ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 100

DOE Review Release 05/07/2014

Summary Site Profile for Sandia National 

Laboratories in Livermore, California

ORAUT-TKBS-0053 Rev 01

Effective Date: 04/28/2014 

Supersedes: Revision 00

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

Document Owner 

Approval: Signature on File Approval Date: 04/14/2014

Lawrence A. Page, Jr., Document Owner

Concurrence: Signature on File Concurrence Date: 04/14/2014

John M. Byrne, Objective 1 Manager

Concurrence: Scott R. Siebert Signature on File for Concurrence Date: 04/14/2014

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 04/14/2014

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/28/2014

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 2 of 100

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 SEC-ER 

and HHS designation letter. Replaced reference to ORAUT-TKBS0048, 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. 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 3 of 100

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

1.1 Purpose...................................................................................................................... 10

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

1.3 Special Exposure Cohort............................................................................................ 10

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

2.1 Introduction................................................................................................................. 11

2.1.1 Purpose .......................................................................................................... 11

2.1.2 Scope ............................................................................................................. 11

2.1.3 Background..................................................................................................... 11

2.2 Site Activities .............................................................................................................. 13

2.2.1 Tritium Research Laboratory........................................................................... 16

2.2.2 Depleted Uranium Machining.......................................................................... 17

2.2.3 Radiography, Accelerators, and Neutron Generators...................................... 17

2.3 Site Processes ........................................................................................................... 18

2.3.1 Incidents ......................................................................................................... 18

2.3.2 Effluents.......................................................................................................... 19

2.4 Radiological Protection and Access Control ............................................................... 20

3.0 Occupational Medical Dose.................................................................................................... 21

3.1 Examination Frequency.............................................................................................. 21

3.2 Equipment and Techniques ........................................................................................ 21

3.3 Organ Dose Estimates................................................................................................ 22

3.3.1 Parameters and Estimation Method ................................................................ 22

3.3.2 Organ Dose Estimates.................................................................................... 24

3.4 Uncertainties .............................................................................................................. 25

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

4.1 Introduction................................................................................................................. 29

4.1.1 Purpose .......................................................................................................... 29

4.1.2 Scope ............................................................................................................. 29

4.2 Internal Dose from Onsite Atmospheric Radionuclide concentrations ........................ 29

4.2.1 Radionuclides of Concern ............................................................................... 29

4.2.2 Source Terms for Internal Dose ...................................................................... 31

4.2.3 Annual Intake of Radioactivity ......................................................................... 31

4.2.3.1 Inhalation Intakes ............................................................................. 32

4.2.3.2 Ingestion Intakes .............................................................................. 35

4.3 External Exposure to Onsite Atmospheric Radionuclide Concentrations .................... 36

4.3.1 Locations of Concern ...................................................................................... 36

4.3.2 Gamma-Emitting Radionuclides...................................................................... 36

4.3.3 Neutron Exposure........................................................................................... 39

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

5.1 Introduction................................................................................................................. 42

5.1.1 Purpose .......................................................................................................... 42

5.1.2 Scope ............................................................................................................. 42

5.1.3 Special Exposure Cohort ................................................................................ 42
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5.2 Summary of Potential Internal Radionuclide Exposures.............................................. 42

5.2.1 Tritium Research Laboratory........................................................................... 43

5.2.2 Uranium Alloy Machining ................................................................................ 43

5.3 Bioassay Programs .................................................................................................... 44

5.3.1 Tritium Bioassay Programs ............................................................................. 44

5.3.2 Uranium Bioassay Programs........................................................................... 44

5.4 Bioassay Laboratory Procedures and Minimum Detectable Intakes ........................... 45

5.4.1 Tritium............................................................................................................. 45

5.4.2 Uranium .......................................................................................................... 47

5.5 Bioassay Results........................................................................................................ 48

5.6 Method of Calculating Dose from Urine Bioassay Data............................................... 49

5.7 Airborne Radionuclide Concentrations........................................................................ 49

5.8 Unmonitored Dose Estimates ..................................................................................... 50

5.9 Radiation Dosimetry Reports...................................................................................... 50

5.10 Summary Tables ........................................................................................................ 51

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

6.1 Introduction................................................................................................................. 53

6.1.1 Purpose .......................................................................................................... 53

6.1.2 Scope ............................................................................................................. 53

6.1.3 Special Exposure Cohort ................................................................................ 53

6.3 Dose Reconstruction Parameters ............................................................................... 54

6.3.1 Site Historical Administrative Practices ........................................................... 54

6.3.1.1 Administrative Practices Before 1989 ............................................... 54

6.3.1.2 Administrative Practices After 1989 .................................................. 54

6.3.2 Site Dosimetry Technology ............................................................................. 57

6.3.2.1 Dosimetry Technology, 1956 to 1988 ............................................... 57

6.3.2.2 Dosimetry Technology, 1989 to Present ........................................... 58

6.3.3 Calibration....................................................................................................... 59

6.3.3.1 Calibration for Dosimeter Technologies, 1965 to 1988...................... 59

6.3.3.2 Calibration for Dosimeter Technologies, 1989 to Present ................. 59

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 ............................................................................... 62

6.4.3.1 Neutron Dose Conversion Factors.................................................... 63

6.4.4 Missed Dose ................................................................................................... 63

6.4.5 Organ Dose Conversion Factors..................................................................... 64

6.4.6 Uncertainty...................................................................................................... 64

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

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

Glossary ............................................................................................................................................ 86

ATTACHMENT A, SUMMARY OF ACTIVITIES AT SITE BUILDINGS .............................................. 89

ATTACHMENT B, DOSIMETRY PROGRAM ADMINISTRATIVE 

DOCUMENTATION EXAMPLES.........................................................................92
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LIST OF TABLES

TABLE TITLE PAGE

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

2-2 Historical T2 inventory at TRL................................................................................................. 16

2-3 Effluents of 3

H to air and sewer during operations and cleanup of TRL .................................. 19

3-1 Entrance kerma by procedure and period .............................................................................. 23

3-2 DCFs; absorbed dose for organs at various Al HVL for radiography ...................................... 24

3-3 Analogs for IREP organs not specified in ICRP (1982)........................................................... 24

3-4 Organ dose estimates for chest X-rays (rem)......................................................................... 25

3-5 Organ doses from lumbar spine X-ray, 1956 to 1971 ............................................................. 26

3-6 Distributions and corresponding statistical parameters for the dose to the Blymphocytes...........................................................................................................................28

4-1 Summary of methodology for estimating intakes of 3

H, 234U, and 239Pu................................... 32

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

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

4-4 External gamma radiation dose based on 2,500 hr/yr exposure duration ............................... 38

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

exposure duration ..................................................................................................................41

5-1 MDAs for and frequency of tritium bioassay ........................................................................... 46

5-2 Preliminary time line for uranium in urine analyses................................................................. 48

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

5-4 Estimated tritium doses based on bioassay, 1982 to 1995..................................................... 49

5-5 DAC Values for tritium and uranium ....................................................................................... 50

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

5-7 Detection limits for urine bioassay for 3

H with LSC ................................................................. 51

5-8 Detection limits for total urine bioassay .................................................................................. 52

5-9 Facility, source, solubility type, and particle size .................................................................... 52

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

6-2 Monthly and quarterly LLDs for the Harshaw Model 8802 dosimeter card.............................. 64

6-3 Selection of radiation energies and percentages.................................................................... 65

6-4 Neutron dose conversion factors for applicable SNL-CA facilities .......................................... 66

6-5 Estimated maximum annual missed photon, beta and neutron dose...................................... 66

