ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 120

DOE Review Release 02/05/2013

Document Title:

Site Profile for Sandia National Laboratories 

in Albuquerque, New Mexico, and the 

Tonopah Test Range, Nevada

Document Number: ORAUT-TKBS-0037

Revision: 01

Effective Date: 01/25/2013

Type of Document: TBD

Supersedes: Revision 00

Subject Expert(s): Rowena S. Argall and Tracy A. Ikenberry

Approval: Signature on File Approval Date: 01/22/2013

Dann C. Smith, Document Owner

Concurrence: Signature on File Concurrence Date: 01/22/2013

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 01/22/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 01/22/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 01/25/2013

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 2 of 120

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

06/22/2007 00 New approved site profile for the Sandia National Laboratories in 

Albuquerque, New Mexico and the Tonopah Test Range, Nevada. 

Incorporates formal internal and NIOSH review comments. There is 

a reduction in assigned dose and no PER is required. Training 

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

Buddenbaum.

01/22/2013 01 Revision initiated to incorporate Special Exposure Cohort (SEC) 

Petition SEC-00162 and added Section 1.3 to incorporate the SEC 

class through 1962. Revised Acronyms and Abbreviations list and 

deleted unused and unnecessary items. Revised Section 1.2 to refer 

to “Section 7.0” for attributions and annotations. Revised Section 2.3 

to address significant use of radioactive materials and external dose 

monitoring starting in 1949. Table 2-2: relocated “Radionuclides” to 

that column for line entry “B-884,” edited listed radiation types for 

consistent terms and case (gamma and neutron), and added 

“Particle” to “AMAD” in “Radionuclides” column heading. Revised 

Section 2.6.2 title to describe section better. Updated reference 

citations throughout the document. Section 3 totally revised for 

alignment with current ORAUT-OTIB-0006 guidance and methods. 

Revised described areas of environmental and onsite dose to be 

outside “radiological” facilities in Section 4.1.1. Revised Section 

4.1.2 to describe incorporation of post-2004 environmental data. 

Revised new Section 4.6 to clarify the use of a geometric standard 

deviation (GSD) of 3 for environmental intake doses. Table 5-5: 

revised citations for to references SNL 1998d through 1998l. 

Clarified language in Section 5.1.2 describing when bioassay was 

performed. Per review comment, deleted Table 5-7 since little is 

currently known of site plutonium absorption types. Deleted alternate 

unit “(MBq),” which was incorrect, listed for Pu-238 1967–1991 in 

new Table 5-7. Revised reference for tritium dose guidance from 

OCAS-TIB-002 to ORAUT-OTIB-0022 in Section 5.2.3. Added 

guidance to Section 5.2.4 to apply natural uranium bioassay results 

even if source may be nonoccupational. Table 5-15: updated IMBA 

reference to current version, 1.0.9, and revised specific activity 

values to those provided by IMBA. Deleted Section 5.3.6, as 

applicable guidance is provided by ORAUT-OTIB-0060. Deleted 

Table 5-29 and related text due to insignificant value for dose 

reconstruction. Revised name of Section 5.4.2 to Incident Intake 

Information. Revised Section 6.3 title to include electron dose. 

Added discussion to Section 6.3 to require use of coworker dose of 

ORAUT-OTIB-0072 as applicable. In Section 6.6, revised photon 

energy listed as “30 to 2,560” to “30 to 250” and described radiation 

type for discussion of <30-keV dose as photons. Revised guidance 

in Section 6.8 for use of neutron-to-photon ratio for missed neutron 

dose as being applicable to missed photon dose rather than recorded 

gamma dose. Added “Attribution and Annotation” for the use of a 

GSD of 3 for environmental intake doses. Deleted term “activity 

fraction” from Glossary. Incorporates Special Exposure Cohort 

(SEC) Petition SEC-00188, allow dose reconstruction for site starting 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 3 of 120

in 1945, and add updated on-site ambient doses and environmental 

intake data for 2005 through 2010. Added START Laboratory to 

Acronyms and Abbreviations list. Revised Sections 1.3 and 5.0 to 

implement SEC-00188 class. Revised Section 2.3 to better define 

site activities during 1945 through 1949. Revised text throughout 

Section 4 to indicate that ambient and environmental dose data are 

now provided through 2010. Revised Section 4.3 to and Table 4-1 

include an additional radionuclide due to Pu-239 release values after 

2004. Added maximizing and best estimate ambient dose guidance 

to Section 4.4 and added reference (ORAUT 2006b) regarding 

ambient dose. Added references throughout Sections 4.4 and 4.5 

and the “References” section to include environmental monitoring 

reports through 2010. Deleted detailed discussion of environmental 

TLD uncertainty since it was not relevant to the applied dose 

reconstruction ambient dose uncertainty and was not supported by 

citable sources for all years. Added a new Table 4-3 for 2005 – 2010 

ambient doses and revised all subsequent Section 4 table numbers. 

