ..............................................67

6.4.4 Missed Dose...................................................................................................68

6.4.5 Organ Dose Conversion Factors....................................................................68

6.4.6 Adjustments to Recorded Neutron Dose........................................................68

6.4.6.1 Correction for Low-Energy Neutrons with NTA Neutron 

Dosimeters ..................................................................................69

6.4.6.2 Angular Dependence of NTA Neutron Dosimeters .....................69

6.4.6.3 Track Fading of Neutron NTA Dosimeters ..................................69

6.4.7 Uncertainty .....................................................................................................69

7.0 Attributions and Annotations .....................................................................................................70

References ...........................................................................................................................................71

Glossary................................................................................................................................................93

ATTACHMENT A SUMMARY OF ACTIVITIES AT SITE BUILDINGS...........................................96
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ATTACHMENT B DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION 

EXAMPLES .......................................................................................................99

LIST OF TABLES

TABLE TITLE PAGE

2-1 Area information and parameters .............................................................................................16

2-2 Historical T2 inventory at TRL ...................................................................................................18

2-3 Effluents of 3H to air and sewer during operations and cleanup of TRL ...................................21

3-1 Entrance kerma by procedure and period.................................................................................25

3-2 DCFs; absorbed dose for organs at various Al HVL for radiography........................................26

3-3 Organs without ICRP Publication 34 DCFs and their substitutes .............................................27

3-4 Organ dose estimates for chest X-rays.....................................................................................28

3-5 Skin area dose estimates from chest X-rays ............................................................................28

3-6 Organ doses from lumbar spine X-ray, 1956 to 1971...............................................................30

3-7 Skin doses from lumbar spine X-ray, 1956 to 1971 ..................................................................30

3-8 Distributions and corresponding statistical parameters for the dose to the B-lymphocytes......31

4-1 Summary of methodology for estimating intakes of 3H, 234U, and 239Pu ....................................35

4-2 Sitewide annual median inhalation intakes, 1956 to 1970 ........................................................36

4-3 Sitewide annual median inhalation intakes, 1971 to 2018 ........................................................36

4-4 External gamma radiation dose based on 2,500 hr/yr exposure duration ................................41

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

exposure duration .....................................................................................................................43

5-1 MDAs for and frequency of tritium bioassay .............................................................................48

5-2 Preliminary timeline for uranium in urine analyses ...................................................................50

5-3 Range of tritium doses based on bioassay, 1979 to 1981........................................................50

5-4 Estimated tritium doses based on bioassay, 1982 to 1995.......................................................51

5-5 DAC values for tritium and uranium..........................................................................................52

5-6 Internal dose control program...................................................................................................53

5-7 Detection limits for urine bioassay for 3H with LSC ...................................................................53

5-8 Detection limits for total uranium bioassay ...............................................................................53

5-9 Facility, source, solubility type, and particle size ......................................................................54

6-1 Documented dosimetry technologies and dose quantities........................................................58

6-2 Monthly and quarterly LLDs for the Harshaw Model 8802 dosimeter card ...............................65

6-3 Selection of radiation energies and percentages......................................................................66

6-4 Neutron DCFs for applicable SNL-CA facilities.........................................................................67

6-5 Estimated maximum annual missed photon, beta, and neutron dose ......................................68

6-6 Dose units for organ DCFs .......................................................................................................68

6-7 Systematic uncertainty estimates .............................................................................................69

A-1 Summary of activities by building..............................................................................................96
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LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Boundaries of SNL-CA property ...............................................................................................14

2-2 Map of SNL-CA facilities...........................................................................................................15

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

tube potential ............................................................................................................................25

4-1 TLD locations since 1989..........................................................................................................39

4-2 Location of LLNL perimeter gamma and neutron dosimeters...................................................40

4-3 Location of LLNL offsite gamma dosimeters.............................................................................40

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

average measurements at LLNL site perimeter and LLNL Site 300 .........................................42

6-1 ALARA reporting action levels ..................................................................................................56

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

radiation and dose equivalent quantities...................................................................................60

6-3 Example Harshaw 8800 card reader calibration procedures, forms, analysis, and output .......62

6-4 Example calibration form for the Harshaw 4000 dosimetry system ..........................................63

6-5 Angular dependence testing results for the Harshaw 8800/8812 dosimetry system. ...............64

B-1 Example external dosimetry history records (output from SANDOS) .....................................100

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

neutron exposures).................................................................................................................101

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

neutron exposures).................................................................................................................102

B-4 Example unreturned dosimeter investigation report ...............................................................103

B-5 Example unreturned dosimeter questionnaire follow-up letter................................................104

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

the Personnel Dosimetry Division...........................................................................................105

B-7 Work Area Radiation Dosimetry Request Form......................................................................106

B-8 Example work area radiation dosimetry results report............................................................107
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 8 of 107

 

ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

ALARA as low as is reasonably achievable

AP anterior-posterior

AWE Atomic Weapons Employer

Bq becquerel

CEDE committed effective dose equivalent

CEP Controls for Environmental Pollution

CFR Code of Federal Regulations

cGy centigray

Ci curie 

cm centimeter

cpm counts per minute

CRDL Chemical and Radiation Detection Laboratory

d day

DAC derived air concentration

DCF dose conversion factor

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ESE entrance skin exposure

ft feet

g gram

GPS Gas Purification System

GSD geometric standard deviation

Gy gray

Hp(10) personal dose equivalent at 10 millimeters depth in tissue

HEPA high-efficiency particulate air

HEU highly enriched uranium

hr hour

HTO tritiated water vapor

HVL half-value layer

ICP-MS inductively coupled plasma–mass spectrometry

ICRP International Commission on Radiological Protection

ICT Insulating Core Transformer

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kVp peak kilovoltage

L liter

LAT lateral
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 9 of 107

LLD lower limit of detection

LLNL Lawrence Livermore National Laboratory

LSC liquid scintillation counting or counter

m meter

mA beam current

MBA Mass Balance Area

mCi millicurie

MDA minimum detectable amount

MDC minimum detectable concentration

MDL minimum detection limit

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mGy milligray

mL milliliter

mm millimeter

mm Al millimeters of aluminum

mR milliroentgen

mrad millirad

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

ORAU Oak Ridge Associated Universities

PA posterior-anterior

pCi picocurie

PER program evaluation report

RDC Radiation Detection Company

RESL Radiological and Environmental Sciences Laboratory

s second

SEC Special Exposure Cohort

SID source-to-image distance

SNL Sandia National Laboratories

SNL-CA SNL Livermore, California, facilities

SNL-NM SNL Albuquerque, New Mexico, facilities

SOP Safe Operating Procedure

SRDB Ref ID Site Research Database Reference Identification (number)

SSD source-to-skin distance

SWP Safe Work Permit

T2 elemental tritium

TBD technical basis document

TLD thermoluminescent dosimeter

TMA/EAL Thermo-Analytical Incorporated/EAL Corporation

TRL Tritium Research Laboratory

U.S.C. United States Code

VERS Vacuum Effluent Recovery System
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 10 of 107

wk week

yr year

μCi microcurie

μg microgram

μR microroentgen

§ section or sections
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 11 of 107

1.0 INTRODUCTION

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

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

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

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

facilities or categories of DOE or AWE facilities. They will be revised in the event additional relevant 

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean nor should it be equated to an 

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined 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 [DOE] (except for buildings, 

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

of Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

and occupational radiation exposure at a DOE or AWE facility during the facility’s designated time 

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and the types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

NIOSH does not consider the following types of exposure as those incurred in the performance of 

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (NIOSH 2010):

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

conventional structures

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

reasons
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 12 of 107

 

1.1 PURPOSE

This site profile for the Sandia National Laboratories (SNL) site in Livermore, California (SNL-CA), 

describes aspects of the SNL-CA site and historical activities and practices pertinent to dose 

reconstruction under EEOICPA.

1.2 SCOPE 

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 from 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 in relation to 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. 

1.3 SPECIAL EXPOSURE COHORT

NIOSH has determined that it is not feasible to reconstruct certain components of internal and 

external dose from October 1, 1957, through December 31, 1994, due to a lack of sufficient 

information, which includes biological and workplace monitoring data and radiological source 

information (Sebelius 2013):

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

and subcontractors who worked in any area at Sandia National Laboratories-Livermore 

in Livermore, California, from October 1, 1957 through December 31, 1994, for a number 

of work days aggregating at least 250 work days, occurring either solely under this 

employment, or in combination with work days within the parameters established for one 

or more other classes of employees in the Special Exposure Cohort.

NIOSH lacks sufficient information, which includes internal and external personnel monitoring data,

process data, and radiological source term information, to allow it to estimate with sufficient accuracy 

the potential internal and external exposures to radionuclides that include but are not limited to 

uranium, uranium tritides and hydrides, tritium, and thorium, as well as potential exposures from 

classified radiological activities to which the class may have been subjected. NIOSH finds that it is 

likely feasible to reconstruct occupational medical dose for SNL-CA workers with sufficient accuracy 

through 1989 (NIOSH 2013). Since 1989, medical X-rays have been performed at Lawrence 

Livermore National Laboratory (LLNL) (ORAUT 2006a). 

2.0 SITE DESCRIPTION

2.1 INTRODUCTION

2.1.1 Purpose

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.

2.1.2 Scope

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.
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2.1.3 Background

The 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 LLNL nuclear weapons designs. The primary 

mission during the Cold War (1956 to 1989) was the design and testing of nonnuclear components of 

nuclear weapons designed by LLNL. SNL-CA was to engineer, or “weaponize,” the nuclear physics 

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

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). Sandia Corporation, a Lockheed Martin Company, operated SNL-CA and 

Sandia National Laboratories-New Mexico (SNL-NM) in Albuquerque. In May 2017, the contract was 

awarded to National Technology and Engineering Solutions of Sandia, a wholly owned subsidiary of 

Honeywell International.

The SNL-CA site presently consists of approximately 70 buildings and other facilities on 410 acres 

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 first completed in September 1957; the rest of the original 

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

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

of engineers and support staff. At first, 14 SNL-NM employees worked with LLNL in LLNL facilities. 

In 1957, SNL-CA began using 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 of LLNL 

activities expanded to include effects test analyses and telemetry for 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 

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

Although SNL-CA retains its 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. The initial move into research on 3H grew out of its 

familiarity with, and use of, 3H in components. In 1974, the first structure dedicated to 3H 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 at 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 were added as a buffer zone on the east side near the TRL in 1979. In 1986, an 

additional 228 acres were 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 
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 14 of 107

 

Figure 2-1. Boundaries of SNL-CA property (Larsen 2005). 

site’s area down to 410 acres where it remains today. In 2004, there were 1,094 employees at the 

site (Larsen 2005).

2.2 SITE ACTIVITIES

In support of the various missions identified in Section 2.1.3, 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 is 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.

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

handled kilogram amounts of depleted uranium (DU), gram amounts of 3H (when the TRL was 

operational), and only microcurie quantities of other isotopes (SNL 1983). However, the 1992 

Environmental Impact Statement for LLNL and SNL-CA noted 100 Ci 192Ir and 60Co radiographic 
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Figure 2-2. Map of SNL-CA facilities (Ullrich 2003). 

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

in a shielded radiography cell (DOE 1992, Volume II).
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Table 2-1. Area information and parameters.a

Area Weapons Laboratory Facility Complex

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

Period 1958–1998

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

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

analysis of materials, Radiflo leak tests

Radiation 

sources

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

Kr-85

Area Radiography

Description Building 923

Period Unknown–early 1990s

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

Radiation 

sources

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

Area Micro and Nano Technologies Laboratories

Description Buildings 941, 942, and 943

Period Unknown–present

Activities Radiography for materials science studies

Radiation 

sources

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

Area

Former Tritium Research Laboratories

(currently the Chemical and Radiological Detection Laboratory)

Description Buildings 967, 968, and 969

Period 1974–1996

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

Radionuclides H-3, 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

Period 1958–present

Activities Environmental testing of mock-up weapons and components

Radionuclides DU

Area Storage Facilities

Description Buildings 921 (became Building 9632 in 1980), 927, 961, 982

Period Unknown–present

Activities Storage and packaging of waste materials

Radionuclides H-3, DU, natural thorium, trace Pu-239 and mixed fission products

a. DU = depleted uranium.

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 1989; Wallace 1988). 

Powdered DU sealed in storage containers for 3H-storage studies (as the tritide) has also been 

present in approximately 1-kg amounts (SNL 1991a). A classified activity involving powdered uranium 

hydride in gram amounts was conducted in gloveboxes in Buildings 979, 916, and the TRL (ORAUT 

2006a). There were some thorium metal parts used on test systems as well (ORAUT 2006a). 

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 that were shipped off site for incineration in Florida (DOE 

1992, Volume II). Details about total quantities of 3H present for each year of operation of the TRL 
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and pertinent to types of research activities carried out in this facility are described in Section 2.2.1

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 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 had no significant environmental 

impact and they involved radionuclides in small enough quantities that special containment features 

and operating procedures were not required to ensure that no radioactivity was released. The only 

exceptions to this finding of no potential significant impact were the SNL-CA radioactivity storage 

vaults, one of which was in Building 927, another small vault within the TRL (Building 968), and the 

TRL itself. The contents of the 927 vault varied according to work needs, but generally included DUcontaining mock-ups of weapons and small, sealed radioactive sources not in use. Tritium was not 

stored in the 927 vault; rather the small quantities 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 3H sources from the 

vaults in Building 968 (the TRL), experiments were conducted with up to 120 g (1.2 × 106 Ci) of 3H 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, 

Volume II) 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); 

therefore, the radiological activities at that time involved radioactive material management of legacy

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 (Wright 1981a):

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.

This information suggests that activities involving DU and 3H 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 this section discusses 

aspects of the SNL-CA site pertinent to these sources of exposure.
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2.2.1 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 3H 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 is 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).

Table 2-2. Historical T2 inventory at TRL.

Year

Mass

(g)

Activity

(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

The TRL was divided into nine laboratories, two office areas, a 3H storage room, a control room, a 

shop, a 3H gas purification equipment room, and eight 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 on the types of research but were focused on the physical and 

chemical characterization of 3H 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 

of 3H (DOE 1992, Volume II).

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 
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have been in place in 1991 (SNL 1991b). Radiological materials were also dispensed through 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 responsible for 

issuing nuclear materials to personnel in that MBA. Material was accounted for to the nearest 0.01 g, 

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 to the stack.

2.2.2 Depleted Uranium Machining

Machine shops in Buildings 913, 914, and in an annex of Building 918 were 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 in which dry cutting 

occurred (ORAUT 2006a). The 1989 Safe Operating Procedure (SOP) for machining of uranium 

metal (SNL 1989) dictated that all machining operations were to be performed wet and that dry 

operations were prohibited. According to a 1985 DOE Headquarters appraisal of SNL-CA and 

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

quarter. The appraisal and response report also indicates that the Hazards Control Division at SNLCA, as of 1987, had 16 years of air sampling data for this procedure, which “demonstrate that 

airborne contamination is not a problem during typical machining operations.”

2.2.3 Radiography, Accelerators, and Neutron Generators

Until the early 1990s, the Radiography Facility was a 3,880-ft2 building (Building 923) that included 

X-ray machines, a gamma-ray source, neutron sources, and many alpha and beta sources (DOE 

1992, Volume II). The radiation sources were primarily used for the radioscopy and electron imaging 

of weapon and nonweapon components. The radiation sources included the radiation-producing 

machines with energies up to 420,000 volts, the 100 Ci 192Ir and 60Co radiographic isotopes, the 

sealed 252Cf spontaneous fission neutron source inside a massive neutron source shield, and many 

other smaller sealed neutron, alpha, beta, and gamma-ray isotope sources of low activity (1 μCi to 

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

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

radiation transmission gauging, backscatter measurements, X-ray fluorescence studies, and neutron 

activation analysis (DOE 1992, Volume II).