6-6 Dose units for organ dose conversion factors......................................................................... 66

6-7 Systematic uncertainty estimates........................................................................................... 67
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LIST OF FIGURES

FIGURE TITLE PAGE

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

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

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

of tube potential .....................................................................................................................27

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

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

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

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

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

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

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

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

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

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

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

6-5 Angular dependence testing results for the Harshaw 8800/8812 dosimetry system. .............. 63

B-1 Example external dosimetry history records (output from SANDOS)...................................... 93

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

(potential neutron exposures)................................................................................................. 94

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

(potential neutron exposures)................................................................................................. 95

B-4 Example unreturned dosimeter investigation report................................................................ 96

B-5 Example unreturned dosimeter questionnaire follow-up letter ................................................ 97

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

the Personnel Dosimetry Division........................................................................................... 98

B-7 Work Area Radiation Dosimetry Request Form...................................................................... 99

B-8 Example work area radiation dosimetry results report .......................................................... 100
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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

ALARA as low as is reasonably achievable

AP anterior-posterior

Bq becquerel

CEDE committed effective dose equivalent

CEP Controls for Environmental Pollution

CFR Code of Federal Regulations

Ci curie

cm centimeter

cpm counts per minute

CRDL Chemical and Radiation Detection Laboratory

d day

DAC derived air concentration

DCF dose conversion factor

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

eq equivalent

ESE entrance skin exposure

ft feet

g gram

GPS Gas Purification System

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

LLD lower limit of detection

LLNL Lawrence Livermore National Laboratory

LSC liquid scintillation counting or counter
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 8 of 100

m meter

MBA Mass Balance Area

mCi millicurie

MDA minimum detectable activity

MDC minimum detectable concentration

MDL minimum detection limit

MeV megaelectron-volt, 1 million electron-volts

mL milliliter

mrem millirem

MPCa maximum permissible concentration in air

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A

ORAU Oak Ridge Associated Universities

PA posterior-anterior

pCi picocurie

POC probability of causation

RDC Radiation Detection Company

RESL Radiological and Environmental Sciences Laboratory

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

SPD site profile document

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

wk week

yr year

μCi microcurie

μg microgram

§ section or sections
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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 sites or categories of sites. They will be revised in the event 

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] guidelines established under 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010):

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

structures

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

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
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1.1 PURPOSE

This site profile document (SPD) for the Sandia National Laboratories site in Livermore, California 

(SNL-CA), describes aspects of the SNL-CA site and historical activities and practices pertinent to 

dose reconstruction under the 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 (HHS 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. After 1989, medical X-rays are not applicable because they were performed off site 

(NIOSH 2013). 
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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.

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 Lawrence Livermore National Laboratory (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.

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

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

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

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

The SNL-CA site presently consists of approximately 70 buildings 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, and central steam plant, office and laboratory 

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

of engineers and support staff. 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 3

H grew out of its 
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Figure 2-1. Boundaries of SNL-CA property (Larsen 2005). 

familiarity with, and use of, 3

H 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 was added as a buffer zone on the east side near the TRL in 1979. In 1986, an 

additional 228 acres was obtained, allowing an alternative exit route from the facility. Finally, in 1998, 

SNL-CA took part in a small land exchange to create a consistent buffer zone line along the western 

boundary, in which 2.82 acres were received in exchange for 5.41 acres. This brought the site’s area 

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

2005).
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Figure 2-2. Map of SNL-CA facilities (Ullrich 2003). 

2.2 SITE ACTIVITIES

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

SNL-CA over the years, some of which involved handling and release of radioactive materials. A 
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complete list of buildings along with known information about present and past uses and the presence 

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

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

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

activities that took place in these complexes. 

Table 2-1. Area information and parameters.

Area Weapons Laboratory Facility Complex

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

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, 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

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

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

H (when the TRL was 

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 

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

a shielded radiography cell (DOE 1992, Volume II).
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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 3

H-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 

(McDowell-Boyer 2006). There were some thorium metal parts used on test systems as well 

(McDowell-Boyer 2006). 

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

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

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

1992, Volume II). Details about total quantities of 3

H present for each year of operation of the TRL 

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

below.

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

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

not known. 

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

the major activities taking place on the SNL-CA site 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 such 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 DU-containing mock-ups of weapons and small, sealed radioactive sources not in use. 

Tritium was not stored in the 927 vault; rather the small quantities 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 3

H source 

from the vaults in Building 968 (the TRL), there were experiments conducted with up to 120 g 

(1.2 × 106 Ci) of 3

H per experiment in sealed gloveboxes. 

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

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

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 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 16 of 100

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 3

H are the only two potentially significant 

sources of internal exposure to workers based largely on the fact that these are the only two 

radioactive materials present in significant quantities throughout the history of SNL-CA. External 

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

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

aspects of the SNL-CA site pertinent to these sources of exposure.

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 3

H research and 

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

control personnel exposure to levels as low as 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 T2 (grams) T2 (Bq)

1979 1.06 3.8E+14

1980 15.47 5.5E+15

1981 29.62 1.1E+16

1982 28.55 1.0E+16

1983 27.62 9.8E+15

1984 37.86 1.3E+16

1985 85.49 3.0E+16

1986 107.58 3.8E+16

1987 174.35 6.2E+16

1988 181.59 6.4E+16

1989 148.89 5.3E+16

1990 115.46 4.1E+16

1991 117.85 4.2E+16

1992 131.67 4.7E+16

1993 101.21 3.6E+16

1994 29.43 1.0E+16

1995 9.96 3.5E+15

1996 4.04 1.4E+15

1997 0 0.0E+00

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

shop, a 3

H 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 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 17 of 100

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 3

H and its compounds, the fabrication of tritium compounds for use as 

engineering components (Wright 1981a), and examining the behavior of hydrogen isotopes and 

helium in metals to understand transport and structural properties (Garcia and Gorman 1996). 

Operations were performed inside sealed gloveboxes for any experiments involving more than 0.1 g 

of 3

H (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 

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 of 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 (McDowell-Boyer 2006). The 1989 Safe Operating Procedure (SOP) for machining of 

uranium metal (SNL 1989) dictates 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). 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 18 of 100

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 2.4.1) was reported to have 

occurred in Building 913, Room 113 in 1979 (Lovell 1980). Radiography is currently conducted in 

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

Two small accelerators have been and remain 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; McDowell-Boyer 2006). 

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 3

H 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

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

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

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

) of the chest and arm (Lovell 1980).

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 Laboratory 115A, an environmental release of 

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

1,650 mrem down to 2 mrem.

Tritium Release, January 1986

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

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

exposures.

Personnel Exposure and Tritium Release, August 1987

An 1,100 Ci environmental release occurred when an operator disassembled a vessel outside a 

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

and the operator received a dose of 15 mrem.

Tritium Release, October 1988

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

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

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

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

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

doses came from both cleanup and pump replacement operations.

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 are 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 2.4.1 

above.

Table 2-3. Effluents of 3

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

Year

Total tritium in stack 

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

Total tritium (as HTO) in 

wastewater discharges to 

sewer (Bq)

1979 2.2E+11 Not estimated 1.5E+08

1980 9.3E+11 Not estimated 1.5E+08

1981 1.6E+12 Not estimated 5.7E+09

1982 7.5E+12 Not estimated 1.1E+10

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

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

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

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

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

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

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

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

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

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

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

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

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

1996 2.9E+09 Not estimated 0.0E+00
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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; Nestor 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).

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 pre-1989 records

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 2.3.1.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 21 of 100

3.0 OCCUPATIONAL MEDICAL DOSE

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

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

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

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

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

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

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

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

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

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

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

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

documents uncertainties.

3.1 EXAMINATION FREQUENCY 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

found.

3.2 EQUIPMENT AND TECHNIQUES

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

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

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

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

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

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

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

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

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

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

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

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

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

administered at SNL/CA from 1956 through 1969, 1970 to 1985, and post-1985 using calculated site-
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 22 of 100

specific estimates with input from Technical Information Bulletin [TIB]: Dose Reconstruction from 

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

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

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

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

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

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

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

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

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

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

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

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

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

pre-employment X-rays.

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

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

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

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

be single phase [3].

3.3 ORGAN DOSE ESTIMATES

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

estimate these doses and Section 3.3.2 discusses results.

3.3.1 Parameters and Estimation Method

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

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

• Total filtration (millimeters of aluminum, mmAl)

• Estimate of tissue thickness (AP and LAT)

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

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

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

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

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

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

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

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

(ORAUT 2005a).

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

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

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

1). The LAT chest X-ray entrance kerma is 2.5 times the PA chest entrance kerma (ORAUT 2005a).
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 23 of 100

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

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

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

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

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

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

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

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

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

Period

PA chest 

entrance kerma 

(cGy) 

LAT chest 

entrance kerma, 

(cGY)

AP lumbar spine 

entrance kerma 

(cGY)

LAT lumbar spine

entrance kerma

(cGy)

Pre-1970 0.013a 0.0325a 0.85b 3.9c

1970–1985a 0.013 0.0325 NA NA

Post-1985d 0.043 0.1075 NA NA

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

thoracic cavity, such as the esophagus and bone surface. For abdominal organs (bladder, colon), the 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 24 of 100

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

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

Organ

Lumbar spine DCFs Chest DCFs

Pre-1970 

(2.5-mmAl HVL)

Pre-1970 

(2.5-mmAl HVL)

Pre-1970 

(2.5-mm Al HVL)

1970–1985

(2.5-mm Al HVL)

Post-1985

(4.0-mm Al HVL)

LAT AP LAT PA LAT PA LAT PA

Thyroid 0.01 0.3 137 174b 115 32 164 78

Eye/brain 0.01 0.3 137 32 115 32 164 78

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

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

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

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

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

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

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

Thymusd 14 79 220 451 220 451 351 674

Esophagusd 14 79 220 451 220 451 351 674

Stomachd 14 79 220 451 220 451 351 674

Bone surfacesd 14 79 220 451 220 451 351 674

Remainderd 14 79 220 451 220 451 351 674

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

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

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

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

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

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

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

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

reported in Lincoln and Gupton (1958).

d. Using analogs listed in Table 3-4.

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

examinations were used instead.

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

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

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

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

Anatomical 

location

ICRP #34 

reference organ IREP organ analogsa

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

gall bladder, remainder organs

Abdominal cavity Ovaries Urinary bladder, colon/rectum

Head and neck Thyroid Eye/brain

a. ORAUT 2005a

3.3.2 Organ Dose Estimates

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

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

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

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

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

1971. These X-rays were often taken at hire.

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 2012d), site-specific information, and 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 25 of 100

International Commission on Radiological Protection (ICRP) Publication 34 dose conversion factors 

(DCFs) (ICRP 1982). The dose distributions and corresponding statistical parameters for the dose to 

the B-lymphocytes for each projection and period is listed in Table 3-6.

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

Organ

Pre-1970

estimated dosea,b,c

HVL = 2.5 mm Al 

(uncollimated)

1970–1985

estimated dosea,b,c

HVL = 2.5 mm Al 

(collimated)

Post-1985

estimated dosea,b

HVL = 4.0 mm Al 

(collimated)

LAT PA LAT PA LAT PA

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

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

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

Liver/gall bladder/

spleen

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

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

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

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

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

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

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

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

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

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

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

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

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

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

a. SID = 183 cm.

b. Image receptor size 35.6 cm by 43.2 cm.

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

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

consistent with ORAUT (2005a).

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

 consistent with ORAUT (2005a).

3.4 UNCERTAINTIES

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

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

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

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

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

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

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

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

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

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

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

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

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

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

the medical community at the time. The following estimates of uncertainty associated 
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 Table 3-5. Organ doses from lumbar spine X-ray, 1956 

 to 1971. 

Organ

Estimated dosea,b

HVL = 2.5 mm Al (collimated)

LAT (rem) AP (rem)

Thyroid 3.88E-05 2.55E-04

Eye/brain 3.88E-05 2.55E-04

Ovaries 1.52E+00c 1.12E+00c

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

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

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

Testes 1.12E+01c 5.40E-02c

Lungs 5.43E-02 6.72E-02

Thymus 5.43E-02 6.72E-02

Esophagus 5.43E-02 6.72E-02

Stomach 5.43E-02 6.72E-02

Bone surfaces 5.43E-02 6.72E-02

Remainder 5.43E-02 6.72E-02

Female breast 5.04E-02 2.13E-02

Uterus 1.20E-01 2.44E-01

Bone marrow 8.54E-02 3.15E-02

Skind 5.39E+00 1.18E+00

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

b. Image receptor size 35.6 cm by 43.2 cm.

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

lumbar spine reflect actual measurements reported in Lincoln and 

Gupton (1958). 

d. Skin dose values include backscatter factors of 1.39 from 

Table B.8 of NCRP 102 (1989).

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

ORAUT (2005a) assume uncertainty in beam output due to timers to be +25%. 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 27 of 100

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

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

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

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

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

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

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

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

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

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

the default for the pre-employment X-rays. 

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

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

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

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

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

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

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

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

30% uncertainty is favorable to claimants.

 Figure 3-1. Kerma in air at 1 m from

X-ray source as a function of total 

 filtration for various values of tube

 potential (ICRP 1982).
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 28 of 100

Table 3-6. Distributions and corresponding statistical parameters for the dose to the B-lymphocytes.

Projection and Period Distribution Parameter 1 Parameter 2 Parameter 3

PA chest 1956–1969 Weibull3 2.860839 0.003534 6.47420E-05

LAT chest 1956–1969 Weibull3 2.679824 0.003979 1.59961E-05

PA chest 1970–1985 Weibull3 2.042838 0.002474 5.65475E-06

LAT chest 1970–1985 Weibull3 2.063026 0.003033 1.81402E-06

PA chest 1986–1989 Weibull3 2.113046 0.012645 2.64696E-05

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

AP Lumbar Spine 1956 –1971 Weibull3 3.278461 0.163130 ̶ 7.85764E-04

LAT Lumbar Spine1956 –1971 Normal 0.155303 0.048359 ---
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 29 of 100

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 on the 

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. The term occupational environmental dose refers to 

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

airborne radionuclides or ionizing radiation. 

4.1.2 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; Althouse et al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 

2004; Petersen 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 the SNL-CA. 

Section 4.4 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 2.4.2). Releases of DU from the SNL-CA site are not considered a significant 

source of environmental exposure at the SNL-CA. Such releases would only occur during machining 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 30 of 100

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). 

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

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

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

the SNL-CA site. In addition to 3

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

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

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

in ORAUT-TKBS-0035-4, Lawrence Livermore National Laboratory – Occupational Environmental 

Dose (ORAUT 2010b), 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 

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

; the maximum 

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

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

perimeter were not less than average measured 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

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

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

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

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

selected based on their relatively higher organ dose factors 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 contribute less 

than 5% of the committed dose, but always contribute less than 1 mrem to the yearly committed organ 

dose associated with environmental exposures for this 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, 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; Althouse et 

al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 2004; Petersen 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

, 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 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 31 of 100

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; Althouse et al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 2004; Petersen et al. 2005). 