X-ray diffraction equipment is reported as being present in Building 913, Room 115, as early as 1965 

(SNL 1964). An incident that occurred with a diffractometer (see Section 2.3.1) 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 in Building 916 (DOE 1992, Volume II; DOE 2003). 

These include a 1-MeV Tandem accelerator and a 700-/200-keV positive ion accelerator (Morse 

1983; SNL 1991c; ORAUT 2006a). 
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There are neutron generators used in the design, assembly, testing, calibration, and repair of neutron 

detectors in Building 910 (SNL 1990). Neutron generator tests are also conducted in Building 974 

(Attachment A).

2.3 SITE PROCESSES

2.3.1 Incidents

According to Garcia and Gorman (1996), a few incidents occurred in relation to the TRL that could 

have led to 3H exposures in excess of routine operations. Other sources (Lovell 1980; SNL 1960–

1983) listed excess exposure to ionizing radiation, but only one was attributable to a work-related 

exposure. These incidents are listed below.

Personnel Exposure to Ionizing Radiation, December 1979

[Redacted]. 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).

A similar event was reported through worker interviews to have occurred in 1978 involving the same 

equipment operated under similar circumstances. NIOSH has determined that this event was of less 

severity than that in 1979 (NIOSH 2007a). The dose evaluation described for the 1979 event as 

described above should be used for the earlier event (Lovell 1980).

Personnel Exposure and Tritium Release, July 1984

During disassembly of an engineering experiment in [redacted], an environmental release of 

approximately 2.5 Ci of [redacted] occurred. [Redacted] 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.

Personnel Exposure and Tritium Release, August 1987

An 1,100-Ci environmental release occurred when an [redacted] disassembled a vessel [redacted] in

[redacted]. This release of T2 caused a building evacuation, and an [redacted] 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 about 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

[Redacted] personnel received doses from 2 to 5 mrem when a [redacted] failed and caused a tritium 

release in [redacted]. An environmental release of 2 Ci HTO was reported by Garcia and Gorman 

(1996), but it was also reported as a 30-Ci stack release by Garcia (1994a). The personnel doses 

came from both cleanup and [redacted] replacement operations.
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2.3.2 Effluents 

The only radionuclides released due to normal operations from SNL-CA have been small amounts of 

tritium and trace amounts of DU (SNL 1982, 1983, 1984; Devlin 1986 to 1988; Siegfriedt 1989; Brekke 

1990, 1991; Brekke and Holland 1992 to 1995; Holland and Brekke 1996; Holland 1997 to 2002; 

Larsen 2003 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 [SNL 1982, 1983, 1984; Devlin 1986]). In 1990, DOE (1990) 

reported that emissions of radioactive particulates from SNL-CA activities that potentially generate 

such particulates were controlled with high-efficiency particulate air (HEPA) filters. A total of 33 HEPA 

filters 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 has often been declared to be the only 

detectable effluent (Devlin 1986 to 1988; Siegfriedt 1989; Brekke 1990, 1991; Brekke and Holland 

1992 to 1995; Holland and Brekke 1996; Holland 1997). Monitoring of tritium effluents from the TRL 

ceased in 1996 after the facility was transitioned for other uses (Holland 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 2.3.1

above.

Table 2-3. Effluents of 3H 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

2.4 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), U.S. 

Energy Research and Development Administration, and DOE criteria (Kingsley 1968; Nester 1994; 

SNL 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 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). A 1994 memorandum 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 (Perez-Romo 1994). It does not appear that the records from before 

1989 for SNL-CA were incorporated into the current database at SNL-NM for SNL employees. 

Dosimetry data for the period from 1973 to 1987 was recently recovered from SNL-NM, and data for 

1988 was recovered from SNL-CA. Cumulative summary data for the period before 1973 has also 

been recovered from SNL-CA. However, annual data for all years before 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 

1958–1978). In 1958, the radiation exposure summary report indicates that 41 employees (of 537) 

were not monitored. However, a 1960 internal memorandum indicates that all employees were 

monitored from the time of establishment of the Sandia Livermore Corporation (SNL 1958–1961). It 

was noted in 1984 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) (Lovell 1984a,b). Lovell (1984a,b) stated 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 the AEC 

Manual Chapter 0524 (AEC 1963) criteria of 6 mR/hr for radiation exposure to the whole body or 

critical organs for wearing dosimeters.

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 1991c; Morse 1983). 

Access was also restricted to the machine shop (Room 119) in Building 913 during uranium 

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

(SNL 1989). 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 2.2.2.

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 Guideline (NIOSH 

2007b) and 42 CFR Part 82 require the X-ray doses to be part of the occupational radiation exposure. 

This section 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.
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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 found. A protocol for the frequency of chest X-ray examinations as a 

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

claimant files available at this time generally indicate that a single posterior-anterior (PA) chest X-ray 

examination was performed at hire, annually, and possibly at termination from 1956 through the 

1980s. Those records also show 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) indicate that LAT chest X-rays were 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 (ORAUT 

2006a). The clinic at LLNL has provided X-ray services for SNL-CA since 1990 (ORAUT 2006a). No 

evidence of the use of photofluorography has been found. It is recommended that biennial chest 

X-rays be assumed for all employees after 1989 (ORAUT 2010a).

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), peak 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 preemployment set of X-rays may have been 

performed at SNL-NM or off the site in California. 

This section provides organ dose estimates from occupational X-ray chest examinations administered 

at SNL-CA from 1956 through 1970, 1971 to 1985, and 1986 to present using calculated site-specific 

estimates with input from ORAUT-OTIB-0006, Dose Reconstruction from Occupational Medical X-Ray 

Procedures (ORAUT 2019a). For the years through 1970, the default values from ORAUT-OTIB0006 assume minimal beam collimation and a half-value layer (HVL) of 2.5 mm Al. For 1971 to 1985, 

the default values assume that the beams were collimated and the HVL was 2.5 mm Al. For 1986 

and after, 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.

The period for lumbar spine examinations is 1956 through 1971. 

For all periods the analysis should assume, as a default, 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 preemployment X-ray record, the 
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dose reconstructor should assign dose from a PA chest and AP and LAT lumbar spine as the default 

for the preemployment X-rays.

Efforts will continue to find related SNL-CA X-ray protocol information. Until that is found, 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

International Commission on Radiological Protection (ICRP) Publication 34 guidance uses the 

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

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

• Total filtration (millimeters of aluminum, mm Al), 

• Estimate of tissue thickness (AP and LAT), and

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

If measured air kerma are available, they should be used. For SNL-CA, air kerma was estimated from 

Figure 3-1 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 1956 to 1970, 1971 to 

1985, and 1986 to 1989, after which time X-rays for SNL-CA were performed at LLNL. For the first 

two periods, 80 kVp, 5 mAs, a total filtration of 2.5 mm Al, and a source-to-skin distance (SSD) of 154 

cm results in an air kerma estimate of 0.06 mGy per mAs at 1 m. Total filtration of 2.5 mm Al is a 

reasonable estimate for those periods (ORAUT 2019a). 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 2019a). 

For chest X-rays in 1986 and after, 110 kVp, 300 mA and 1/30-s exposure time, and SID of 183 cm 

(from claim files), with a total filtration estimate of 2.5 mm Al and HVL of 4.0 (ORAUT 2019a) result in 

an estimated air kerma of 0.126 mGy/mAs and calculated entrance kerma of 0.53 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 2019a). 

For lumbar spine X-rays (1956 to 1971), 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 lumbar spine X-rays 

based on information in claim files. A total filtration of 2.5 mm Al and HVL of 2.5 mm Al were 

assumed (ORAUT 2019a), resulting in an estimated air kerma of 0.05 mGy/mAs for AP lumbar spine 

X-rays and 0.070 mGy/mAs for LAT lumbar spine X-rays. SSDs of 63 cm for AP and 52 cm for LAT 

X-rays were used, resulting in entrance kerma of 9.45 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.
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 25 of 107

 

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).

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)

1956–1970 chest 0.013a 0.033 Not applicable Not applicable

1956–1971 lumbar spine Not applicable Not applicable 0.945b 3.88c

1971–1985a chest 0.013 0.033 Not applicable Not applicable

1986–1989d,e chest 0.053 0.133 Not applicable Not applicable

a. Based on 80 kVp and 5 mAs, as observed in SNL-CA claim files and 2.5 mm Al total filtration (ORAUT 2019a). LAT 

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

(100/154 cm)2 = 0.13 mGy.

b. Assumes 75 kVp and 75 mAs, as observed in SNL-CA claim files, and total filtration of 2.5 mm Al (ORAUT 2019a). 

Entrance kerma = (0.05 mGy/mAs) × (75 mAs) × (100/63 cm)2 = 9.45 mGy.

c. Assumes 85 kVp and 150 mAs, as observed in SNL-CA claim files, and total filtration of 2.5 mm Al (ORAUT 2019a). 

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

d. Based on 110 kVp, 300 mA and 1/30-s exposure time (noted in one claim file), and total filtration of 2.5 mm Al (ORAUT 

2019a). LAT chest entrance kerma is 2.5 times PA chest entrance kerma. PA entrance kerma = (0.126 mGy/mAs) × 

(10 mAs) × (100/154 cm)2 = 0.53 mGy. 

e. Medical procedures were performed at LLNL after 1989 (ORAUT 2006a). Information post-1989 is found in ORAUT 

(2010a). 
 

Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 26 of 107

Table 3-2. DCFs (mGy per mGy air kerma); absorbed dose (1 mGy) for organs at various Al HVL for radiography (ORAUT 2019a; ICRP 

1982), Appendix 1, Tables A.2–A.8).a,b,c

Organ

Lumbar spine

DCFs

1956–1971

(2.5-mm Al 

HVL)

LAT

Lumbar spine

DCFs

1956–1971

(2.5-mm Al 

HVL)

AP

Chest

DCFs

1956–1970

(2.5-mm Al 

HVL)

LAT

Chest

DCFs

1956–1970

(2.5-mm Al 

HVL)

PA

Chest

DCFs

1971–1985

(2.5-mm Al 

HVL)

LAT

Chest

DCFs

1971–1985

(2.5-mm Al 

HVL)

PA

Chest

DCFs

1986–1989

(4.0-mm Al 

HVL)

LAT

Chest

PA DCFs

1986–1989

(4.0-mm Al 

HVL)

PA

Thyroid 1.00E−05 3.00E−04 1.37E−01 1.74E−01d 1.15E−01 3.20E−02 1.64E−01 7.80E−02

Eye/brain 1.00E−05 3.00E−04 1.37E−01 3.20E−02 1.15E−01 3.20E−02 1.64E−01 7.80E−02

Ovaries Measurede Measurede Measurede Measurede 6.00E−04 1.00E−03 2.50E−03 5.20E−03

Liver/

gall bladder/

spleen

Measurede Measurede 2.20E−01 4.51E−01 2.20E−01 4.51E−01 3.51E−01 6.74E−01

Urinary 

bladder

Measurede Measurede Measurede N/A 6.00E−04 1.00E−03 2.50E−03 5.20E−03

Colon/rectum Measurede Measurede Measurede N/A 6.00E−04 1.00E−03 2.50E−03 5.20E−03

Testes Measurede Measurede Measurede Measurede 1.00E−04 1.00E−05 1.00E−04 1.00E−05

Lungs (male) 1.40E−02 7.90E−02 1.93E−01 4.19E−01 1.93E−01 4.19E−01 3.13E−01 6.28E−01

Lungs (female) 1.40E−02 7.90E−02 2.20E−01 4.51E−01 2.20E−01 4.51E−01 3.51E−01 6.74E−01

Thymusf 1.40E−02 7.90E−02 2.20E−01 4.51E−01 2.20E−01 4.51E−01 3.51E−01 6.74E−01

Esophagusf 1.40E−02 7.90E−02 2.20E−01 4.51E−01 2.20E−01 4.51E−01 3.51E−01 6.74E−01

Stomachf Measurede Measurede 2.20E−01 4.51E−01 2.20E−01 4.51E−01 3.51E−01 6.74E−01

Bone surfacesf Measurede Measurede 2.20E−01 4.51E−01 2.20E−01 4.51E−01 3.51E−01 6.74E−01

Remainderf Measurede Measurede 2.20E−01 4.51E−01 2.20E−01 4.51E−01 3.51E−01 6.74E−01

Female breast 1.30E−02g 2.50E−02g 2.55E−01 4.90E−02 2.55E−01 4.90E−02 3.43E−01 1.16E−01

Uterus 3.10E−02 2.87E−01 Measurede Measurede 6.00E−04 1.30E−03 2.10E−03 5.20E−03

Bone marrow 

(male)

2.20E−02 3.70E−02 3.70E−02 9.20E−02 3.70E−02 9.20E−02 7.60E−02 1.78E−01

Bone marrow 

(female)

2.20E−02 3.70E−02 2.90E−02 8.60E−02 2.90E−02 8.60E−02 5.90E−02 1.72E−01

Skinh 1.355E+00 1.355E+00 1.35E+00 1.355E+00 1.35E+00 1.355E+00 1.42E+00 1.42E+00

a. DCFs for lumbar spine are from ICRP (1982, Appendix 1, Tables A.2–A.8) and DCFs for chest are from ORAUT (2019a). 

b. N/A = not applicable.

c. Medical procedures were performed at LLNL after 1989 (ORAUT 2006a). Information post-1989 is found in ORAUT (2010a). 

d. Per ORAUT (2019a), DCF for AP cervical spine corrected for depth by 0.2.

e. Organ dose values for the testes and ovaries for lumbar spine reflect actual measurements reported in Lincoln and Gupton (1958).

f. Using analogues listed in Table 3-4.

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

h. Backscatter factor from NCRP (1997, Table B-8).
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 27 of 107

 

The ICRP tables used to estimate absorbed dose (ICRP 1982) do not include all the organs included 

in the Interactive RadioEpidemiological Program (IREP). 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 dose coefficient for ovaries is used. This approach should be either favorable to 

the claimant or neutral. Table 3-3 lists analogues for IREP organs from ORAUT (2019a). 

Table 3-3. Organs without ICRP Publication 34 DCFs and their substitutes.a

External organ

ICRP 34

substitute DCFb

Thymus, esophagus Lung

Stomach, bone surface, liver/gall bladder/spleen, small intestine, remainder Lungc or ovary

Urinary bladder, colon/rectum Ovary

Eye/brain Thyroid

a. Applies to most properly collimated radiographic procedures.

b. Source: ORAUT (2019a, Table 5-1). 

c. The ovary is used as the substitute DCF for these organs for the lumbar spine.

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 (through 1970, 1971 to 1985, and 1986 to 1989). LAT chest X-rays were, for the most part, 

discontinued after 1970; however records for LAT chest X-rays have been discovered for some claims 

after that time, and should be included. 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 (2019a). Chest X-rays after 1989 were performed at LLNL (ORAUT 2006a).

Table 3-5 presents skin doses from PA and LAT chest X-rays that should be used for each chest 

X-ray period. These X-rays were often taken at hire.

Table 3-6 presents organ doses 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. Table 3-7 provides skin doses for lumbar spine 

X-rays during this period. AP skin doses to the hands and forearms are reduced for 1971, based on 

guidance in ORAUT (2019a). 