The alpha-emitting radionuclides 234U and 239Pu (represented as gross alpha for pre-1971) from LLNL 

are retained as potentially significant contributors to internal environmental dose at SNL-CA. 

Inhalation intakes for these radionuclides are addressed in Section 4.2.3 below.

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

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

described in Section 4.2.3. 

4.2.2 Source Terms for Internal Dose

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

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

activities at LLNL 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 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 and of airborne 3H are estimated in this 

section. Ingestion of drinking water contaminated by 3

H is also considered. 

To calculate inhalation intake, it was necessary to consider the onsite air concentrations of 234U, 239Pu, 

and 3

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

Between 1979 and 1996, it was also necessary to consider onsite 3

H concentrations from TRL 

emissions.

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

2,400 m3

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

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

available; 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. Tritium should be assumed to be 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 32 of 100

associated with HTO because this will give the highest dose associated with the two forms (T2 and 

HTO) known to be released from both LLNL and SNL-CA (Garcia and Gorman 1996; Peterson 2005).

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. 

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

H, 234U, and 239Pu. 

Radionuclide Applicable period Method used

3

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

ratio of south perimeter H-3 concentration to release, 

derived from 1973 to 2003 data, to obtain estimate of 

maximum SNL-CA concentration (Bq/m3

). Multiply 

maximum concentration by 2,400 m3

/yr inhalation intake 

rate and by 1.5 to account for skin absorption.

1973–1978 and 

1997–2004

Reported south LLNL perimeter H-3 concentrations are 

used directly to estimate maximum SNL-CA concentration 

(Bq/m3

). Multiply maximum concentration by 2,400 m3/yr 

inhalation intake rate and by 1.5 to account for skin 

absorption.

1979–1996 Reported south LLNL perimeter H-3 concentrations are 

used directly to estimate LLNL contribution to maximum 

SNL-CA concentration (Bq/m3

). TRL contribution (onsite) 

derived by multiplying reported TRL releases by the 

maximum calculated ratio of near-TRL air concentrations 

to TRL releases for 1994 and 1995. Multiply both 

perimeter and near-TRL concentrations by 2,400 m3

/yr 

inhalation intake rate, and by 1.5 to account for skin 

absorption. Add intakes at LLNL perimeter to intakes near 

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

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

1961–1970 Reported south LLNL perimeter gross alpha concentration 

measurements were corrected for average Livermore 

valley background contributions and multiplied by an 

inhalation intake rate of 2,400 m3

/yr to estimate maximum 

SNL-CA intakes (Bq/yr).

1971–2004 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 234U concentrations were 

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

Activity intakes for all radionuclides are added and 

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

1961–1970 Reported south LLNL perimeter gross alpha concentration 

measurements were corrected for average Livermore 

valley background contributions and multiplied by an 

inhalation intake rate of 2,400 m3

/yr to estimate maximum 

SNL-CA intakes (Bq/yr).

1971–2004 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).
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 33 of 100

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

(Bq/yr), 1956 to 1970.

a 

Year H-3b,c U-234 or Pu-239d

1956 1.2E+04 (e)

1957 4.0E+04 (e)

1958 2.1E+04 (e)

1959 1.2E+04 (e)

1960 1.1E+04 (e)

1961 3.9E+03 1.2E-01

1962 1.5E+04 1.4E-01

1963 2.1E+04 2.7E-02

1964 9.3E+04 6.2E-02

1965 1.3E+06 0.00f

1966 6.0E+04 0.00f

1967 2.3E+04 0.00f

1968 2.4E+04 0.00f

1969 2.6E+04 8.9E-03

1970 1.0E+06 0.00f

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

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

estimated by multiplying the inhalation intake by a factor of 1.5.

c. Intake of H-3 is based on historical LLNL source terms from Peterson (2005); 

see Section 4.2.3.1 and Table 4-1.

d. 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.

e. No measurements reported for these years.

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

the offsite background gross alpha concentration.

3

H Intakes

Measured concentrations of 3

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

reports from 1973 to 2004 (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; Althouse et 

al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 2004; Petersen et al. 2005). These values were 

used to estimate environmental intakes of 3

H for SNL-CA workers during these years without 

considering additional dilution in transport to the SNL-CA site, which provides an estimate that is 

favorable to the claimant. Two LLNL annual environmental reports were not found (for 1982 and 

1984); for these 2 years, the 3

H concentrations were estimated to be the average of the preceding and 

following year’s concentrations.

Before 1973, 3

H concentrations at the LLNL perimeter were not reported in annual reports. For these 

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

concentration of 3

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

2.7 × 10-14) multiplied by the reported 3

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

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

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

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

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

and are given in Table 4-3. 

Between 1979 and 1996, the environmental 3

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

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

H 

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

SNL-CA monitors to detect environmental levels of 3

H at that time (Brekke and Holland 1994). In 
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Table 4-3. Maximum site-wide annual median inhalation intakes (Bq/yr), 1971 to 2004.

a 

Year H-3b,c U-234c Pu-239c

1971 4.9E+03d 3.5E-03 6.3E-03

1972 4.8E+03d 6.0E-03 2.8E-03

1973 8.3E+03 5.5E-03 1.5E-03

1974 1.1E+04 4.8E-03 2.9E-03

1975 1.2E+04 4.6E-03 2.0E-03

1976 1.9E+04 5.4E-03 8.0E-04

1977 1.6E+04 7.3E-03 2.1E-03

1978 1.7E+04 9.8E-03 2.7E-03

1979 8.7E+03e 5.2E-03 1.5E-03

1980 9.6E+03e 3.6E-03 4.4E-04

1981 1.4E+04e 7.1E-03 1.4E-03

1982 2.6E+04e 8.0E-03 1.4E-03

1983 1.5E+04e 8.9E-03 2.1E-03

1984 2.0E+04e 7.1E-03 1.2E-03

1985 5.2E+04e 5.3E-03 1.8E-04

1986 7.2E+04e 4.7E-03 1.2E-04

1987 1.8E+05e 5.6E-03 5.3E-05

1988 1.5E+05e 1.3E-02 8.0E-05

1989 8.2E+04e 9.4E-03 1.0E-04

1990 2.9E+04e 5.7E-03 1.2E-04

1991 4.5E+04e 5.9E-03 1.1E-04

1992 2.7E+04e 5.1E-03 8.9E-05

1993 1.8E+04e 4.2E-03 8.6E-05

1994 9.0E+03 3.3E-03 8.2E-05

1995 2.8E+03 2.7E-03 5.9E-05

1996 5.8E+02 2.7E-03 5.8E-05

1997 4.7E+02 2.9E-03 1.6E-05

1998 3.3E+02 2.0E-03 1.4E-05

1999 2.3E+02 3.9E-03 1.3E-05

2000 1.5E+02 0.0E+00f 2.2E-05

2001 1.5E+02 0.0E+00f 9.2E-06

2002 1.7E+02 6.9E-04 7.6E-06

2003 2.0E+02 1.4E-03 6.0E-06

2004 9.2E+01 1.3E-03 7.4E-06

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

b. Assumes total intake is the sum of inhalation and skin absorption, which is estimated by multiplying the 

inhalation intake by a factor of 1.5.

c. Intakes based on LLNL south perimeter concentrations; see Table 4-1.

d. Intake of 3

H is based on historical LLNL source terms from Peterson (2005); see Table 4-1.

e. Intakes include both LLNL south perimeter concentrations and TRL-generated concentrations; see 

Table 4-1.

f. Zero values indicate the analytical background exceeded the measured concentration of uranium 

isotopes.

1994 and 1995, 3H releases from the TRL were between 70% and 80% of the LLNL releases. 

Therefore, although most of the measured 3

H near the TRL was likely to be due to TRL releases 

during these 2 years, some of it would have been due to LLNL releases. Further, the ratio of the 

measured 3

H concentration in air near the TRL to the 3H release rate from the TRL in 1994 and 1995 

is an overestimate of the true ratio. Despite this, the maximum derived ratio for these 2 years was 

6.9 × 10-13, and this multiplier was used to estimate air concentrations near the TRL for the years for 

which measurements were not available for this location (1979 to 1993).

To consider both LLNL sources and the TRL, the intakes estimated for the south LLNL perimeter were 

added to the intakes estimated near the TRL. Due to the varying relationship between the amount of 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 35 of 100

LLNL 3H releases and the TRL releases, it was not possible to assume that the method of estimating 

concentrations near the TRL adequately considered LLNL contributions, especially for years when 

LLNL 3

H releases were 10 to 100 times greater than the TRL releases. Therefore, intakes due to the 

LLNL south perimeter concentrations and the intakes due to the onsite concentrations near the TRL 

were added.

Particulate Intakes (239Pu and 234U)

Measured concentrations of 239Pu, 235U, and 238U at the south perimeter of the LLNL, near the SNL-CA 

site, were reported in annual environmental reports from 1971 through 2004 (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; Althouse et al. 2001; Gallegos 

et al. 2002; Sanchez et al. 2003, 2004; Petersen 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 4-3). 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.

4.2.3.2 Ingestion Intakes

Ingestion of 3

H from drinking water available to workers was evaluated by considering the maximum 

sitewide intakes estimated for LLNL workers in ORAUT (2010b). It was noted in ORAUT (2010b) that 

several sources of drinking water for the LLNL site were sampled, along with water from the LLNL 

onsite swimming pool, which is close to the main sources of tritium at LLNL. The median activity in 

the drinking water sources was below the limits of detection. However, concentrations in the pool 

were reported to range from 0.8 to 200 Bq/L. Therefore, the pool water was adopted as the sitewide 

maximum source of drinking water for LLNL workers (ORAUT 2010b). There is no reason to believe 

that SNL-CA workers would be exposed to drinking water at a higher concentration than the pool 

water at LLNL, so it was assumed here that the maximum concentration of 3

H in drinking water at any 

time might be represented by 200 Bq/L. This concentration would lead to an annual ingestion dose of 

less than 1 mrem/yr, regardless of whether the assumed form of 3

H is tritiated water or organically 

bound 3

H. This calculation assumes a 3 L/d ingestion rate and a 250-day work year (i.e., 750 L/yr 

ingestion of drinking water). Based on this calculation, ingestion of 3

H in drinking water is not a 

significant route of worker exposure.
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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) 

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 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 data preference 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 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 37 of 100

Figure 4-1. Thermoluminescent dosimeter locations since 1989.

Figure 4-2. Location of LLNL perimeter gamma and neutron 

dosimeters (Griggs and Buddemeier 1986). 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 38 of 100

Figure 4-3. Location of LLNL offsite gamma 

dosimeters (Holland and Brekke 1988). 

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

Year

Average 

perimetera Errorb 

Maximum dose 

ratec

1956 120d 36 156

1957 120d 36 156

1958 120d 36 156

1959 120d 36 156

1960 120d 36 156

1961 120d 36 156

1962 120d 36 156

1963 120d 36 156

1964 120d 36 156

1965 120d 36 156

1966 120d 36 156

1967 11 3 15

1968 120 36 156

1969 24 7 31

1970 40 12 52

1971 25 7 32

1972 35 10 45

1973 27 8 36

1974 39 12 51

1975 90 27 118

1976 84 25 109

1977 65 19 84

1978 27 8 35

1979 27 8 36

1980 23 7 30

1981 17 5 23

1982 19e 6 25

1983 21 6 27

1984 22e 7 28

1985 23 7 30

1986 17 5 23

1987 17 5 22

Year

Average 

perimetera Errorb 

Maximum dose 

ratec

1988 16 5 21

1989 15 5 20

1990 15 4 19

1991 16 5 20

1992 16 5 20

1993 15 5 20

1994 16 5 21

1995 16f 5 20

1996 16f 5 20

1997 16f 5 20

1998 16f 5 20

1999 16f 5 20

2000 16f 5 20

2001 16f 5 20

2002 16f 5 20

2003 16f 5 20

2004 16f 5 20

a. Until 1990, unless otherwise noted, values are the 

south perimeter LLNL dose rate; after 1989, values 

are the average of SNL-CA perimeter dosimeters.

b. Error assumed to be ± 30% of the higher of the 

average perimeter (ORAUT 2012b).

c. Maximum dose rate is the average perimeter value 

plus error term.

d. Assumed maximum dose rate estimated for the 

average perimeter – no measurements reported for 

the LLNL south perimeter or for SNL-CA from 1956 

through 1966.

e. Assumed the average of the year prior and year 

following – missing annual environmental reports for 

1982 and 1984.

f. Assumed average of SNL-CA reported average 

perimeter values for 1990 to 1994.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 39 of 100

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 

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; 

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

2000; Althouse et al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 2004; Petersen et al. 2005). 

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 

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; 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; Althouse et al. 2001; Gallegos et al. 2002; Sanchez et al. 2003, 2004; Petersen et al. 2005). 

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 background 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.

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, 

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).
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 40 of 100

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).

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).
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 41 of 100

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

hr/yr exposure duration (mrem/yr).

a

Year

North perimeter 

doseb Errorc Maximum dosed

1966 18e 5 23

1967 18e 5 23

1968 18e 5 23

1969 18e 5 23

1970 18e 5 23

1971 18e 5 23

1972 18e 5 23

1973 18 5 23

1974 26 8 34

1975 50 15 65

1976 43 13 55

1977 39 12 51

1978 9 3 12

1979 13 4 16

1980 6 2 8

1981 2 1 3

1982 8f 2 10

1983 8 2 10

1984 8f 2 10

1985 <1 <0.3 <1

1986 2 1 3

1987 <1 <0.3 <1

a. Reported for years neutron dose rate was believed, or unknown, to have exceeded background levels.

b. Value for south LLNL perimeter dose rate at Location 3 or 5, 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. Maximum dose rate is the average perimeter value plus error term.

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

f. Assumed the 1983 value; annual environmental reports for 1982 and 1984 not found. 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 42 of 100

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, 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 SPD 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, 

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 

Tritium Research Laboratory (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. 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 43 of 100

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 

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.
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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 3

H at all SNL facilities has been urine bioassay (Potter ca. 

1997). There is no evidence that urine bioassay samples were analyzed for any radionuclides except 

3

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

employed at SNL-CA. The 3

H 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, 

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 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 45 of 100

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 4 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 on and off the air samplers. 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 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). 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 activities (MDAs) for tritium bioassay from the available 

quarterly Health Physics Reports and Standard Operating Procedures are given in Table 5-1.

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 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 46 of 100

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 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

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 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 Wright (1998) memorandum to Debbie Miller DOE/AL

Special bioassay

calculation forms. The background count in the tritium channel was reported by Hafner to be 

approximately 22 cpm. Calibration forms for 1995 give background counts in the tritium channel as 

approximately 15 cpm with counting efficiencies of approximately 0.67. Daily performance checks 

from 1993 to 1995 showed efficiencies in the range of 0.65 and background counts from 17 to 

21 cpm. Based on these parameter values, the minimum detectable concentration (MDC) for tritium 

in urine would be approximately 13 nCi/L.