The B-lymphocyte cells are the tissue at risk for chronic lymphocytic leukemia. The dose equivalent 

to the B-lymphocytes was determined using the method in ORAUT-OTIB-0082, Dose Reconstruction 

Method for Chronic Lymphocytic Leukemia (ORAUT 2012a), site-specific information, and ICRP 

Publication 34 DCFs (ICRP 1982). The dose distributions and corresponding statistical parameters 

for the dose to the B-lymphocytes for each projection and period are listed in Table 3-8.

3.4 UNCERTAINTIES

As stated in the International Vocabulary of Metrology (JCGM 2008), “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 implies knowledge of what the correct or actual value is, which is of course not known. 

Therefore, the more appropriate factor is uncertainty, which is expressed in terms of a coverage 

probability (e.g., a 99% coverage probability indicates that the correct or true value, although not 

actually known, has a 99% probability of falling within the cited range). Uncertainty includes both 

precision and adjustments for bias. 
 

Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 28 of 107

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

Organ

1956–1970

estimated 

dosea,b,c

HVL = 2.5 mm Al 

(uncollimated) 

LAT

1956–1970

estimated 

dosea,b,c

HVL = 2.5 mm Al 

(uncollimated) 

PA

1971–1985

estimated 

dosea,b,c 

HVL = 2.5 mm Al 

(collimated)

LAT

1971–1985

estimated 

dosea,b,c 

HVL = 2.5 mm Al 

(collimated)

PA

1986–1989 

estimated dosea,b,f

HVL = 4.0 mm Al 

(collimated)

LAT

1986–1989

estimated dosea,b,f

HVL = 4.0 mm Al 

(collimated)

PA

Thyroid 4.52E−03 2.26E−03 3.80E−03 4.16E−04 2.18E−02 3.35E−03

Eye/brain 4.52E−03 4.16E−04 3.80E−03 4.16E−04 2.18E−02 3.35E−03

Ovaries 1.30E−02 2.50E−02 1.98E−05 1.30E−05 3.33E−04 2.24E−04

Liver/gall bladder/ 

spleen

7.26E−03 5.86E−03 7.26E−03 5.86E−03 4.67E−02 2.90E−02

Urinary bladder 1.30E−02 2.50E−02 1.98E−05 1.30E−05 3.33E−04 2.24E−04

Colon/rectum 1.30E−02 2.50E−02 1.98E−05 1.30E−05 3.33E−04 2.24E−04

Testes 1.30E−02 5.00E−03 3.30E−06 1.30E−07 1.33E−05 4.30E−07

Lungs 7.26E−03 5.86E−03 7.26E−03 5.86E−03 4.67E−02 2.90E−02

Thymus 7.26E−03 5.86E−03 7.26E−03 5.86E−03 4.67E−02 2.90E−02

Esophagus 7.26E−03 5.86E−03 7.26E−03 5.86E−03 4.67E−02 2.90E−02

Stomach 7.26E−03 5.86E−03 7.26E−03 5.86E−03 4.67E−02 2.90E−02

Bone surfaces 7.26E−03 5.86E−03 7.26E−03 5.86E−03 4.67E−02 2.90E−02

Remainder 7.26E−03 5.86E−03 7.26E−03 5.86E−03 4.67E−02 2.90E−02

Female breast 8.42E−03 6.37E−04 8.42E−03 6.37E−04 4.56E−02 4.99E−03

Uterus 1.30E−02 2.50E−02 1.98E−05 1.69E−05 2.79E−04 2.24E−04

Bone marrow 1.22E−03 1.20E−03 1.22E−03 1.20E−03 1.01E−02 7.65E−03

Skind,e 4.46E−02 1.76E−02 4.46E−02 1.76E−02 1.89E−01 6.11E−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 (2019a). 

d. Calculated using backscatter factor of 1.35 from NCRP (1997, Table B-3) for skin dose estimates through 1985; consistent with ORAUT (2019a). 

e. Calculated using backscatter factor of 1.42 from NCRP (1997, Table B-3) for skin dose estimates after 1985; consistent with ORAUT (2019a). 

f. Procedures after 1989 were performed at LLNL (ORAUT 2006a); doses are as found in ORAUT (2010a).

Table 3-5. Skin area dose estimates from chest X-rays (rem).a 

Skin area

1956–1970

LAT

1956–1970

PA

1971–1985

LAT

1971–1985

PA

1986–1989b

LAT

1986–1989b

PA

R front shoulder 4.46E−02 3.84E−04 4.46E−02 3.84E−04 1.89E−01 2.08E−03

R back shoulder 4.46E−02 1.76E−02 4.46E−02 1.76E−02 1.89E−01 6.11E−02

L front shoulder 1.96E−04 3.84E−04 1.96E−04 3.84E−04 1.41E−03 2.08E−03

L back shoulder 1.96E−04 1.76E−02 1.96E−04 1.76E−02 1.41E−03 6.11E−02

R upper arm to elbow 4.46E−02 1.76E−02 4.46E−02 1.76E−03 1.89E−01 6.11E−03

L upper arm to elbow 4.46E−02 1.76E−02 1.96E−04 1.76E−03 1.41E−03 6.11E−03

L hand 4.46E−03 1.76E−02 4.46E−03 1.76E−03 1.89E−02 6.11E−03
 

Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 29 of 107

Skin area

1956–1970

LAT

1956–1970

PA

1971–1985

LAT

1971–1985

PA

1986–1989b

LAT

1986–1989b

PA

R hand 4.46E−03 1.76E−02 4.46E−03 1.76E−03 1.89E−02 6.11E−03

L elbow, forearm, wrist 4.46E−03 1.76E−02 4.46E−03 1.76E−03 1.89E−02 6.11E−03

R elbow, forearm, wrist 4.46E−03 1.76E−02 4.46E−03 1.76E−03 1.89E−02 6.11E−03

R side of head (including ear) 4.52E−03 1.76E−03 4.46E−03 1.76E−03 1.89E−02 6.11E−03

L side of head (including ear) 4.52E−03 1.76E−03 4.46E−03 1.76E−03 1.89E−02 6.11E−03

Front left thigh 6.00E−06 5.05E−06 6.00E−06 5.05E−06 3.58E−05 2.40E−05

Back left thigh 6.00E−06 5.05E−06 6.00E−06 5.05E−06 3.58E−05 2.40E−05

Front right thigh 6.00E−06 5.05E−06 6.00E−06 5.05E−06 3.58E−05 2.40E−05

Back right thigh 6.00E−06 5.05E−06 6.00E−06 5.05E−06 3.58E−05 2.40E−05

L knee and below 2.19E−06 1.85E−06 2.19E−06 1.85E−06 1.31E−05 8.77E−06

R knee and below 2.19E−06 1.85E−06 2.19E−06 1.85E−06 1.31E−05 8.77E−06

L side of face 4.52E−03 4.16E−04 4.46E−03 4.16E−04 1.89E−02 3.35E−03

R side of face 4.52E−03 4.16E−04 4.46E−03 4.16E−04 1.89E−02 3.35E−03

L side of neck 4.52E−03 1.76E−02 4.46E−03 1.76E−03 1.89E−02 6.11E−03

R side of neck 4.52E−03 1.76E−02 4.46E−03 1.76E−03 1.89E−02 6.11E−03

Back of head 4.52E−03 1.76E−03 4.46E−03 1.76E−03 1.89E−02 6.11E−03

Front of neck 4.52E−03 4.16E−04 4.46E−03 4.16E−04 1.89E−02 3.35E−03

Back of neck 4.52E−03 1.76E−02 4.46E−03 1.76E−03 1.89E−02 6.11E−03

Front torso: base of neck to end of sternum 7.26E−03 3.84E−04 7.26E−03 3.84E−04 4.67E−02 2.08E−03

Front torso: end of sternum to lowest rib 7.26E−03 3.84E−04 7.26E−03 3.84E−04 4.67E−02 2.08E−03

Front torso: lowest rib to iliac crest 7.26E−03 3.84E−04 7.26E−04 3.84E−05 4.67E−03 2.08E−04

Front torso: iliac crest to pubis 7.26E−04 3.84E−05 7.26E−04 3.84E−05 4.67E−03 2.08E−04

Back torso: base of neck to mid-back 7.26E−03 1.76E−02 7.26E−03 1.76E−02 4.67E−02 6.11E−02

Back torso: mid-back to lowest rib 7.26E−03 1.76E−02 7.26E−03 1.76E−02 4.67E−02 6.11E−02

Back torso: lowest rib to iliac crest 7.26E−03 1.76E−02 7.26E−04 1.76E−03 4.67E−03 6.11E−03

Back torso: buttocks (iliac crest and below) 7.26E−04 1.76E−03 7.26E−04 1.76E−03 4.67E−03 6.11E−03

Right torso: base of neck to end of sternum 4.46E−02 1.76E−02 4.46E−02 1.76E−02 1.89E−01 6.11E−02

Right torso: end of sternum to lowest rib 4.46E−02 1.76E−02 4.46E−02 1.76E−02 1.89E−01 6.11E−02

Right torso: lowest rib to iliac crest 4.46E−02 1.76E−02 4.46E−03 1.76E−03 1.89E−02 6.11E−03

Right torso: iliac crest to pubis (R hip) 4.46E−03 1.76E−03 4.46E−03 1.76E−03 1.89E−02 6.11E−03

Left torso: base of neck to end of sternum 1.96E−04 1.76E−02 1.96E−04 1.76E−02 1.41E−03 6.11E−02

Left torso: end of sternum to lowest rib 1.96E−04 1.76E−02 1.96E−04 1.76E−02 1.41E−03 6.11E−02

Left torso: lowest rib to iliac crest 1.96E−04 1.76E−02 1.96E−05 1.76E−03 1.41E−04 6.11E−03

Left torso: iliac crest to pubis (L hip) 1.96E−05 1.76E−03 1.96E−05 1.76E−03 1.41E−04 6.11E−03

a. The general method for calculating skin doses from chest projections is found in ORAUT-OTIB-0006, Section 6.0 and Tables B-4 and B-8 (ORAUT 2019a). 

b. Procedures after 1989 were performed at LLNL (ORAUT 2006a); doses are as found in ORAUT (2010a).
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 30 of 107

 

Table 3-6. Organ doses from lumbar spine X-ray, 1956 to 1971 (rem). 

Organ

Estimated dosea,b

HVL = 2.5 mm Al (collimated)

LAT

Estimated dosea,b

HVL = 2.5 mm Al (collimated)

AP

Thyroid 3.88E−05 2.84E−04

Eye/brain 3.88E−05 2.84E−04

Ovaries 7.10E−01c 5.60E−01c

Liver/gall bladder/spleen 7.10E−01c 5.60E−01c

Urinary bladder 7.10E−01c 5.60E−01c

Colon/rectum 7.10E−01c 5.60E−01c

Testes 5.60E−02 2.70E−02c

Lungs 5.44E−02 7.47E−02

Thymus 5.44E−02 7.47E−02

Esophagus 5.44E−02 7.47E−02

Stomach 7.10E−01c 5.60E−01c

Bone surfaces 7.10E−01c 5.60E−01c

Remainder 7.10E−01c 5.60E−01c

Female breast 5.05E−02 2.36E−02

Uterus 1.20E−01 2.71E−01

Bone marrow 8.54E−02 3.50E−02

Skind 5.26E+00 1.28E+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 analogues) for lumbar spine reflect actual 

measurements reported in Lincoln and Gupton (1958).

d. Skin dose values include backscatter factors of 1.355 from Table B.8 of NCRP (1997).

Table 3-7. Skin doses from lumbar spine X-ray, 1956 to 1971.

a

Skin area LAT AP

R front shoulder 5.26E−01 1.28E−01

R back shoulder 5.26E−01 2.79E−03

L front shoulder 2.31E−03 1.28E−01

L back shoulder 2.31E−03 2.79E−03

R upper arm to elbow 5.26E−01 1.28E−01

L upper arm to elbow 2.31E−03 1.28E−01

L hand 2.31E−03 1.28E+00b

R hand 5.26E−01 1.28E+00b

L elbow, forearm, wrist 2.31E−03 1.28E+00b

R elbow, forearm, wrist 5.26E−01 1.28E+00b

R side of head (including ear) 3.88E−05 2.84E−04

L side of head (including ear) 3.88E−05 2.84E−04

Front left thigh 2.31E−03 1.28E−01

Back left thigh 2.31E−03 2.79E−03

Front right thigh 5.26E−01 1.28E−01

Back right thigh 5.26E−01 2.79E−03

L knee and below 5.32E−04 2.76E−04

R knee and below 5.32E−04 2.76E−04

L side of face 3.88E−05 2.84E−04

R side of face 3.88E−05 2.84E−04

L side of neck 3.88E−05 2.84E−04

R side of neck 3.88E−05 2.84E−04

Back of head 3.88E−05 2.84E−04

Front of neck 3.88E−05 2.84E−04

Back of neck 3.88E−05 2.84E−04

Front torso: base of neck to end of sternum 5.44E−02 1.28E−01

Front torso: end of sternum to lowest rib 5.44E−02 1.28E+00
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 31 of 107

 

Skin area LAT AP

Front torso: lowest rib to iliac crest 5.44E−02 1.28E+00

Front torso: iliac crest to pubis 5.44E−02 1.28E+00

Back torso: base of neck to mid-back 5.44E−02 2.79E−03

Back torso: mid-back to lowest rib 5.44E−02 2.79E−02

Back torso: lowest rib to iliac crest 5.44E−02 2.79E−02

Back torso: buttocks (iliac crest and below) 5.44E−02 2.79E−02

Right torso: base of neck to end of sternum 5.26E−01 1.28E−01

Right torso: end of sternum to lowest rib 5.26E+00 1.28E+00

Right torso: lowest rib to iliac crest 5.26E+00 1.28E+00

Right torso: iliac crest to pubis (R hip) 5.26E+00 1.28E+00

Left torso: base of neck to end of sternum 2.31E−03 1.28E−01

Left torso: end of sternum to lowest rib 2.31E−02 1.28E+00

Left torso: lowest rib to iliac crest 2.31E−02 1.28E+00

Left torso: iliac crest to pubis (L hip) 2.31E−02 1.28E+00

a. The general method for calculating skin doses from chest projections is found in ORAUT-OTIB-0006, Section 6.0 and 

Tables B-5 and B-9 (ORAUT 2019a). 

b. Guidance for these AP doses in 1971 is 10% of the entrance skin dose or 1.18E–01 rem (ORAUT 2019a). 

Table 3-8. Distributions and corresponding statistical parameters for the dose to the B-lymphocytes.

Projection and period Distribution Parameter 1 Parameter 2 Parameter 3

PA chest 1956–1970 Weibull3 2.923411 0.002773 8.07816E−06

LAT chest 1956–1970 Weibull3 2.679824 0.003979 1.59961E−05

AP lumbar spine 1956–1971 Weibull3 3.118187 0.4762602 −0.002715842

LAT lumbar spine 1956–1971 Weibull3 3.171354 0.6082226 −0.002241317

PA chest 1971–1985 Weibull3 2.035671 0.001904 2.77200E−06

LAT chest 1971–1985 Weibull3 2.063026 0.003033 1.81402E−06

PA chest 1986–1989 Weibull3 2.0769196 0.008583 3.23658E−06

LAT chest 1986–1989 Weibull3 2.131524 0.016496 −2.07377E−06

In theory, a large number of factors can introduce uncertainties or affect X-ray machine output and 

dose to the worker. However, in practice, only five factors can be reasonably considered to have a 

meaningful or significant impact on dose uncertainty. These are (1) measurement error, (2) variation 

in applied kilovoltage, (3) variation in beam current, (4) variation in exposure time, and (5) SSD. The 

influence of other factors such as film speed, the use of screens, the use of grids, and development, 

while potentially variable, do not affect the beam output per se, except indirectly because they 

influence the machine settings (i.e., kVp, mA, and time). 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 with X-ray 

exposures are from ORAUT (2019a), which this analysis relied on for default information when sitespecific 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 (2019a) 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 
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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 (2019a) 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 (2019a) assumes uncertainty in beam output due to timers to be ±25%. 