Vosburg (1993) indicates:

(4.66 2.71) MDA Sb

EARV

+ = (5-1)

where:

Sb = blank standard deviation

E = efficiency

A = decays per disintegration = 1 for tritium

R = chemical recovery (not applicable for LSC for tritium)

V = volume of the sample = 0.5 mL
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 47 of 100

Assuming an MDA of 13.5 dpm and a 0.5-mL aliquot, the MDC would be 29,000 dpm/L or 13 nCi/L. 

This is consistent with the MDA of 10 nCi/L that 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 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 DR 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 detectable limit (MDL) by the fluorimetric 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 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 48 of 100

Table 5-2. Preliminary time line uranium in urine analyses. 

Period Laboratory Method MDA Basis and comments

<1968 No information No information No information No information

1968–

unknown

Radiation Detection 

Company (RDC)

Fluorimetry 5 μg/L Reported error (±2σ)

1973–1986 SNL-NM Industrial 

Hygiene Services

Fluorimetry

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 

inductively coupled plasma mass 

spectrometry. 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 MDC given Audit memos show that CEP 

analyses were problematic and 

therefore cannot be used.

1993–

present

SNL-NM Industrial 

Hygiene Services

Not defined but 

probably ICPMS

0.1 μg/L Reasonable MDAs for ICP-MS

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 a 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.

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 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 

3

H 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. 
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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 

Table 5-4. Estimated tritium doses based on bioassay, 1982 to 1995 (Garcia and Gorman 1996). 

Year

Maximum dose

(mrem/yr)

Average dose

(mrem/yr)

Total person-dose

(mrem/yr)

No. of monitored 

Individuals (inferred)

1982 70 30 183 6

1983 79 49 148 3

1984 1,620 152 3,040 20

1984a 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, 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, the stack, and 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 (MPCa) and the more recent derived air concentration 
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(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 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 air are given in Table 5-5. 

Table 5-5. DAC Values for tritium and uranium. 

Date

DAC (μCi/mL)

HTO U-238 Source

Type F Type M Type S

Post-1994 2 E-5 5 E−10 3 E−10 8 E−11 10 CFR 835, Appendix A

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 2000 h/yr, would be 4.8 × 103 μCi. After 

an adjustment of 1.5 for skin absorption, the intake would become 7.2 X103 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 

1.2 X 10-1 μCi for Type F uranium, 7.2 X10-2 μCi for Type M uranium, and 1.92 X10-2 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 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 51 of 100

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 

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 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. 

Period MDA (pCi/L)

1977–1983 10,000 (reported)

1980–1985 20,000 (calculated for 1980 based on background and efficiency for the LSC)

1986 20,000 (calculated)

1986–1997 10,000 (estimated based on previous reported MDAs)

1998–present 5,500

Note: Reporting limits were based on dose rather than bioassay results.
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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 memos show that CEP analyses were problematic and 

therefore 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 2010c) and the Hanford site (ORAUT 2012c).

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 (2007d).
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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 SPD 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). Dose reconstructions for individuals employed at SNL-CA during the period from

October 1, 1957, through December 31, 1994, but who do not qualify for inclusion in the SEC, may be 

performed using these data as appropriate. Therefore, dose reconstructions for individuals employed 

at SNL-CA from October 1, 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.

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 2.3, 

the laboratory typically handled kilogram amounts of DU, typically in the form of alloyed metal 

components, gram amounts of 3

H, 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 

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.
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LLNL performed external film dosimetry monitoring services for SNL-CA from 1956 to 1959 (SNL 

1958–1961). After that, a commercial vendor called 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 an SPD 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).

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 
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history output (SNL 1995a) is shown in Attachment B (Figure B-1). Although dose units are not stated 

in SANDOS output, review of other documents (SNL 1995a) 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.

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, 

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 
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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 check-out locations while at a given SNL site 

so that only site-specific background radiation is accumulated. The locations of badge storage areas 

for SNL-CA have not been identified.

Routine field dosimeters are assigned and issued by department managers and are subject to specific 

protocols for 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, 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.

In addition to personnel radiation dosimetry, SNL-CA conducts a Special Dosimetry Program that 

includes work area monitoring (Ward 1992; SNL 1995b). 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 
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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 

1995b).

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 the purposes of this SPD, 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.

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). 

Apparently, neutron film was exchanged on a monthly basis while a quarterly exchange period was 

generally implemented for beta/photon film. There is a discrepancy in the available record as to 

exactly when SNL-CA discontinued using film dosimeters and began using 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 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 
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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, 

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 

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, 
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analyzed, and stored with TLD Radiation Evaluation and Management System software on a desktop 

computer. Card reader results are converted into dose equivalent by a complex algorithm that was 

developed specifically for the Harshaw 8800/8812 system and SNL sites. 

Figure 6-2. Harshaw Model 8812 dosimeter assembly, element 

specifications, and respective radiation and dose equivalent quantities 

measured by Rhea and Bradley (1990).

Along with explicit protocols for card reader calibration, 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 that 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. At present, no information is available about the calibration of TLD badges SNL-CA 

used between 1971 and 1988 (the period in which RESL performed dosimetry services).

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 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.
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Figure 6-3. Example Harshaw 8800 card reader calibration procedures, forms, analysis, and output. 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 61 of 100

Figure 6-4. Example calibration form for the Harshaw 

4000 dosimetry 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, adjustments to recorded dose, 

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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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 

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 1995a). 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, Laudermilk, 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.”

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.

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 Beta particles Neutrons

<30 keV <15 keV <10 keV

30–250 keV >15 keV 10–100 keV

>250 keV 100 keV–2 MeV

2–20 MeV

>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. 

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.
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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).

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 2006a), 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 in which neutron exposures were possible. Table 6-4 shows multiplier values to use 

for dose reconstruction.

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 
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Table 6-2. Monthly and quarterly LLDs for the Harshaw Model 8802 

dosimeter card (adapted from Walker, Laudermilk, and Thompson 1997).

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.

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. No 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 dose conversion factors.

6.4.6 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-0006-6, Hanford Site – Occupational External Dose (ORAUT 2010d), with additional values for 

NTA film adapted from ORAUT-TKBS-0036-6, Argonne National Laboratory-East – Occupational 

External Dosimetry (ORAUT 2006b). Some general analogies can be drawn between Hanford and 

SNL-CA in terms of dosimetry technologies in use during various periods of interest.

Exposure Radiation Test Monthly LLD Quarterly LLD

Category Type(s) Depth (rem) (rem)

I Accident X-rays Deep 0.003 0.004

II Accident Gammas Deep 0.003 0.005

IIIA General X-rays Shallow 0.004 0.008

Deep 0.004 0.006

IV Cs-137 Gammas Shallow 0.003 0.007

Deep 0.003 0.004

VA Beta Particles Shallow 0.003 0.007

VI Moderated Neutrons Deep 0.003 0.005

VII (mixtures) 

III + IV Shallow 0.005 0.010

Deep 0.004 0.006

III + VA Shallow 0.004 0.008

Deep 0.005 0.008

III + VI Deep 0.003 0.004

IV + VA Shallow 0.004 0.008

Deep 0.004 0.006

IV + VI Deep 0.003 0.004
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Table 6-3. Selection of radiation energies and percentages.