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 (2019a), 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 preemployment X-ray record form. If the claim file is missing the preemployment X-ray 

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

the default for the preemployment 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 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 document are 

favorable to claimants and consistent with practices during each period.

Consistent with ORAUT (2019a), 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%. An estimate of 30% uncertainty is favorable to claimants.

4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

4.1 INTRODUCTION

The SNL-CA site is directly south of LLNL. The southern perimeter fence of LLNL is directly across 

the street from the northern perimeter fence of SNL-CA. Therefore, environmental dose is addressed 

in terms of potential exposures that could have occurred on the SNL-CA site as a result of both 

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

are assumed to be due to LLNL operations alone. According to Holland (1998), no other measurable 

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

perimeter, as measured by thermoluminescent dosimeters (TLDs), has been reported from 1967 to 

the present. These measurements are also considered in estimating external dose for SNL-CA.

4.1.1 Purpose

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

beginning of operations (1956) to the present. 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.
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4.1.2 Scope

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

separately in this section. Section 4.2 presents information necessary to estimate internal 

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

rates) for radionuclides that are 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 (Gudiksen et al. 1972, 1973; Silver et al. 1974 to 1980; Toy 

1981; Auyong, Griggs, and Buddemeier 1982; Griggs, Myers, and Buddemeier 1984; Griggs and 

Buddemeier 1986; Holland, Buddemeier, and Brekke 1987; Holland and Brekke 1988; Kamelgarn 

1989; Gallegos et al. 1990; Schwoegler et al. 1991; Gallegos et al. 1992 to 1994; Harrach et al. 1995 

to 1998; Larson et al. 1999, 2000; Biermann et al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 

2004; Peterson et al. 2005; LRL 1962a,b, 1963 to 1971; Devlin 1986 to 1988; Siegfriedt 1989; Brekke 

1990,1991; Brekke and Holland 1992 to 1995; Holland and Brekke 1996; Holland 1997 to 2002; 

Larsen 2003 to 2005; SNL 1982, 1983, 1984).

Section 4.3 contains information necessary for estimating external environmental dose. Ambient 

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

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

Section 4.3 also considers uncertainties in the information provided for estimating occupational 

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

information where possible.

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE CONCENTRATIONS

4.2.1 Radionuclides of Concern

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

was released (Section 2.3.2); however, such tritium releases provided annual doses of less than 

1 mrem/yr (<0.001 rem/yr) and are considered insignificant. 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 Brekke (1990) 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 1989). These mitigating factors would prevent the contribution of DU to any 

environmental intakes.

However, the presence of measurable concentrations of 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. Environmental tritium measurements for LLNL for all years indicate environmental 

doses of less than 1 mrem (<0.001 rem) and are not considered significant (ORAUT 2020). LLNL has 

processed and handled a number of radionuclides, including uranium and transuranic elements, 

mixed fission products, and accelerator-produced isotopes. Plutonium-239 and isotopes of uranium 

have been identified as radionuclides of significance in ORAUT-TKBS-0035-4, Lawrence Livermore 

National Laboratory – Occupational Environmental Dose (ORAUT 2020), 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 in relation to environmental dose, the concentrations at the LLNL perimeter 
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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 1962a, 

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

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

offsite 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 (8.1 × 102

pCi/L); the maximum net alpha concentration was 5.9 × 10-5 Bq/m3 (1.6 pCi/L). When not zero (i.e., 

when the concentrations at the perimeter were not less than average measured offsite 

concentrations), committed organ doses associated with the alpha- and beta-emitting particulate 

radionuclides were calculated, assuming an annual inhalation rate of 2,400 m3, 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 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 in comparison with other possible 

representative radionuclides, so 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 contributes less 

than 5% of the committed dose but always contributes less than 1 mrem to the yearly committed 

organ dose associated with environmental exposures for this 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 

(Gudiksen et al. 1972, 1973; Silver et al. 1974 to 1980; Toy 1981; Auyong, Griggs, and Buddemeier 

1982; Griggs, Gonzalez, and Buddemeier 1983; Griggs, Myers, and Buddemeier 1984; Griggs, 

Meyers, and Buddemeier 1985; Griggs and Buddemeier 1986; Holland, Buddemeier, and Brekke 

1987; Holland and Brekke 1988; Kamelgarn 1989; Gallegos et al. 1990; Schwoegler et al. 1991; 

Gallegos et al. 1992 to 1994; Harrach et al. 1995 to 1998; Larson et al. 1998, 1999, 2000; Althouse et 

al. 2000, Biermann et al. 2001; Gallegos et al. 2001, 2002; Sanchez et al. 2002, 2003, 2004; LLNL 

2004 to 2017; Peterson et al. 2005). 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 239Pu between 1971 and 2004 was 2.6 × 10-6 Bq/m3 (7.0 × 10-2 pCi/L), 

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

2,400 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 assumption that these intakes are both of 

234U is favorable to the claimant; the corresponding committed maximum organ dose for a 1-year 

intake is 6.7 mrem.

These measured concentration data and dose calculations indicate that particulate airborne betaemitting radionuclides 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, which indicate 

that the gross beta activity is due to global fallout and that it fluctuates in a manner typical of that 

source (Gudiksen et al. 1972, 1973; Silver et al. 1974 to 1980; Toy 1981; Auyong, Griggs, and 

Buddemeier 1982; Griggs, Myers, and Buddemeier 1984; Griggs and Buddemeier 1986; Holland, 

Buddemeier, and Brekke 1987; Holland and Brekke 1988; Kamelgarn 1989; Gallegos et al. 1990; 

Schwoegler et al. 1991; Gallegos et al. 1992 to 1994; Harrach et al. 1995 to 1998; Larson et al. 1999, 

2000; Biermann et al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 2004; Peterson et al. 2005). 
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The alpha-emitting radionuclides 234U and 239Pu (represented as gross alpha before 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 4.2.3 below.

4.2.2 Source Terms for Internal Dose

No onsite environmental sources or outdoor concentrations of radionuclides were reported for SNLCA before the beginning of operation of the TRL in 1979. Release of radionuclides from activities at 

LLNL could have affected intakes by workers at SNL-CA, as noted in Section 4.2.1 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 are variable throughout the 

remainder of the year (Holland and Brekke 1988); therefore, this assumption is most favorable to the 

claimant during these months.

4.2.3 Annual Intake of Radioactivity

Inhalation intakes of airborne particulate 234U and 239Pu are estimated in this section. 

To calculate inhalation intake, it was necessary to consider the onsite air concentrations of 234U and

239Pu due to the presence of these isotopes near the LLNL south boundary (closest to the SNL-CA). 

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

2,400 m3/yr. For airborne particulates (i.e., 234U and 239Pu), no particle size information was available; 

therefore, the default ICRP Publication 66 value of 5-μm activity median aerodynamic diameter is 

recommended (ICRP 1994). Further, no solubility information is available for airborne uranium or 

plutonium particulates; therefore, assumed solubility should be selected based on what is most 

favorable to the claimant in light of the organ of interest. 

4.2.3.1 Inhalation Intakes

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

provided in Tables 4-2 and 4-3. The basis for the intake values in Tables 4-2 and 4-3 is documented 

in ORAUT (2020).

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

Radionuclide

Applicable 

period Method used

H-3 1956–1978, 

1997–2018

Potential environmental H-3 releases from LLNL from all sources were 

determined to provide internal doses of less than 1 mrem/year (<0.001 

rem/year). They are not considered significant for environmental dose for 

LLNL employees and are, therefore, not considered significant for employees 

at SNL-CA (ORAUT 2020).

H-3 1979–1996 Potential intakes during the operation of the TRL, along with intake data from

all sources at LLNL (ORAUT 2020), were evaluated and determined to result 

in potential internal doses of less than 1 mrem/year (<0.001 rem/year). They 

are not considered significant for environmental dose for employees at 

SNL-CA.

U-234 1956–1960 No data available for this period.

U-234 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 2,400 m3/yr to estimate maximum 

SNL-CA intakes (Bq/yr).
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Radionuclide

Applicable 

period Method used

U-234 1971–2018 Reported south LLNL perimeter U-235 and -238 activity concentrations 

(Bq/m3) were multiplied by an inhalation intake rate of 2,400 m3/yr to estimate 

maximum SNL-CA intakes (Bq/yr). The unreported U-234 concentrations 

were addressed by assuming U-234 intake was equivalent to U-238. Activity 

intakes for all radionuclides are added and assumed to be U-234. 

Pu-239 1956–1960 No data available for this period.

Pu-239 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 2,400 m3/yr to estimate maximum 

SNL-CA intakes (Bq/yr).

Pu-239 1971–2018 Reported south LLNL perimeter Pu-239 activity concentrations (Bq/m3) were 

multiplied by an inhalation intake rate of 2,400 m3/yr to estimate maximum 

SNL-CA intakes (Bq/yr).

Table 4-2. Sitewide annual median inhalation 

intakes (Bq/yr), 1956 to 1970.

a,b 

Year U-234 or Pu-239c

1956 (d)

1957 (d)

1958 (d)

1959 (d)

1960 (d)

1961 1.24E-01

1962 1.36E-01

1963 5.77E-02

1964 6.39E-02

1965 0.00e

1966 0.00e

1967 0.00e

1968 0.00e

1969 0.00e

1970 0.00e

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

b. Intakes are assigned as lognormal distributions, 

with a geometric standard deviation (GSD) of 3.

c. Gross alpha is reported for 1961 to 1970; 

recommend assuming the isotope of either U-234 

or Pu-239 that gives highest dose to the organ of 

interest.

d. No measurement results were reported for these 

years. 

e. A zero value indicates the measured concentration 

was less than or equal to the offsite background 

concentration.

Table 4-3. Sitewide annual median inhalation 

intakes (Bq/yr), 1971 to 2018.

a,b

Year U-234c Pu-239c

1971 3.63E-03 6.30E-03

1972 6.26E-03 2.84E-03

1973 5.75E-03 1.51E-03

1974 5.07E-03 2.93E-03

1975 4.82E-03 2.04E-03

1976 5.63E-03 7.99E-04

1977 7.63E-03 2.13E-03
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Year U-234c Pu-239c

1978 1.07E-03 2.66E-03

1979 5.45E-03 1.51E-03

1980 3.75E-03 4.44E-04

1981 7.44E-03 1.42E-03

1982 6.77E-03 4.44E-04

1983 9.31E-03 2.17E-03

1984 6.01E-03 3.55E-04

1985 5.57E-03 1.78E-04

1986 4.89E-03 1.15E-04

1987 5.88E-03 7.10E-05

1988 1.37E-02 7.99E-05

1989 4.53E-03 6.04E-05

1990 5.92E-03 1.24E-04

1991 6.12E-03 1.69E-04

1992 5.27E-03 0.00d

1993 4.37E-03 8.64E-05

1994 3.41E-03 8.18E-05

1995 2.82E-03 5.88E-05

1996 2.86E-03 5.76E-05

1997 3.04E-03 1.56E-05

1998 2.13E-03 1.42E-05

1999 4.11E-03 1.32E-05

2000 0.00d 2.18E-05

2001 0.00d 9.19E-06

2002 2.43E-03 5.74E-06

2003 1.45E-03 5.98E-06

2004 1.38E-03 7.39E-06

2005 9.58E-04 7.73E-05

2006 1.03E-03 3.36E-06

2007 1.07E-03 0.00d

2008 1.30E-03 0.00d

2009 1.12E-03 5.98E-06

2010 6.51E-04 6.77E-06

2011 8.64E-04 1.93E-06

2012 1.25E-03 0.00d

2013 1.01E-03 2.76E-06

2014 1.01E-03 2.10E-06

2015 8.14E-04 0.00d

2016 8.58E-04 0.00d

2017 9.39E-04 2.19E-03

2018 7.95E-04 0.00d

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

b. Intakes are assigned as lognormal distributions, with a 

GSD of 3.

c. Intakes based on LLNL south perimeter 

concentrations; see Table 4-1.

d. Median activities for uranium and plutonium samples 

were negative and are reported as zero.

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 2018 (Gudiksen et al. 1972, 

1973; Silver et al. 1974 to 1980; Toy 1981; Auyong, Griggs, and Buddemeier 1982; Griggs, Gonzalez, 

and Buddemeier 1983; Griggs, Myers, and Buddemeier 1984; Griggs, Meyers, and Buddemeier 1985; 

Griggs and Buddemeier 1986; Holland, Buddemeier, and Brekke 1987; Holland and Brekke 1988; 
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Kamelgarn 1989; Gallegos et al. 1990; Schwoegler et al. 1991; Gallegos et al. 1992 to 1994; Harrach 

et al. 1995 to 1998; Larson et al. 1998, 1999, 2000; Althouse et al. 2000, Biermann et al. 2001; 

Gallegos et al. 2001, 2002; Sanchez et al. 2002, 2003, 2004; LLNL 2004 to 2017; Peterson et al. 

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 (Gallegos et al. 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 1962a, 1963 to 1971). From 1961 

through 1970, the net alpha measurements, which were 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 4-2). Net concentrations calculated to be less than zero were 

assumed to be zero. The assumed radionuclide should be the one that gives the highest dose to the

organ of interest. For the 5 years of operation before 1961, there are no measurements available with 

which to estimate the intakes of 234U and 239Pu. 

4.3 EXTERNAL EXPOSURE TO ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

4.3.1 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 Building 

(Brekke and Holland 1993). A 1975 aerial survey of the LLNL and SNL-CA in 1975 found gammaemitting 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, 133Ba, and 235U (Building 921, incorrectly referred to as B-291 in 

the report; Tipton 1977). The highest estimated exposure rate at the center of the Radiography 

Building (as seen in the flyover) in excess of background gamma was 40 to 70 μR/hr (100 to 

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 (5 to 20 mrem/yr) for a 2,500 hr/yr occupational exposure. It was 

indicated in the 1977 annual monitoring report (Silver et al. 1978) that none of the elevated areas 

posed a radiation hazard to workers. Further, the survey indicated that the sources were restricted to 

work areas where access was limited. Therefore, 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 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 that had 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 (Gudiksen et al. 1972). One of these locations (Location 5 – 

neutron dosimeter) was adjacent to the LLNL cyclotron building and is at the south perimeter of the 

LLNL site (directly across the street from the north perimeter of SNL-CA). Therefore, this LLNL 

perimeter location (south location) was considered in evaluating external exposures. Environmental 
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neutron measurements were also made at this location and reported in annual reports beginning in 

1973. These are discussed in Section 4.3.3 below.

4.3.2 Gamma-Emitting Radionuclides

The use of TLDs to measure environmental radiation exposure at the SNL-CA site perimeter 

apparently began around 1989, which was the first time that the five SNL-CA perimeter TLDs 

(Figure 4-1) are mentioned in annual environmental reports (Brekke 1990). Before that time, the site 

relied on LLNL perimeter and offsite measurements because 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 

are applicable to an exposure duration of 2,500 hr/yr (50 hr/wk, 50 wk/yr). The basis for the external 

doses in Table 4-4 is documented in ORAUT (2020).

Figure 4-1. TLD locations since 1989.