Process description Radiation type Energy selection (keV) Percentagea

Weapons Laboratory Facility Complex

(buildings 910, 912, 913, 914, 916, 918)

Years: 1958–1998

Test/repair of neutron and X-ray 

detectors (neutron and X-ray 

generators)

Photon

Neutron

>250

10–100

100–2,000

2,000–20,000

100

5

5

90

Wet machining of DU Beta

Photon

>15

30–250

>250

100

50b

50b

Radiography for weapons 

components

Beta

Photon

>15

30–250

>250

100

50b,c

50b,c

Radiography for materials science 

studies (X–ray diffraction operations 

later moved to Building 941)

Beta

Photon

>15

<30

30–250

>250

100

40

30

30

H-3 storage studies Photon 30–250

>250

50b,d

50b,d

Ion beam analysis of materials Photon 30–250

>250

30

70

Radiflo leak tests Beta

Photon

>15

30–250

>250

100

10

90

Radiography

(building 923)

Years: unknown–early 1990s

Radiography using X–rays, gamma 

rays, neutrons, alpha and beta 

particles

Beta

Photon

Neutron

>15

<30

30–250

>250

10–100

100–2,000

2,000–20,000

100

5

45

50

5

70

25

Micro and Nano Technologies Laboratories

(buildings 941, 942, and 943)

Years: unknown–present

Radiography for materials science 

studies (radiological materials in 

Building 941 only)

Beta

Photon

>15

<30

30–250

>250

100

40

30

30

Explosives and Environmental Testing Complex

(buildings 955, 956, 966, 972, 974, 976, 977, 978, 979, 981, 983)

Years: 1958–present

Environmental testing of mock-up 

weapons and components (DU)

Beta

Photon

>15

30–250

>250

100

50

50

Storage Facilities

(buildings 921, 927, 961, 982)

Years: unknown–present

Storage and packaging of waste 

materials

Beta

Photon

>15

30–250

>250

100

70

30

a. Estimated primarily by site personnel.

b. Based on default assumptions favorable to claimants for DU in ORAUT (2007c).

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.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 66 of 100

Table 6-4. Neutron dose conversion factors 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

0.1–2.0 MeV 10.49 20 0.05 0.095

2.0–20.0 MeV 7.56 10 0.90 1.190

Radiography 10–100 keV 5.38 10 0.05 0.093

0.1–2.0 MeV 10.49 20 0.70 1.335

2.0–20.0 MeV 7.56 10 0.25 0.331

a. From Table 6-3.

b. Multiply reported dose by these factors to determine the corrected neutron dose equivalent for each applicable neutron 

energy interval.

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; quarterly 180; 60

1959–1971 Two-element beta/photon film 30b Monthly; quarterly 180; 60

1956–1971 Neutrons (NTA film) 50b Monthly; quarterly 300; 100

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 2007c,e).

c. Estimated MDL based on contemporary TLD systems of the period (ORAUT 2010d,e, 2006b).

d. Although neutron doses are reported for the two-chip system, no documentation of the neutron dosimetry methodology 

employed has been found.

Table 6-6. Dose units for organ dose 

conversion factors.

Years

Photon dose units for 

use with organ dose 

conversion factors

1956 - 1959 R

1959 - 1971 R

1972 - 1982 R

1982 - 1988 R

1989 - 1990 Hp(10)

1991-current Hp(10)
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 67 of 100

Table 6-7. Systematic uncertainty estimates (adapted from

 ORAUT 2010d).

Dosimeter Period of use

Systematica

Uncertainty factors

Two-element film 1956–1971 1.1

Two-chip TLD 1972–1982 1.05

Three-chip TLD 1982–1988 1.05

Multielement TLD 1989–present 1.05

NTA filmb 1956–1971 1.5

a.Systematic uncertainty from lack of knowledge about energy distributions

 and geometries.

b. The estimates most favorable to claimants from the various energy ranges

 in ORAUT (2006b) are applied to all applicable energy ranges.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 68 of 100

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 Dosimetrist. October 2006. The X-ray 

record form included in claim files has a place for the technician to record the projection, 

mA, kVp, distance, and time. For PA chests, the mA is almost always listed as 5, the 

distance is listed as 72", and the kVp is usually around 70. Exposure times are not usually 

listed. It is assumed that the person completing this part of the X-ray record form was 

recording the mAs, not the mA, primarily since radiographic machines are not designed to 

operate at such low mA settings, and also because time settings were not recorded. Five 

mAs would not be unusual for a PA chest exposure.

[2] Lopez, Theresa. ORAU Team. Senior Toxicologist. September 2006. 

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. 

Per review of X-ray records and at the direction of ORAU in comments dated October 

2006, the X-ray machines have been assumed to be single-phase.

[4] Lopez Theresa. ORAU Team. Senior Toxicologist. September 2006.

Lumbar spine X-rays were performed at hire from 1956 through 1971 as evidenced by medical 

records in claim files.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 69 of 100

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Kingsley, Sandia National Laboratories, Albuquerque, New Mexico, October 13. [SRDB Ref 

ID: 20326, p. 15]

Tucker, G. E. Jr., 1978, “Division 3313 Progress Report - 1977 Fourth Quarter,” memorandum to W. 

D. Burnett, Sandia National Laboratories, Albuquerque, New Mexico, February 6. [SRDB Ref 

ID: 20326, p. 7]

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, 

University of California, Lawrence Livermore National Laboratory, Livermore, California, April 

15. [SRDB Ref ID: 14835]

Turner, J. E., 1996, Atoms, Radiation and Radiation Protection, 2nd Edition, John Wiley & Sons, New 

York, New York.

Ullrich, R. A., 2003, Cold War Context Statement, Sandia National Laboratories, California Site, 

SAND2003-0112, Sandia National Laboratories, Albuquerque, New Mexico, January. [SRDB 

Ref ID: 22926]

Vosburg, S. K, 1993, “Sandia National Laboratories, Statement of Work for Bioassay Radiochemical 

Analysis Laboratory,” letter to S. N. Bakhtiar (Reynolds Electrical & Engineering Company), 

Sandia National Laboratories, Albuquerque, New Mexico, July 16. [SRDB Ref ID: 22864]

Walker, S., 1995, “Dose Assignment for [Privacy Act information],” memorandum to L. Lippert, Sandia 

National Laboratories, Albuquerque, New Mexico, June 22 [SRDB Ref ID: 18285]

Walker, S., 1996, “Environmental Background Accumulation Rates for SNL-NM, SNL-CA, and SNLNV,” memorandum to DPR/REP Files, Sandia Corporation, Sandia National Laboratories, 

Albuquerque, New Mexico, December 5. [SRDB Ref ID: 19251]

Walker, S., 1997a, “Interim Processing Steps for EXTRAD and Model 8807 Dosimeters,” 

memorandum to L. Garcia-Larkin and J. Zamora, Sandia Corporation, Sandia National 

Laboratories, Albuquerque, New Mexico, August 11. [SRDB Ref ID: 19247]

Walker, S., 1997b, “Lens of the Eye Dose Equivalent (LDE) Monitoring at SNL,” memorandum to 

PCS/TRK Files, Sandia Corporation, Sandia National Laboratories, Albuquerque, New Mexico, 

February 5. [SRDB Ref ID: 19249]

Walker, S., 1997c, “SANDOS Background Subtraction Problem,” memorandum to PCS/TRK Files, 

Sandia Corporation, Sandia National Laboratories, Albuquerque, New Mexico, April 8. [SRDB 

Ref ID: 19248]

Walker, S. W., F. R. Loudermilk, and D. J. Thompson, 1997, Lower Limit of Detection (LLD) for the 

Harshaw 8802 TLD. [SRDB Ref ID: 23577, p. 194–202]
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Wall, W. R., 1981, Tritium Control and Accountability Instructions, SAND80-8057, Sandia National 

Laboratories, Livermore, California, March. [SRDB Ref ID: 22994, p. 302–-314]

Wallace, S. A., 1988, “Current Status of 1985 Headquarters (HQ) Nuclear Facilities Appraisal (Ref. 