The data preference used in determining the values included in Table 4-4 was to include SNL-CA 

measurements when available, but this only occurred from 1990 to 1994 (Brekke 1991; Brekke and 

Holland 1992 to 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 expected 

environmental exposure rates for the SNL-CA. The 1975 aerial survey indicated this measurement 

point coincided with the LLNL accelerator and therefore had a slightly elevated exposure rate on the 

order of that seen for the storage vaults and holding area at SNL-CA (Tipton 1977). Because 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.

There were no reliable measurements for either SNL-CA or the south perimeter LLNL location before 

1967. The values for 1967 to 1969 in Table 4-4 assume that the “experimental physics facility” 

mentioned in the corresponding annual reports (LRL 1968 to 1970) represents the south perimeter 
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Figure 4-2. Location of LLNL perimeter gamma and neutron 

dosimeters (Griggs and Buddemeier 1986). 

Figure 4-3. Location of LLNL offsite gamma dosimeters (Holland and 

Brekke 1988). 

location that is labeled Location 5 from 1970 to 1980 (LRL 1971, Gudiksen et al. 1972, 1973; Silver et 

al. 1974 to 1980; Toy 1981), but is relabeled Location 11 after 1980 (Auyong, Griggs, and 

Buddemeier 1982; Griggs, Myers, and Buddemeier 1984; Griggs and Buddemeier 1986; Holland, 

Buddemeier, and Brekke 1987; Holland and Brekke 1988; Kamelgarn 1989; Gallegos et al. 1990; 
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 41 of 107

 

Table 4-4. External gamma radiation dose based on 2,500 hr/yra exposure duration (mrem/yr).

Year

Average 

perimeterb Errorc 

Bounding 

dose rated

1956 51e 15 66

1957 51e 15 66

1958 51e 15 66

1959 51e 15 66

1960 51e 15 66

1961 51e 15 66

1962 51e 15 66

1963 51e 15 66

1964 51e 15 66

1965 51e 15 66

1966 51e 15 66

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

1981 16 5 23

1982 19f 6 25

1983 21 6 27

1984 22f 7 28

1985 23 7 30

1986 17 5 23

1987 17 5 22

Year

Average 

perimeterb Errorc 

Bounding 

dose rated

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 21 6 27

1995 16 5 21

1996 16 5 21

1997 17 5 22

1998 17 5 22

1999 17 5 22

2000 16 5 21

2001 16 5 22

2002 19 6 25

2003 16 2 18

2004 16 2 18

2005 17 2 19

2006 17 2 18

2007 17 3 19

2008 16 2 18

2009 17 1 18

2010 16 1 17

2011 16 2 18

2012 16 2 18

2013 17 2 18

2014 16 2 18

2015 16 2 18

2016 16 2 18

2017 16 1 17

2018 17 1 18

a. 2,500 hr/yr taken from ORAUT (2019b). 

b. Until 1990, unless otherwise noted, values are the south perimeter LLNL dose rate; from 1990 through 2002, 

values are the average of SNL-CA perimeter dosimeters; from 2003 through 2018, values are the average of 

the LLNL perimeter dosimeters.

c. Error assumed to be ±30% of the higher of the average perimeter (ORAUT 2012b) through 2002. From 2003, 

the error is the difference between the LLNL maximum and average values.

d. Bounding dose rate is the average perimeter value plus error term. Bounding dose rates can vary due to 

rounding of average perimeter and error values. 

e. Assumed average dose rate estimated for the perimeter from 1967 through 1977 – no measurements reported 

for the LLNL south perimeter or for SNL-CA from 1956 through 1966.

f. Assumed the average of the previous year and the year after – missing annual environmental reports for 1982 

and 1984.

Schwoegler et al. 1991; Gallegos et al. 1992 to 1994; Harrach et al. 1995 to 1998; Larson et al. 1999, 

2000; Biermann et al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 2004; Peterson et al. 2005; 

Matthews et al. 2006, 2007; Gallegos et al. 2008, 2009; Jones et al. 2010 through 2014; Rosene et al.

2015 through 2018). This is a reasonable assumption because 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 the LLNL 

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 
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values due to the inability to find annual reports for 1982 and 1984. Last, SNL-CA TLD 

measurements are included for the years they were reported: 1990 through 1994. For these years, 

the “average perimeter” dose rate in Table 4-4 represents the average of the reported SNL-CA 

measurements. After 1994, the average perimeter dose rate was estimated by averaging the 5 years’ 

worth of SNL-CA data.

Background radiation measurements for the Livermore valley have been reported in LLNL annual 

reports since 1971 (Gudiksen et al. 1972, 1973; Silver et al. 1974 to 1980; Toy 1981), but is relabeled 

Location 11 after 1980 (Auyong, Griggs, and Buddemeier 1982; Griggs, Myers, and Buddemeier 

1984; Griggs and Buddemeier 1986; Holland, Buddemeier, and Brekke 1987; Holland and Brekke 

1988; Kamelgarn 1989; Gallegos et al. 1990; Schwoegler et al. 1991; Gallegos et al. 1992 to 1994; 

Harrach et al. 1995 to 1998; Larson et al. 1999, 2000; Biermann et al. 2001; Gallegos et al. 2002; 

Sanchez et al. 2003, 2004; Peterson et al. 2005; Matthews et al. 2006, 2007; Gallegos et al .2008, 

2009; Jones et al. 2010 through 2014; Rosene et al. 2015 through 2018).

Figure 4-4 is a summary of the quarterly results from 1988 through 1996; it compares LLNL average 

perimeter measurements and average LLNL Site 300 measurements to the offsite 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 onsite ambient dose rate, which was calculated 

from the yearly rate for these 9 years (Harrach et al. 1997), is 57 mrem/yr for continuous exposure or 

16 mrem/yr for a 2,500-hr/yr occupational exposure, as reflected in Table 4-4.

Figure 4-4. Livermore valley radiation background measurements in comparison with annual average 

measurements at LLNL site perimeter and LLNL Site 300 (Harrach et al. 1997). 

4.3.3 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) (Silver et al. 1974 to 1980; Toy 1981; Auyong, Griggs, and 

Buddemeier 1982; Griggs, Myers, and Buddemeier 1984; Griggs and Buddemeier 1986; Holland, 
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Buddemeier, and Brekke 1987; Holland and Brekke 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 

because the operations responsible for neutron radiations were discontinued (Harrach et al. 1996). 

Measurements were initially made at Location 5 (Figure 4-2) (Silver et al. 1974), but a monitor at 

Location 3 (originally called 5a) was added in 1975 (Silver et al. 1976). By 1987, the dose rates were 

at background neutron levels, which is approximately 4 mrem/yr for continuous exposure (Harrach et 

al. 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 ft) (Willhoite 1979). Assuming that this is also the

approximate distance to the nearest SNL-CA perimeter (which currently is across a four-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 in Table 4-5 were also corrected for an occupational 

exposure duration of 2,500 hr/yr and background contributions. These values are favorable to the 

claimant because the dose rate drops off fairly rapidly throughout other areas of the SNL-CA site. 

Due to the lack of information for years before 1973 about 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). The 

basis for the external doses in Table 4-5 is documented in ORAUT (2020).

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 Bounding 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 6 2 8

1980 6 2 8

1981 2 1 3

1982 8f 2 10

1983 8 2 10

1984 8f 2 10

1985 <1 N/Ag 1.3h

1986 2 1 3

1987 <1 N/Ag 1.3h

1988 <1 N/Ag 1.3h

1989 <1 N/Ag 1.3h

1990 <1 N/Ag 1.3h

1991 <1 N/Ag 1.3h

1992 <1 N/Ag 1.3h
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Year North perimeter doseb Errorc Bounding dosed

1993 <1 N/Ag 1.3h

1994 <1 N/Ag 1.3h

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, which was 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 2012b).

d. Bounding dose rate is the average perimeter value plus error term. Bounding dose 

rates may vary due to rounding of average perimeter and error values.

e. Assumed value reported for 1973 applied to previous years back to 1966.

f. Assumed the 1983 value; annual environmental reports for 1982 and 1984 not found.

g. N/A = not applicable.

h. Bounding dose rates for 1985 and 1987 are based on rounding up north perimeter 

dose to 1.000 mrem/yr and adding a 30% error, as described above.

To be favorable to claimants and consistent with ORAUT (2010b), a neutron energy range of 0.1 to 

2.0 MeV should be chosen along with an International Commission on Radiological Protection 

Publication 60 weighting factor of 1.91 (ICRP 1991).

5.0 INTERNAL DOSIMETRY

5.1 INTRODUCTION

5.1.1 Purpose

The purpose of this section is to describe internal dosimetry practices at SNL-CA to support dose 

reconstructions under EEOICPA.

5.1.2 Scope

Section 5.2 summarizes potential internal exposures to radionuclides. Section 5.3 describes the 

bioassay monitoring programs SNL-CA used over the years, and Section 5.4 discusses laboratory 

procedures and minimum detectable intakes. Section 5.5 discusses the results of the bioassay 

programs in terms of dose. Section 5.6 provides a method of calculating dose from urine bioassay 

data, and Section 5.7 discusses airborne radionuclide concentrations. Section 5.8 discusses 

estimates of unmonitored dose, and Section 5.9 describes the radiation dosimetry reports. 

Section 5.10 provides summary tables of the above information.

5.1.3 Special Exposure Cohort

NIOSH has determined that it is not feasible to reconstruct internal dose from October 1, 1957, 

through December 31, 1994, due to a lack of sufficient information, which includes biological and 

workplace monitoring data and radiological source information that would allow it to estimate potential 

internal exposure to tritium, highly enriched uranium (HEU) and DU, uranium tritides and hydrides, 

thorium, and classified activities (NIOSH 2013). Therefore, this period has been included in the SEC. 

Dose reconstruction guidance in this document for this period is presented to provide a technical 

basis for partial dose reconstructions for nonpresumptive cancers not covered in the SEC class or for 

claims with less than 250 working days in the SEC period.

Although NIOSH found that it is not possible to reconstruct internal radiation doses completely for the 

proposed class, it intends to use internal monitoring data that might become available for an individual 

claim (and that can be interpreted using existing NIOSH dose reconstruction processes or 

procedures). Therefore, dose reconstructions for individuals employed at SNL-CA from October 1, 
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1957, through December 31, 1994, but who do not qualify for inclusion in the SEC, can be performed 

using these data as appropriate to support a partial dose reconstruction.

5.2 SUMMARY OF POTENTIAL INTERNAL RADIONUCLIDE EXPOSURES

Work with radionuclides that created a potential for internal exposure at SNL-CA included limited 

uranium operations that consisted primarily (but not exclusively) of DU machining and research in the 

TRL. Radiation workers were monitored using external dosimetry and urine bioassay. A radiation 

worker was defined as an “employee who received 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 estimating 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.

5.2.1 Tritium Research Laboratory

The TRL was established in 1976 to perform research and development for the DOE Office of 

Defense Programs to support weapons development. At its peak of operations, the TRL employed 

approximately 35 experimenters and 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. 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 g was doubly contained in gloveboxes, special Sandia-designed 

containers, or U.S. Department of Transportation-approved containers. The Safety Analysis Report 

(Wright 1981a) anticipated that 50 g of tritium might be handled in one doubly contained system and 

that the total tritium inventory in the facility would be approximately 300 g.

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).

5.2.2 Uranium Alloy Machining

Uranium alloy machining was performed at the SNL-CA facility starting before 1972. Several 

memoranda indicate that machining and testing of uranium alloys had previously been performed at 

the 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), which 

required initiation of an onsite program.

The radiation safety requirements for the machining operation are described in SOP Machining 

Depleted Uranium Metal (SNL 1989). The DU is designated in the SOP as D-38. The radiation safety 

requirements included air sampling and urine bioassay (SNL 1989). Daily air samples were required 

during all machining operations even though air-sampling data had shown that neither wet machining 
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nor “burning (D-38) chips” (SNL 1989) resulted in airborne particulates. Urine bioassay was required 

semiannually.

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 was installed at the TRL in 1991 and also presented a potential for exposure 

to uranium powder [TRL, SOP No. 757 (SNL 1991a)]. The operating procedure did not specify 

bioassay requirements. In addition, parts contaminated with UO2 dust were received by SNL-CA 

(Lovell 1982) and posed a potential airborne dust hazard.

5.3 BIOASSAY PROGRAMS

Limited bioassay data for tritium and uranium were found from SNL-CA. 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 1996). 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 3H 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 

3H and natural uranium, or that other types of bioassay (i.e., fecal analyses or in vivo counting) were 

employed at SNL-CA. The 3H bioassay was limited to TRL experimenters and staff. The uranium 

bioassay was performed on individuals who were involved in machining DU as well as others who 

were 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.

5.3.1 Tritium Bioassay Programs

According to Garcia and Gorman (1996), tritium bioassays were performed weekly for individuals who 

were involved in experimental work at the TRL from 1979 through 1995 when the TRL was 

decommissioned. Individuals working in the Waste Handling Facility might 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 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 that were analyzed by liquid scintillation counting (LSC). Laboratory analyses were 

performed in house by the Health Physics Division. Doses were calculated and reported on a monthly 

basis.

A 1993 memorandum 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 program, in 1988. In 1990, the files were sent to SNL-NM for entry into personnel dosimetry 

histories. As of 1993, the date of the memorandum, the bioassay data were collected by SNL-CA and 

a hardcopy was 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 were determined by LSC. At first, 
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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).

5.3.2 Uranium Bioassay Programs

Uranium bioassays were required for SNL-CA workers who were 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 1 × 10-10 μCi/cm3; 

• For routine handling of uranium hydrides, solutions of uranium compounds, and uranium 

powders (more than four times per quarter);

• Machining of uranium; and

• During any operation that the Hazards Control Division deemed hazardous or for which an 

SOP or SWP required air sampling.

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; and

• Skin contact with a solution of uranium.

DU alloy machining was performed under SOP 1066 (SNL 1989). The SOP required that all 

machining operations be performed wet, which reduced the risk of fire and generation of airborne 

dust. Revision C, dated October 1, 1989 (SNL 1989), required semiannual 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 the air samplers on and off. No air concentration data were available to the 

authors.

5.4 BIOASSAY LABORATORY PROCEDURES AND MINIMUM DETECTABLE INTAKES 

The bioassay measurements were performed by various groups during the period of operation of 

SNL-CA. No specific laboratory procedure manuals were available for either tritium or uranium 

analyses at SNL-CA.

5.4.1 Tritium

Tritium bioassay was performed at SNL-CA during the entire history of the TRL by LSC. 

Revision 3 of the technical basis document (TBD) for internal dosimetry at SNL-CA and SNL-NM 

states that workers were instructed to fill the entire 1,500 mL urinalysis container (Potter ca. 1997). 
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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 listed references.)

The frequency and minimum detectable amounts (MDAs) for tritium bioassay 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 ±2σ)

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 Unknown Calculated from LSC efficiency and background 

information in H-3 dose worksheets for 1980 

1981 Approximately 

10 nCi/L

(background ±2σ)

Weekly TRL Health Physics Summary, First Quarter, 1981

(Wright 1981a)

1986 Approximately 

20 nCi/L

Unknown Calculated based on computer printout for positive 

exposures for 1986

1988 None given Weekly Tritium Research Laboratory Health Physics 

Quarterly Summaries (October) (Author unknown 

1988)

1989 None given Unknown Tritium Research Laboratory Health Physics 

Quarterly Summaries (Author unknown 

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, but weekly for 

individuals working 

primarily in the TRL

SOP No. 709, Tritium Research Laboratory 

Building 968 (SNL 1991b)

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 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 Unknown Wright (1998) memorandum to Debbie Miller DOE

Albuquerque 

Special bioassay

According to Hafner (2006), tritium bioassay samples were counted in an LSC 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. 

An MDA of 10 nCi/L was 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 
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aliquot 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 from April 1982 through October 1982 

provided the results of the bioassay and the calculated dose.

The reviewed SNL-CA personnel records 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 . t = 0 286Δ μCi/L (5-2)

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.

5.4.2 Uranium

Uranium bioassay samples were sent to SNL-NM for analysis either in house or by a contract 

laboratory (SNL 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 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 1993a). Presumably because 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 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 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 1993a,b). However, subsequent intercomparison 

studies indicated that the data from CEP were not reliable. Therefore, dose reconstructors 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 came 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 dose reconstructor should assume a specific activity for DU when 

converting mass concentration to activity concentration. The most reasonable specific activity for DU 

is 4.38 × 10-7 Ci/g as stated by Wright (1979a).

Wright (1979a) states that the minimum detection limit (MDL) by the fluorometric analysis method at 

SNL-NM was 1 × 10-2 μg/L or 4 × 10-3 μCi/L. These two values are inconsistent, assuming the 

specific activity of 4.38 × 10-7 Ci/g for DU also stated by Wright. A mass concentration of 1 × 10-2 μg/L 

would be equivalent to an activity concentration of 4.38 × 10-9 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, which indicates an MDC of 0.1 μg/L (SNL 1975–1977). Other sample data 

sheets show MDAs of 0.01 μg (SNL 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.
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Table 5-2. Preliminary timeline for 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

Method 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 

Hygiene Services

ICP listed as 

Procedure No.

No information; all 

results reported as 0

It is unlikely that the method 

was ICP-MS. It did not come 

into general use until the early 

90s. ICP alone is not 

sensitive

1986–1987 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 information No MDC given Audit memoranda show that 

CEP analyses were 

problematic and therefore 

cannot be used

1993–

present

SNL-NM Industrial 

Hygiene Services

Not defined but 

probably ICP-MS

0.1 μg/L Reasonable MDAs for ICPMS

5.5 BIOASSAY RESULTS

Garcia and Gorman (1996) provide data on doses that were calculated from urine bioassay of TRL 

experimenters and staff. However, they provide 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

Number of workers 

receiving 101–500 mrem/yr

1979 4 1

1980 4 1

1981 6 0

For 1982 through 1995, Garcia and Gorman (1996) show maximum and average doses from intake of 

3H at TRL as well as total person-mrem (Table 5-4). The number of monitored workers can be 

inferred by dividing the total person-mrem by the average dose in millirem.

5.6 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; it does not include the actual urine bioassay concentration data. The methods of 
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Table 5-4. Estimated tritium doses (mrem/yr) based on bioassay, 1982 to 

1995 (Garcia and Gorman 1996). 

Year

Maximum

dose

Average 

dose

Total 

person-dose

No. of monitored 

individuals 

(inferred)

1982 70 30 183 6

1983 79 49 148 3

1984a 1,620 152 3,040 20

1984b 234 75 1,420 19

1985 347 65 2,270 35

1986 229 67 1,330 32

1987 178 42 580 14

1988 218 63 1,652 26

1989 232 46 2,465 54

1990 262 30 1,056 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

a. Maximum individual single event. 

b. Maximum, average, and total person-dose excluding maximum individual single event.

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 reviewed SNL-CA personnel records 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 based on Equation 5-2 above. Again, the origin of the 

constant, 0.286, is not defined.

5.7 AIRBORNE RADIONUCLIDE CONCENTRATIONS

Work in the TRL was conducted in sealed gloveboxes and/or in gloveboxes used in high-velocity air 

hood mode. The TRL used two systems for decontaminating glovebox air before release to the 

environment. The GPS removed tritium from sealed gloveboxes, and the VERS removed tritium from 

the glovebox pressure control system and the gases exhausted from all of the vacuum pumps in the 

laboratory. The TRL high-flow ventilation provided 6 to 10 room air changes per hour (SNL 1991b).

Tritium monitors were installed in TRL. The monitors were used to detect the release of tritium into 

room air, gloveboxes, and the stack as well as to monitor the performance of the GPS and VERS 

(SNL 1991b). The monitors activated audible and 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 from 1988 

through 1995. 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, 60 tritium monitors were in use (Author unknown 1989a,b,c,d).

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 Room Air High alarm was set at 1 mCi/m3. The 

maximum permissible concentration in air and the more recent derived air concentration (DAC) 

remained constant at 2 × 10-5 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 
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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 1 minute (SNL 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.

5.8 UNMONITORED DOSE ESTIMATES 

The vast majority of SNL-CA workers were employed in areas with little or no potential for intake of 

radionuclides that were generated at their work areas (Wright 1979b; Bryson 1972). Therefore, only a 

small proportion of workers were required to participate in the bioassay program. Individuals who 

worked in the laboratory areas as experimenters or support personnel could have been exposed to 

airborne tritium (SNL 1991a,b). Machinists and others who worked in areas where DU was machined 

or other uranium operations were conducted could have been exposed to airborne uranium dust (SNL 

1989).

For years after 1994, 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% of the airborne radionuclide concentrations limits because the procedures 

required monitoring for individuals who could have potentially received internal doses in excess of 

10% of the dose limit (Wright 1993) or when air sampling results demonstrated that air concentrations 

could reach or exceed 10% of the concentration guides for air (Wright 1979b). The applicable DAC 

values in place at the time in air are given in Table 5-5.

Table 5-5. DAC values (μCi/mL) for tritium and uranium (DOE 1993, Appendix A). 

Period HTOa

U-238b

type F

U-238b

type M

U-238b

type S

After 1994 2E−5 6E−10 3E−10 2E−11

a. Assigned through 1996, environmental intakes assigned thereafter.

b. Assigned through 1998, environmental intakes assigned thereafter.

The calculated annual intake for an unmonitored worker inhaling tritium at 10% of a DAC or 

2 × 10-6 μCi/mL, assuming a breathing rate of 1.2 m3/hr for 2,000 hr/yr, would be 4.8 × 103 μCi. After 

an adjustment of 1.5 for skin absorption, the intake would become 7.2 × 103 μCi. This would be 

applicable for 1995 and 1996 only, because decommissioning of the TRL was completed in 1996. 

After 1996, environmental intakes should be assigned. 

The calculated annual intakes for an unmonitored worker inhaling uranium at 10% of a DAC would be 

6 × 10-11 μCi for type F uranium, 3 × 10-11 μCi for type M uranium, and 2 × 10-12 μCi for type S 

uranium. These intakes would be applicable for 1995 to 1998. After 1998, environmental intakes 

should be assigned.

Because SNL-CA is close 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.

5.9 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 from 1973 

showed no internal body depositions or exposure to airborne materials that resulted in internal body 

deposition as determined by bioassays. 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 
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there were no exposures resulting in internal body deposition of radioactive materials. The reporting 

form changed in 1976; subsequent reports for 1977 and 1978 did not show internal exposures.

The Personnel Monitoring and Laboratory Services department at SNL-NM issues 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 CFR Part 835 limiting values (SNL 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.

Before 1973, external doses were documented by SNL-CA. Tritium doses acquired at sites other 

than SNL were noted occasionally in the records (SNL undated c). The notation showed values less 

than 3 μCi/L (presumably in urine). One record in this undated document showed “3 MC/L,” but the 

entries in this record were all in capital letters, and it was 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 

1980–1984).

5.10 SUMMARY TABLES 

Tables 5-6 to 5-9 incorporate the best general information available about MDAs. However, it should 

be noted that some individual data in the reviewed records showed lower MDAs or reported urine 

bioassay values less than the listed MDA.

Table 5-6. Internal dose control program. 

Routine monitoring type Period Frequency

Urine – tritium 1977–1995 Weekly

Urine – uranium 1968–present Semiannual or as required by the Hazards Control Department

Table 5-7. Detection limits for urine bioassay for 3H with LSC.

a

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

a. Reporting limits were based on dose rather than bioassay results.

Table 5-8. Detection limits for total uranium bioassay.a

Method Period MDA (μg/L)b

No data Before 1968 5 (based on state of technology)c

Fluorimetry 1968 5 (based on reported error)

Fluorimetry 1969–1973 5 (estimated based on 1968 data)

Fluorimetry 1973–1986 5 (individual employee bioassay reports as low as 0.7)

Fluorimetry 1986–1987 3

Fluorimetry 1987–1990 10

CEP 1991–1992 Audit memoranda show that CEP analyses were problematic and thus cannot be used.

ICP-MS 1993–present 0.1

a. No reporting limits were available in the documents provided to the authors.

b. Dose reconstructors should use the MDA data provided in the bioassay records when available.

c. MDA of 5 μg/L is consistent with the value reported for bioassay conducted at nearby LLNL site (ORAUT 2016) and the 

Hanford Site (ORAUT 2015). 
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Table 5-9. Facility, source, solubility type, and particle size. 

Facility Source Solubility type Particle size

TRL (Building 968) H-3 Tritiated water Not applicable

Building 913 Machine Shop DU Metallic form – unspecified solubility type but 

likely Type S

No data

Building 968 Uranium tritide Type Fa No data

Building 913 UO2 Type S No data

a. Described in Section 3.0 of ORAUT (2007b). 

6.0 OCCUPATIONAL EXTERNAL DOSE

6.1 INTRODUCTION

6.1.1 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. 

6.1.2 Scope

Historical documentation about 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 in this section draws on review of currently available records. The review for 

years before 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 will be 

revised accordingly.

6.1.3 Special Exposure Cohort

NIOSH has determined that it is not feasible to estimate with sufficient accuracy beta, gamma, and 

neutron external exposures from October 1, 1957, through December 31, 1994, resulting from HEU, 

uranium hydrides, Radiography Facility isotope sources, thorium, classified work, and resulting doses 

for the class of employees covered by this evaluation (NIOSH 2013). Therefore, this period has been 

included in the SEC. Dose reconstruction guidance in this document for this period is presented to 

provide a technical basis for partial dose reconstructions for nonpresumptive cancers not covered in 

the SEC class or for claims with less than 250 working days in the SEC period.

Although NIOSH found that it is not possible to completely reconstruct external radiation doses for the 

designated class, it intends to use any external monitoring data that might become available for an 

individual claim (and that can be interpreted using existing NIOSH dose reconstruction processes or 

procedures). Therefore, for individuals employed at SNL-CA from October 1, 1957, through 

December 31, 1994, who do not qualify for inclusion in the SEC, dose reconstructions can be 

performed using these data as appropriate to support a partial dose reconstruction.

6.2 BACKGROUND

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 2.2, 

the laboratory typically handled kilogram amounts of DU (typically in the form of alloyed metal 

components), gram amounts of 3H, and microcurie quantities of other isotopes. The Radiography 

Facility (Building 923) also contained 100-Ci 192Ir and 60Co sources, a sealed 252Cf source, and many 
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other smaller sealed sources of activity ranging from 500 mCi to 1 Ci (see Section 2.3). According to 

a DOE website, 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. Highlevel or transuranic waste is not generated or stored at SNL-CA (SNL 1992). Table 2-1 summarizes 

the SNL-CA buildings of potential radiological interest. 

LLNL performed external film dosimetry monitoring services for SNL-CA from 1956 to 1959 (SNL 

1958–1961). After that, a commercial vendor, Radiation Detection Company (RDC), provided film 

dosimetry services until the early 1970s (SNL 1958–1961, 1961–1962, 1962; DeSelm 1965; SNL 

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 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, Idaho (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 accredited under the DOE 

Laboratory Accreditation Program (DOELAP) (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 1992; Ward 1994). Most of the dosimetry information presented, described, and evaluated 

in this section for years after 1991 was obtained at SNL-NM by the Oak Ridge Associated Universities 

(ORAU) Team for the purpose of developing a site profile for that site (those records are directly 

applicable to the dosimetry program at SNL-CA from 1991 forward). Although the currently available 

records 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 that were used, 

exchange periods, and in most cases, types of radiation dose quantities that were measured and 

recorded.

6.3 DOSE RECONSTRUCTION PARAMETERS

6.3.1 Site Historical Administrative Practices

6.3.1.1 Administrative Practices Before 1989

Documents from early in the SNL-CA site history indicate a policy of maintaining permanent dosimetry 

records (SNL 1963). Between 1956 and 1959, most SNL-CA employees, contractors, and visitors 

were required to wear dosimeter badges (SNL 1958–1961, 1958–1978). During this time, LLNL was 

performing dosimetry services for SNL-CA, and records management was 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 about the need to 

badge all personnel on the site versus badging only personnel with access to “exclusion zones” 

(areas where radiation fields were present). Over time, 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 1958–1978). Through about 1972, dosimetry results were recorded manually on 4- by 6-in. 

cards or were stored as text file hardcopies, all of which were transferred to SNL-NM in 1993 (Wright 

1993). Electronic external dosimetry records for 1973 to 1988 were transferred to SNL-NM in 1989 

(Wright 1993) and apparently became 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 1994).
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6.3.1.2 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 Personnel Dosimetry Department 7715; onsite oversight and 

implementation was conducted by SNL-CA Personnel Dosimetry Department 8541 (Ward 1994). 

Dosimetry records from Department 7715 are entered into SANDOS. An example personnel dose 

history output (SNL 1993c) is shown in Attachment B (Figure B-1). Although dose units are not stated 

in SANDOS output, review of other documents (SNL 1993c) suggests that they are reported in rem. 

There are specific protocols and required training for dosimetry data entry and records management 

(Ward 1994).

Consistent with DOE Order 5480.11, Section 9.g (DOE 1988), 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; and

• 1.5 rem (0.015 sievert) annual dose equivalent to the lens of the eye.

Any site personnel with potential to receive annual doses exceeding these limits are required to wear 

personal dosimeters (Thompson 1991; Ward 1994). Monitored workers with interim dosimetry results 

indicating that annual doses could approach or exceed these limits can have work restrictions 

imposed. A 1997 memorandum announced an apparent change in dose limit reporting protocols 

under the SNL ALARA policy (Figure 6-1; SNL 1997). 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 (SNL 1997).

There are two dosimeter categories for monitoring personnel on the 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, 
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or personnel performing special radiation work outside the scope of their normal activities (i.e., 

nonroutine work tasks where the potential for additional or special types of exposures might exist).

While not made explicitly clear in Section 8.4.1 of the SNL External Dosimetry Program Manual (Ward 

1994), routine field dosimeter exchanges appear to occur on a monthly or quarterly basis while 

nonroutine exchanges occur on a biweekly basis (for special radiation work cases, exchange periods 

can 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 reread values for daily measurements of calibration 

cards. Environmental background consists of accumulated exposures from cosmic and terrestrial 

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 checkout 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 onsite use as well as for subsequent handling and shipment to the Personnel Dosimetry 

Division in the New Mexico office for processing. Personnel who are issued routine field dosimeters 

are required to have documented training on 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, and 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 1994).

Nonroutine field dosimeters for temporary employees and visitors are issued by the Security Patrol 

Division at specified site check-in locations. The Security Inspector is responsible for ensuring proper 

documentation of personnel being issued nonroutine field dosimeters and for instruction in their use 

while on the 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 

for routine field dosimeters.

Nonroutine field dosimeters for special radiation work cases are requested from Division 7715 and 

issued by department managers for personnel who are 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 unreturned field dosimeters result in a formal investigation to resolve the issue as well as to 

document and justify the dose equivalent assigned to the individual’s dosimetry record. Radiation 

work restrictions can be imposed pending resolution of an investigation. Example documentation of 

an unreturned dosimeter investigation and assignment of missed dose are provided in Attachment B 

(Figures B-4 to B-6). 