Audit E431),” memorandum to W. V. Ormand, Sandia National Laboratories, Albuquerque, 

New Mexico, September 9. [SRDB Ref ID: 23585]

Ward, D. C., 1992, Special Dosimetry Program, PRDP-00-03, Issue 01, Sandia National Laboratories, 

Albuquerque, New Mexico, October 23. [SRDB Ref ID: 18278]

Ward, D. C., 1994, External Dosimetry Program, PRDP-00-01, Issue 05, Sandia National 

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Willhoite, S. B., 1979, Determination of Gamma and Neutron Yearly Doses at the Building 212 BusStop, December 12. [SRDB Ref ID: 15066]

Wright, D., 1993, “History of Radiation Dosimetry at SNL-CA as Remembered by D.A. Wright, 8641,” 

memorandum to L. Perez, Sandia National Laboratories, Livermore, California, June 1. 

[SRDB Ref ID: 24811]

Wright, D. A., 1979a, “Sandia Livermore’s Bioassay Program from Depleted Uranium,” memorandum 

to Distribution, Sandia National Laboratories, May 14. [SRDB Ref ID: 24813]

Wright, D. A., 1979b, “Radiation Workers at SLL,” memorandum to H. M. Patterson, Sandia National 

Laboratories, December 20. [SRDB Ref ID: 24824]

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Operations Office), Sandia National Laboratories, Livermore, California, February 19. [SRDB 

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National Laboratories, Livermore, California, August. [SRDB Ref ID: 22994]

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GLOSSARY

alpha particle (α)

See alpha radiation. 

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 particle (β)

See beta radiation. 

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

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, neutron 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.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 87 of 100

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.

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. 

fissionable

Capable of undergoing fission by capturing neutrons, including fast neutrons. Uranium-238 is 

fissionable. Fissionable indicates both spontaneous and induced fission.

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 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray 

radiation.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

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. See element.

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.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 88 of 100

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.

operating area

Usually refers to a major operational work area at a site.

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. 

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.

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. 01 Effective Date: 04/28/2014 Page 89 of 100

ATTACHMENT A 

SUMMARY OF ACTIVITIES AT SITE BUILDINGS

Building

Number

Building Use Radiological

Activities

Dates of 

Use

Nuclides Comments

904 Auditorium None 1980s–

present

905 Offices None 1980s–

present

906 Combustion Laboratory 

Research

None 1980s–

present

907 Equipment room None 1980s–

present

910 Weapons Laboratory 

Facility

Neutron 

detection

Used to design, 

assemble, test, 

calibrate and repair 

neutron detectors

911 Administration None

912 Administration, 

computers 

None 1958–

present

913 Test assembly, machine 

shop, plating, 

metallography, tritium 

high-velocity air hoods

Tritium storage 

studies, U 

machining

1976–

1998

H-3, U-238

914 Test laboratories Radiflow, 

nondestructive 

testing, U 

machining, 

Kr-85, beta 

sources, 

U-238

No radiological work 

for about 15 years

915 New building, office 

space

None

916 Mostly light chemistry 

laboratories, Previously 

a warehouse, Ar 

gloveboxes

E-microscopes, 

accelerators, 

various small 

sources, H-2 

storage tests

700 keV, 

1 MeV, 

U-238

Sealed sources

917 Cylinder storage

918 Raw stock U-machining U-238 Deconstructed in the 

late 1980s

920 Office space None 1977–

present

921 Decommissioned in 

1980

Radiological 

material storage 

and 

decontamination 

of Nevada Test 

Site test units

U-238, trace 

Pu-239 and 

mixed 

fission 

products

Became Building 

9632 in 1980 

(Carter 2009)

922 Office space None

923 Radiography Laboratory Radiography 

operations; X-ray 

machines 

Co-60, 

Ir-192 X-ray 

machines, 

Cf-252 

Converted to records

storage in early 

1990s

924 Mo Mobile counting 

laboratory

Health Physics

laboratory moved to 

Building 973

925 Health services None

927 Radiological Material 

Storage

Warehouse and 

vault

U-238, 

natural 
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 90 of 100

ATTACHMENT A

SUMMARY OF ACTIVITIES AT SITE BUILDINGS (continued)

Building

Number

Building Use Radiological

Activities

Dates of 

Use

Nuclides Comments

thorium

928 Shipping and receiving None

929 Office space None

940 Office space None

941 Light mechanical, 

electrical, chemical 

laboratories

Radiography X-ray U-238, 

beta 

sources

Sealed sources

942 Light mechanical, 

electrical, chemical 

laboratories

None

943 Plating operations

955 Environmental Test 

Facility

Environmental

testing of various 

test units

U-238 metal

956 Vibration Test Facility Environmental 

testing of various 

test units

U-238 metal

960 Offices

961 Radiological and mixed 

waste storage and 

packaging

Storage and

packaging

H-3, U-238

9611 Chemical waste storage None

963 Maintenance Facilities None

964 Security None

966 High Pressure Gas 

Dynamics Test Facility

None

967 Office None

968 TRL Tritium research 1976–

1989

H-3, U-238 

beds

Decommissioned in 

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

972 Centrifuge Test package 

centrifuge, 

penetrator 

studies

U-238

973 Environment, safety, 

and health laboratories

974 Explosives test tanks Neutron 

generator tests

H-3

976 Four test cells for high 

press work

H-2 storage tests U-238

978 Flight test unit testing Mass properties 

tests

U-238

979 H-2 storage, research 

and development

U-238 Three Ar gloveboxes, 

two contaminated 

machines

983 Test cells; flight test 

assembly

U-238
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 91 of 100

ATTACHMENT A

SUMMARY OF ACTIVITIES AT SITE BUILDINGS (continued)

Building

Number

Building Use Radiological

Activities

Dates of 

Use

Nuclides Comments

Explosives 

Storage 

Area

Magazines; explosives 

storage

H-3

Sources: DOE (1982); Wright (1981a, 2006).
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 92 of 100

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES

LIST OF FIGURES

FIGURE TITLE PAGE

B-1 Example external dosimetry history records (output from SANDOS)...................................... 93

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

(potential neutron exposures)................................................................................................. 94

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

(potential neutron exposures)................................................................................................. 95

B-4 Example unreturned dosimeter investigation report................................................................ 96

B-5 Example unreturned dosimeter questionnaire follow-up letter ................................................ 97

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

the Personnel Dosimetry Division........................................................................................... 98

B-7 Work Area Radiation Dosimetry Request Form...................................................................... 99

B-8 Example work area radiation dosimetry results report .......................................................... 100
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 93 of 100

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. 01 Effective Date: 04/28/2014 Page 94 of 100

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. 01 Effective Date: 04/28/2014 Page 95 of 100

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. 01 Effective Date: 04/28/2014 Page 96 of 100

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-4. Example unreturned dosimeter investigation report.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 97 of 100

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. 01 Effective Date: 04/28/2014 Page 98 of 100

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. 01 Effective Date: 04/28/2014 Page 99 of 100

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-7. Work Area Radiation Dosimetry Request Form.
Document No. ORAUT-TKBS-0053 Revision No. 01 Effective Date: 04/28/2014 Page 100 of 100

ATTACHMENT B

DOSIMETRY PROGRAM ADMINISTRATIVE DOCUMENTATION EXAMPLES (continued)

Figure B-8. Example work area radiation dosimetry results report.