The overall performance of the SNL Personnel Radiation Dosimetry Program is monitored quarterly 

by a blind audit (Ward 1994), but who performs these audits is unclear.
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In addition to personnel radiation dosimetry, SNL-CA conducts a Special Dosimetry Program that 

includes work area monitoring (Ward 1992; SNL 1995). These dosimeters are also considered 

“nonroutine” and respective exchange periods can vary depending on the intended monitoring 

purpose. An example request form and 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 1995).

6.3.2 Site Dosimetry Technology

The historical progression of dosimeter technologies for external dose monitoring at SNL-CA is shown 

in Table 6-1. As broken down and discussed in the following sections, site dosimetry technologies 

have been partitioned into two historical timeframes (1956 to 1988 and 1989 to the present). This is 

because, for dose reconstruction, more complete and detailed information is available from after 1989 

to describe the dosimetry program after 1989.

Table 6-1. Documented dosimetry technologies and dose quantities (SNL 1958–1961, 1958–1978, 

1962, 1963, 1964; Wright 1993). 

Dosimeter type Years

Dosimetry 

service 

provider Measured quantities

Exchange 

frequency

Compliance dose 

quantities

Two-element 

beta/photon film + 

NTA neutron film

1956–1959 LLNL Photons, beta, 

neutron, penetrating, 

nonpenetrating

Monthly AEC Manual chapter 0524

Two-element 

beta/photon film + 

NTA neutron film

1959–1971 RDC Photons, beta, 

neutron, penetrating, 

nonpenetrating

Monthly AEC Manual chapter 0524

2-chip TLD (one 

filter type only)

1972–1982 RESL No discrimination of 

different radiations. 

Photons, beta, 

neutron, penetrating, 

nonpenetrating

Semiannual AEC Manual chapter 0524

3-element Eberline 

TLD 

1982–1988 RESL Photons, beta, 

neutron, penetrating, 

nonpenetrating

Annual or 

semiannual

ANSI N13/WD-2 (N324) 

1978 penetrating/ 

nonpenetrating, photons: 

<0.3 MeV, 0.3–10MeV, 

betas: 0.2–2.0MeV, 

photon/beta mixtures

Multielement 

Harshaw TLD 

(separate special 

dosimeter for 

neutrons)

1989–1990 SNL-NM Shallow (Sh), deep 

(Dp), neutron (Nt)

Quarterly Skin = Sh + Dp + Nt WB = 

Dp + Nt, extremities = Sh 

+ Dp + Nt 

Multielement 

Harshaw TLD

1991–

present

SNL-NM 

(DOELAP)

Shallow (Sh), deep 

(Dp), neutron (Nt)

Quarterly Skin = Sh + Dp + Nt WB = 

Dp + Nt, extremities = Sh 

+ Dp + Nt 

6.3.2.1 Dosimetry Technology, 1956 to 1988

As suggested by some of the documents Table 6-1 cites, dosimeters that were worn by SNL-CA 

personnel between 1956 and the early 1970s consisted of two-element DuPont Type 554 beta/photon 

film along with nuclear track emulsion, type A (NTA) film. Film badges consisted of four windows with 

one open and the others filtered with various densities of lead, cadmium, and aluminum (RDC 1963). 

There is a discrepancy in the available record as to exactly when SNL-CA discontinued using film 

dosimeters and began using TLDs. 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 
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that use. Lovell (1983) estimates the use of TLDs began in about 1968. RDC (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 TLDs 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 use 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.

As reported in Wright (1993), up until 1982 the SNL-CA TLD technology consisted of a two-chip 

badge with only one filter type [30 mg/cm2 total filtration according to Wright (1981b)] and as such, it 

could not discriminate between different radiation types or energies. However, the actual dosimetry 

records for this period report both penetrating and nonpenetrating doses including beta, photon, and 

neutron radiations. If the TLDs 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 could 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.

6.3.2.2 Dosimetry Technology, 1989 to Present

Since 1989, radiation dose monitoring at SNL-CA has been based primarily on the use of Harshaw 

TLDs. The Harshaw 8800 series TLD systems were the first to be DOELAP accredited at SNL and 

are the predominant systems used at all Sandia sites. Harshaw 4000 series TLDs are apparently 

used as well (Tucker 1977a,b, 1978; Kay 1979; Stanley 1987a,b,c; SNL 1996a), but little information 

is currently available on the specific application and extent of use of this series. There is some 

evidence that the 4000 series might 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 of its basic technologies. Model 8801 and 8802 TLD 

cards consist of four thermoluminescence elements 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 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, and field cards for worker monitoring have no 

added coloring).

The assembled dosimeter consists of the TLD card inside a Model 8812 cardholder. 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 1,400 cards can be loaded at a 

time and read automatically. Barcode identification information 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 of radiation they have received. The light is converted to an 

electrical signal by a photomultiplier tube. The relative strength of the electrical signal is measured in 
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Figure 6-2. Harshaw Model 8812 dosimeter assembly, element 

specifications, and respective radiation and dose equivalent quantities 

measured by Rhea and Bradley (1990).

units of charge (nanocoulombs) to create glow curves that are then analyzed against card reader 

calibration parameters. Noncontact heating of the cards with hot nitrogen gas improves dosimeter 

reuse, 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 on a desktop 

computer. Card reader results are converted into dose equivalents by a complex algorithm that was 

developed specifically for the Harshaw 8800/8812 system and SNL sites.

Along with explicit protocols for card reader calibration, accredited DOELAP programs involve an 

ongoing 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.

6.3.3 Calibration

6.3.3.1 Calibration for Dosimeter Technologies, 1965 to 1988

Calibrations for film dosimeters processed by RDC in the 1960s involved exposing film in SNL-CA 

badges to known 60Co gamma and X-ray fields (effective X-ray energies were 35 keV and 90 keV) 

based on previous calibration data for RDC badges (RDC 1963). SNL-CA badges had aluminum, 

cadmium, and lead filters, each of which was cross calibrated against similarly filtered RDC badges in 

parallel runs. Calibration of TLD badges SNL-CA used between 1971 and 1988, the period in which 

RESL performed dosimetry services, are provided as follows. 

• Cesium-137 was considered for a calibration source in 1959 (AEC 1960, p. 83) and was 

installed in the instrument calibration facility in 1961 (Horan 1962). An automatic badge 

irradiator developed in the 1960s did not use a phantom to provide backscatter

(Cipperley 1966). 

• As reported in 1981, an extrapolation chamber was built for the measurement of beta doses 

(Gupta 1981). The chamber window was polycarbonate, the gas was air, and the thick 

collecting electrode was Shonka tissue-equivalent plastic. The chamber was used to calibrate 

a 2.5-Ci 90Sr/Y source to tissue rad. The source, with an area of 2.5 cm2, was constructed by 

the Amersham Searle Corporation in February 1975. This source was used to measure beta 
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correction factors for several instruments after the Three Mile Island-2 reactor accident in 

1978. TLD badges were calibrated to 500 mrad tissue using a 1.78-cm-thick phantom 50 cm 

from the source (300 rad/hr).

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

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

• Use of a phantom in calibration apparently started in 1981 with the National Voluntary 

Laboratory Accreditation Program (NVLAP) certification process developed for non-DOE 

dosimetry processors. About this time, calibration developed in terms of absorbed dose to 

tissue rather than exposure. Beginning in January 1981, in response to a draft NVLAP (a 

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

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

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

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

Kalbeitzer 1984).

6.3.3.2 Calibration for Dosimeter Technologies, 1989 to Present

Harshaw Model 8800 TLD card readers are set up 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 1996b).

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 to be different for calibration of the 

Harshaw 4000 system (SNL 1996a), and an example calibration form is shown in Figure 6-4. No 

official manuals or procedural documents have been found about the Harshaw 4000 dosimetry 

system calibration or system applications, and no similar documentation has been found about the 

Harshaw/Bicron EXTRAD system.

6.4 DOSE CALCULATION AND REPORTING

6.4.1 Dose Calculation and Reporting, 1956 to 1988

As shown in Table 6-1, reported doses for SNL-CA employees, contractors, and visitors through 1988 

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 reporting categories by comparison against calibration curves that were developed for each 

batch of film. No information is currently available about MDLs, uncertainty or bias, or adjustments to 

recorded dose during this period of film dosimeter use.

As mentioned previously, the two-chip single-filter TLD dosimeters in use from about 1972 to 1982 

apparently could not directly measure each of the various dose quantities that were reported in the 

record for this period. No definitive information has been found about adjustments to recorded dose, 

and no information has been found about dose calculation algorithms, system MDLs, or uncertainty or 

bias during this period.

Like the earlier period in which two-chip TLDs were used at SNL-CA and processed by RESL in 

Idaho, there is no documentation about dose calculation algorithms, MDLs, adjustments to recorded 

dose, or uncertainty or bias for the three-element Eberline TLD badges that were used from 1982 to 

1988. 
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Figure 6-3. Example Harshaw 8800 card reader calibration procedures, forms, analysis, and output. 

6.4.2 Dose Calculation and Reporting, 1989 to Present

The algorithms SNL used for calculating dose equivalents are specific to each particular dosimeter 

system (Bradley et al. 1994). Both shallow (0.007 cm) and deep (1.0 cm) dose equivalents are 

measured, calculated, and reported. Although dose to the lens of the eye is not measured directly, it 
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Figure 6-4. Example calibration form for the Harshaw 

4000 dosimetry system. 

can be calculated indirectly from routine dosimetry results. Due to the complexity of this calculation, 

however, SNL does not routinely calculate dose to the 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 1993c). Calculation or database programming errors are also included but do not necessarily 

involve an adjustment. For example, in 1997 a SANDOS programming oversight was discovered that 

had resulted in incorrect background subtractions since 1992 and therefore in slightly overestimated 

doses (Walker 1997c). The programming correction was made when discovered, but no adjustments 

to the affected records were implemented because the slight overestimations were considered 

conservative.

The angular dependence of the Harshaw 8812 TLD was studied by SNL as part of the DOELAP 

accreditation process (Friedman, Loudermilk, and Thompson 1991). A batch of 8812 TLDs was sent 

to the Pacific Northwest Laboratory, which maintains a National Institute of Standards and 

Technology-traceable irradiation and measurement quality assurance program, to assess the angular 

dependence of 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 SNL uses are 

acceptable and “comparable to or better than results for modern dosimetry systems.”
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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 6-2. A 1990 LLD study 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. Because the shallow dose 

LLD for the newer 8802 cards is 10 mrem or less across all exposure categories, SNL uses a 

10-mrem LLD value in SANDOS for shallow dose equivalent. 
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Table 6-2. Monthly and quarterly LLDs for the Harshaw Model 8802 dosimeter card (adapted from 

Walker, Loudermilk, and Thompson 1997).

Exposure 

category Radiation types

Test 

depth

Monthly LLD 

(rem)

Quarterly LLD 

(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

IIIA General X-rays Deep 0.004 0.006

IV Cs-137 gammas Shallow 0.003 0.007

IV Cs-137 gammas Deep 0.003 0.004

VA Beta particles Shallow 0.003 0.007

VI Moderated 

neutrons

Deep 0.003 0.005

Exposure category VII (mixtures)

Component exposure 

categories

Test 

depth

Monthly LLD 

(rem)

Quarterly LLD 

(rem)

III + IV Shallow 0.005 0.010

III + IV Deep 0.004 0.006

III + VA Shallow 0.004 0.008

III + VA Deep 0.005 0.008

III + VI Deep 0.003 0.004

IV + VA Shallow 0.004 0.008

IV + VA Deep 0.004 0.006

IV + VI Deep 0.003 0.004

6.4.3 Exposure Energy Spectra

Dose reconstruction under the NIOSH program requires estimates of exposure percentages within 

specific energy bands for each type of radiation as follows (NIOSH 2007b):

• Photons:

– <30 keV, 

– 30 to 250 keV, and

– >250 keV. 

• Beta particles:

– >15 keV. 

• Neutrons:

– <10 keV,

– 10 to 100 keV,

– 100 keV to 2 MeV, 

– 2 to 20 MeV, and

– >20 MeV. 

For external exposures, betas with energies of less than 15 keV are not applicable. Table 6-3 

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. 
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Other sources and assumptions are provided as applicable in the table footnotes. The TRL is not 

listed because the primary radionuclide was tritium with beta energies of less than 15 keV.

Table 6-3. Selection of radiation energies and percentages. 

Weapons Laboratory Facility Complex (Buildings 910, 912, 913, 914, 916, 918), 1958 to 1998

Process description

Radiation type, 

energy selection (keV), and 

percentagea

Test/repair of neutron and X-ray detectors 

(neutron and X-ray generators)

Photon 30–250 100

Neutron 10–100 5 

 100–2,000 5 

2,000–20,000 90

Wet machining of DU Beta >15 100

Photon 30–250 90b

>250 10b

Radiography for weapons components Beta >15 100

Photon 30–250 50b,c

>250 50b,c

Radiography for materials science studies (X–ray 

diffraction operations later moved to Building 941)

Beta >15 100

Photon <30 40

 30–250 30

>250 30

H-3 storage studies Photon 30–250 50b,d

>250 50b,d

Ion beam analysis of materials Photon 30–250 30

>250 70

Radiflo leak tests Beta >15 100

Photon 30–250 10

>250 90

Radiography (Building 923), unknown to early 1990s

Process description

Radiation type, 

energy selection (keV), and 

percentagea

Radiography using X–rays, gamma rays, neutrons, 

alpha and beta particles

Beta >15 100

Photon <30 5 

 30–250 45

 >250 50

Neutron <10–100 5 

 100–2,000 70

>2,000–20,000 25
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Micro and Nano Technologies Laboratories (Buildings 941, 942, 943), unknown to present 

Process description

Radiation type, 

energy selection (keV), and 

percentagea

Radiography using X–rays, gamma rays, neutrons, 

alpha and beta particles

Beta >15 100

Photon <30 40

30–250 30

>250 30

Explosives and Environmental Testing Complex 

(Buildings 955, 956, 966, 972, 974, 976, 977, 978, 979, 981, 983), 1958 to present 

Process description

Radiation type, 

energy selection (keV), and 

percentagea

Environmental testing of mock up weapons and 

components (DU)

Beta >15 100

Photon 30–250 90b

>250 10b

Storage Facilities (Buildings 921, 927, 961, 982), unknown to present

Process description

Radiation type, 

energy selection (keV), and 

percentagea

Storage and packaging of waste materials Beta >15 100

Photon 30–250 70

>250 30

a. Estimated primarily by site personnel.

b. Based on default assumptions favorable to claimants for DU in ORAUT (2011). 

c. Assumes radiography primarily associated with weapons mock-ups and DU.

d. Assumes only external exposures would be associated with DU beds in tritium storage facility.

6.4.3.1 Neutron Dose Conversion Factors

As described in ORAUT-OTIB-0055, Technical Basis for Conversion from NCRP Report 38 Neutron 

Quality Factors to ICRP Publication 60 Radiation Weighting Factors for Respective IREP Input 

Neutron Energy Ranges (ORAUT 2006b), adjustments to neutron dose are necessary to account for 

changes in quality factors between historical and current scientific guidance. Using the method in 

ORAUT-OTIB-0055, adjustment factors were determined for the various energy groups at the two 

SNL-CA facilities where neutron exposures were possible. Table 6-4 shows multiplier values to use 

for dose reconstruction.

Table 6-4. Neutron DCFs for applicable SNL-CA facilities. 

Facility

Neutron energy 

intervals

NCRP 38

(NCRP 1971) 

quality factor

ICRP 60

(ICRP 1991) 

weighting factor

Dose 

fractiona

Corrected 

dose 

equivalent 

multiplierb

Weapons laboratory complex 10–100 keV 5.38 10 0.05 0.093

Weapons laboratory complex 0.1–2.0 MeV 10.49 20 0.05 0.095

Weapons laboratory complex 2.0–20.0 MeV 7.56 10 0.90 1.190

Radiography 10–100 keV 5.38 10 0.05 0.093

Radiography 0.1–2.0 MeV 10.49 20 0.70 1.335

Radiography 2.0–20.0 MeV 7.56 10 0.25 0.331

a. From Table 6-3.

b. Multiply reported dose by these factors to determine the corrected neutron dose equivalent for each applicable neutron 

energy interval.
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6.4.4 Missed Dose

Given the lack of documentation about MDLs for dosimeter systems for the period from 1956 to 1988, 

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 

in use at other AEC or DOE facilities during corresponding periods. Maximum annual missed doses 

were estimated using exchange periods and MDL/2 as recommended by NIOSH (NIOSH 2007b). For 

1989 to the present, the MDL information in the previous section was used to estimate missed dose 

using exchange periods and MDL/2. Table 6-5 lists documented or estimated MDLs and maximum 

missed doses for each dosimetry technology SNL-CA used.

Table 6-5. 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 180

1959–1971 Two-element beta/photon film 30b Monthly 180

1956–1971 Neutrons (NTA film) 50b Monthly 300

1972–1982 Two-chip TLD 20c Semiannual 20

1982–1988 Three-chip Eberline TLD 20c Annual; semiannual 10; 20

1972–1988 Neutrons (two- and three-chip TLD 

systems)d

20c Annual; semiannual 10; 20

1989–1990 Multielement Harshaw TLD 10 Quarterly 20

1991–present Multielement Harshaw TLD 10 Quarterly 20

1989–present Neutrons (Harshaw TLD systems) 5 Quarterly 10

a. Maximum annual missed dose calculated using N × MDL/2 from NIOSH (2007b).

b. Estimated MDL based on contemporary film systems of similar sensitivity (ORAUT 2007a).

c. Estimated MDL based on contemporary TLD systems of the period (ORAUT 2016, 2010b, 2006b).

d. Although neutron doses are reported for the two-chip system, no documentation of the neutron dosimetry methodology 

employed has been found.

6.4.5 Organ Dose Conversion Factors

During the period in which film dosimeters were used at SNL-CA, dosimeters were calibrated in units 

of roentgen. Limited information is available about 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 

(roentgen). 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 6-6 shows dose units to use for organ DCFs.

Table 6-6. Dose units for organ DCFs. 

Years

Photon dose units for 

use with organ DCFs

1956–1959 R

1959–1971 R

1972–1982 R

1982–1988 R

1989–1990 Hp(10)

1991–current Hp(10)

6.4.6 Adjustments to Recorded Neutron Dose

Based on multiplying the data from the following three sections, the total correction factor for NTA 

neutron doses is 2.65.
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6.4.6.1 Correction for Low-Energy Neutrons with NTA Neutron Dosimeters

Because NTA film is a poor detector of neutron energies below 500 to 800 keV (ORAUT 2014), a

correction factor of 1.5 is applied, based on information from ORAUT-TKBS-0036-6, Argonne National 

Laboratory-East – External Dosimetry (ORAUT 2014). 

6.4.6.2 Angular Dependence of NTA Neutron Dosimeters

The results of a study by Kathren, Prevo, and Block (1965) that indicated a correction of 1.3 is 

appropriate for NTA film for neutrons of energy greater than about 2 MeV. The data provided in the 

Kathren, Prevo, and Block study also indicated the dosimeters did not need correction at lower 

neutron energies. 

6.4.6.3 Track Fading of Neutron NTA Dosimeters

RDC participated in testing of their film dosimeters, and the results for neutron exposures averaged 

40% higher than the reported administered neutron exposures (RDC 1961, pp. 15–18). The

exposures and results were reported in terms of neutron flux. A report describing their film dosimeters 

included the following (RDC 1958–1961, p. 23):

The NTA emulsion is subject to latent image track fading. The longer the period between 

exposure and development, the more pronounced is the fading. Fading is greatest for 

short tracks and is accelerated by high humidity. Where humidities are moderate, biweekly or monthly periods do not produce serious latent image fading unless low neutron 

energies are encountered. It is recommended that films be returned for processing 

immediately after the exposure period, preferably via air mail. Films are developed on 

the day they are received and reports are issued within 48 hours.

The Mound Site determined that track fading was determined to be 9% per week. The NTA film was 

exchanged on a monthly cycle and therefore the correction for this effect is assumed to be 1.36

(Meyer 1994, Volume IV, pp, 171–177, 181–185).

6.4.7 Uncertainty

Given the lack of specific technical information available about dosimetry systems for much of the 

SNL-CA site history, it is necessary to estimate measurement uncertainty based on reported values 

for contemporary systems in use at other facilities. Table 6-7 shows estimates based on ORAUTTKBS-0012-6, Oak Ridge National Laboratory – Occupational External Dose (ORAUT 2007a). Some 

general analogies can be drawn between ORNL and SNL-CA in terms of dosimetry technologies in 

use during various periods of interest.

Table 6-7. Systematic uncertainty estimates (adapted from ORAUT 2007a). 

Dosimeter Period of use Systematica uncertainty factors

Two-element film 1956–1971 1.3

Two-chip TLD 1972–1982 1.15

Three-chip TLD 1982–1988 1.15

Multielement TLD 1989–present 1.15

a. Systematic uncertainty from lack of knowledge about energy distributions and geometries.

For neutrons, the biases are corrected as described in Section 6.4.6. The uncertainty factor is 1.3 [5]. 
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 70 of 107

7.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database (SRDB).

[1] Thomas, Elyse. ORAU Team. Principal Medical Dosimetry Scientist. 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 because 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.

Lumbar 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.

According to a 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.

[5] Rohrig, Norman D. ORAU Team. Health Physicist. April 2006.

This standard ORAU Team TBD table is based on information in Attachment A and an 

understanding of error propagation. 
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a Class of Employees from Sandia National Laboratories – Livermore, California, U.S. 

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Silver, W. J., C. L. Lindeken, J. W. Meadows, E. H. Willes, and D. R. McIntyre, 1976, Environmental 

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SNL (Sandia National Laboratories), 1961–1962, external dosimetry data (1961–1962), Livermore, 

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SNL (Sandia National Laboratories), 1962, dosimetry program suppliers (1962), Livermore, California. 

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SNL (Sandia National Laboratories), 1964, miscellaneous dosimetry program details, 1964, 

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SNL (Sandia National Laboratories), 1975–1977, sample urinalysis tests for tellurium and uranium 

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SNL (Sandia National Laboratories), 1980–1984, termination occupational exposure reports (1975–

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SNL (Sandia National Laboratories), 1983, Sandia National Laboratories Livermore, Environmental 

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SNL (Sandia National Laboratories), 1997, “Monthly ALARA Radiation Dosimetry Report,” 

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SNL (Sandia National Laboratories), undated a, Sandia National Laboratories Livermore Program to 

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Stanley, A. L., 1987a, “3318 Progress Summary, October - December, 1986,” memorandum to W. D. 

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Stanley, A. L., 1987b, “Progress Report, April - June 1987,” memorandum to W. D. Burnett, Sandia 

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Stanley, A. L., 1987c, “Third Quarter Progress Report, July-September, 1987,” memorandum to W. D. 

Burnett, Sandia National Laboratories, Albuquerque, New Mexico, October 27. [SRDB Ref ID: 

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Stanley, A. L., 1991, “Basis for Wearing Only One Dosimeter,” memorandum to N. R. Ortiz, Sandia 

National Laboratories, Albuquerque, New Mexico, April 22. [SRDB Ref ID: 18275]

Thompson, B. C., 1991, Sandia National Laboratories Radiation Protection Overview Program, 

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California), Date of Survey: August 1975, EGG-1183-1693, EG&G/Energy Measurements, 

Las Vegas, Nevada, October. [SRDB Ref ID: 34498]

Tucker, G. E. Jr., 1977a, “Division 3313 Progress Report, Second Quarter 1977,” memorandum to W. 

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Tucker, G. E. Jr., 1977b, “Division 3313 Progress Report, Third Quarter,” memorandum to W. H. 

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Tucker, G. E. Jr., 1978, “Division 3313 Progress Report - 1977 Fourth Quarter,” memorandum to W. 

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Toy, A. J. , C. L. Lindeken, K. S. Griggs, and R. W. Buddemeier, 1981, Environmental Monitoring at 

the Lawrence Livermore National Laboratory: 1980 Annual Report, UCRL-50027-80, 

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Ullrich, R. A., 2003, Cold War Context Statement, Sandia National Laboratories, California Site, 

SAND2003-0112, Sandia National Laboratories, Albuquerque, New Mexico, January. [SRDB 

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[SRDB Ref ID: 23577, p. 194]

Wall, W. R., 1981, Tritium Control and Accountability Instructions, SAND80-8057, Sandia National 

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Wright, D. A., 1998, untitled memorandum to D. Miller (U.S. Department of Energy, Albuquerque 

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memorandum to Distribution, Sandia National Laboratories, April 20. [SRDB Ref ID: 23669, 

p. 64] 
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GLOSSARY

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

becquerel (Bq)

International System unit of radioactivity equal to 1 disintegration per second; 1 curie equals 

37 billion (3.7 × 1010) Bq.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See film dosimeter and thermoluminescent dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

film

(1) In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. (2) X-ray film.

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 94 of 107

fission

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

gamma ray, particle, or photon (γ)

See gamma radiation.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

gamma radiation, neutron radiation, and X-ray radiation.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron.

nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation.
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 95 of 107

radioactive

Of, caused by, or exhibiting radioactivity.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. See radionuclide.

radionuclide

Radioactive nuclide. See radioactive and nuclide.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

thermoluminescence

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose. TLDs replaced film dosimeters at essentially all U.S. 

Department of Energy sites beginning in the 1960s.

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 96 of 107

 

ATTACHMENT A 

SUMMARY OF ACTIVITIES AT SITE BUILDINGS

Table A-1. Summary of activities by building.a

Building

number Building use

Radiological

activities

Dates of 

use Nuclides Comments

904 Auditorium None 1980s–

present

None None

905 Offices None 1980s–

present

None None

906 Combustion Laboratory 

Research

None 1980s–

present

None None

907 Equipment room None 1980s–

present

None None

910 Weapons Laboratory 

Facility

Neutron detection Unknown Unknown Used to design, 

assemble, test, 

calibrate, and repair 

neutron detectors

911 Administration None None None None

912 Administration, 

computers 

None 1958–

present

None None

913 Test assembly, machine 

shop, plating, 

metallography, tritium 

high-velocity air hoods

Tritium storage 

studies, U 

machining

1976–

1998

H-3, U-238 None

914 Test laboratories Radiflo, 

nondestructive 

testing, U 

machining 

Unknown Kr-85, beta 

sources, 

U-238

No radiological work 

for about 15 years

915 New building, office 

space

None None None None

916 Mostly light chemistry 

laboratories, previously 

a warehouse, Ar 

gloveboxes

E–microscopes, 

accelerators, 

various small 

sources, H-2 

storage tests

Unknown 700 keV, 

1 MeV, 

U-238

Sealed sources

917 Cylinder storage None None None None

918 Raw stock U-machining Unknown U-238 Deconstructed in the 

late 1980s

920 Office space None 1977–

present

None None

921 Decommissioned in 

1980

Radiological 

material storage 

and 

decontamination of 

Nevada Test Site 

test units

Unknown U-238, 

trace 

Pu-239, 

and mixed 

fission 

products

Became Building 9632 

in 1980 (Carter 2009)

922 Office space None None None None

923 Radiography Laboratory Radiography 

operations, X-ray 

machines 

Unknown Co-60

and Ir-192; 

X-ray 

machines, 

Cf-252 

Converted to records 

storage in early 1990s

924 Mo Mobile counting 

laboratory

Unknown Unknown Unknown Health Physics 

laboratory moved to 

Building 973

925 Health services None None None None
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 97 of 107

ATTACHMENT A

SUMMARY OF ACTIVITIES AT SITE BUILDINGS (continued)

Building

number Building use

Radiological

activities

Dates of 

use Nuclides Comments

927 Radiological Material 

Storage

Warehouse and 

vault

Unknown U-238, 

natural 

thorium

None

928 Shipping and receiving None None None None

929 Office space None None None None

940 Office space None None None None

941 Light mechanical, 

electrical, chemical 

laboratories

Radiography Unknown X-ray 

U-238, 

beta 

sources

Sealed sources

942 Light mechanical, 

electrical, chemical 

laboratories

None None None None

943 Plating operations None None None None

955 Environmental Test 

Facility

Environmental 

testing of various 

test units

Unknown U-238 

metal

None

956 Vibration Test Facility Environmental 

testing of various 

test units

Unknown U-238 

metal

None

960 Offices None None None None

961 Radiological and mixed 

waste storage and 

packaging

Storage and 

packaging

Unknown H-3, U-238 None

9611 Chemical waste storage None None None None

963 Maintenance Facilities None None None None

964 Security None None None None

966 High Pressure Gas 

Dynamics Test Facility

None None None None

967 Office None None None None

968 TRL Tritium research 1976–

1989

H-3, U-238 

beds

Decommissioned in 

1996, 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 None None None None

972 Centrifuge Test package 

centrifuge, 

penetrator studies

Unknown U-238 None

973 Environment, safety, 

and health laboratories

None None None None

974 Explosives test tanks Neutron generator 

tests

Unknown H-3 None

976 Four test cells for high 

press work

H-2 storage tests Unknown U-238 None

978 Flight test unit testing Mass properties 

tests

Unknown U-238 None

979 H-2 storage, research 

and development

Unknown Unknown U-238 Three Ar gloveboxes, 

two contaminated 

machines
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 98 of 107

ATTACHMENT A

SUMMARY OF ACTIVITIES AT SITE BUILDINGS (continued)

Building

number Building use

Radiological

activities

Dates of 

use Nuclides Comments

983 Test cells, flight test 

assembly

Unknown Unknown U-238 None

Explosives 

Storage Area

Magazines, explosives 

storage

Unknown Unknown H-3 None

a. Sources: DOE (1982); Wright (1981a, 2006).
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 99 of 107

 

ATTACHMENT B 

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

LIST OF FIGURES

FIGURE TITLE PAGE

B-1 Example external dosimetry history records (output from SANDOS) .....................................100

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

neutron exposures).................................................................................................................101

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

neutron exposures).................................................................................................................102

B-4 Example unreturned dosimeter investigation report ...............................................................103

B-5 Example unreturned dosimeter questionnaire follow-up letter................................................104

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

the Personnel Dosimetry Division...........................................................................................105

B-7 Work Area Radiation Dosimetry Request Form......................................................................106

B-8 Example work area radiation dosimetry results report............................................................107
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 100 of 107

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-1. Example external dosimetry history records (output from SANDOS).
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 101 of 107

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-2. Example nonroutine dosimeter evaluation form for special radiation work case (potential 

neutron exposures).
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 102 of 107

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-3. Example nonroutine dosimeter evaluation form for special radiation work case (potential 

neutron exposures).
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 103 of 107

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-4. Example unreturned dosimeter investigation report.
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 104 of 107

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-5. Example unreturned dosimeter questionnaire follow-up letter. 
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 105 of 107

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

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

by the Personnel Dosimetry Division.
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 106 of 107

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-7. Work Area Radiation Dosimetry Request Form.
Document No. ORAUT-TKBS-0053 Revision No. 02 Effective Date: 01/13/2021 Page 107 of 107

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-8. Example work area radiation dosimetry results report.