Revised Section 4.4.2 TTR ambient dose text that indicated the 

ambient dose was zero due to positive ambient doses after 2004. 

Provided guidance regarding noble gases intakes in Section 4.5.1. 

Revised Section 4.5.2 to increase the assumed light activity 

breathing rate to that used for SNL – NM, which is an overestimated 

value. Revised text in Section 4.6, “Uncertainty” to delete 

superfluous text that did not address uncertainty and to apply 

ambient dose uncertainties consistent with ORAUT 2006b guidance. 

Revised Table 5-8 to correct activity values and ratios. Removed 

information associating specific radionuclides with specific facilities 

and building numbers; removed text that was not necessary for SNL 

dose reconstructions, including associated references, attributions, 

Figure 5-1, and Tables 5-28 and 5-29; and removed text discussing 

employees by name in the body of the TBD aside from citing 

attributions and references. Table 3-1: changed 1957–1966 X-ray 

frequency to annual. Sections 4.4.1 and 4.4.2 and Tables 4-3 and 4-

4: added detailed 2008 and 2009 environmental intake data. Table 

5-1: deleted “Manzano” as a TA. Tables 5-7, 5-11, 5-13, 5-16, 5-20, 

and 5-22: changed table titles to incorporate the term “MDAs,” 

consistent with text and with internal DR terminology. Changed 

Sections 5.3.1 and 5.3.4, including deletion of Tables 5-21 and 5-27, 

to refer to LANL TBD for guidance on internal DR using LANL in vivo

measurements. Section 5.4 title changed to reflect that “selected 

buildings” are no longer addressed. Sections 6-1and 6-2 and Tables 

6-1 and 6-7: changed to indicate that LANL dosimetry parameters 

apply prior to 1949. In Section 6-7, added more detailed guidance to 

apply a neutron-to-photon dose ratio. In Section 6-8, added more 

detailed guidance regarding the applicability of neutron missed dose 

and added the associated reference (ORAUT 2008b). Incorporates 

formal internal and NIOSH review comments. Constitutes a total 

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

Objective Manager. Initiated by Dann C. Smith.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 4 of 120

TABLE OF CONTENTS

SECTION TITLE PAGE

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

1.0 Introduction ............................................................................................................................ 12

1.1 Purpose...................................................................................................................... 13

1.2 Scope......................................................................................................................... 13

1.3 Special Exposure Cohort Information for SNL-NM...................................................... 13

2.0 Site Description...................................................................................................................... 14

2.1 Purpose and Scope.................................................................................................... 14

2.2 Introduction................................................................................................................. 14

2.3 Site Activities and Processes...................................................................................... 14

2.4 Major Facilities and Activities...................................................................................... 15

2.4.1 TA-I................................................................................................................. 19

2.4.2 TA-II................................................................................................................ 20

2.4.3 TA-III............................................................................................................... 20

2.4.4 TA-IV .............................................................................................................. 20

2.4.5 TA-V ............................................................................................................... 20

2.4.6 Other Areas .................................................................................................... 21

2.5 Major Site Incidents .................................................................................................... 21

2.6 Health Protection Practices ........................................................................................ 21

2.6.1 Personnel Monitoring ...................................................................................... 21

2.6.1.1 Badging ............................................................................................ 21

2.6.1.2 Area Monitoring ................................................................................ 22

2.6.2 Access Control................................................................................................ 22

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

3.1 Introduction................................................................................................................. 22

3.1.1 Purpose .......................................................................................................... 22

3.1.2 Scope ............................................................................................................. 22

3.2 Examination Types and Frequencies.......................................................................... 22

3.3 Technique Factors and Incident Air Kerma................................................................. 24

3.3.1 Photofluorography, 1943-1956 (at LANL)........................................................ 24

3.3.2 Radiography, 1943 Through 1952 (at LANL)................................................... 24

3.3.3 Radiography, 1953 Through 1977 (at SNL)..................................................... 24

3.3.4 Radiography, 1978 Through 1995................................................................... 25

3.3.5 Radiography, 1996 to Present......................................................................... 26

3.4 X-Ray Doses to Workers ............................................................................................ 27

3.4.1 Conversion of ESE to Dose............................................................................. 27

3.4.2 Organ Doses from Chest Photofluorography................................................... 27

3.4.3 Organ Doses from 14- by 17-in. PA Chest Radiography ................................. 27

3.4.4 Organ Doses from Lumbar Spine Radiography............................................... 27

3.5 Uncertainty Analysis ................................................................................................... 28

4.0 Occupational Environmental Dose ......................................................................................... 37

4.1 Introduction................................................................................................................. 37

4.1.1 Purpose .......................................................................................................... 37

4.1.2 Scope ............................................................................................................. 38

4.2 Operations Overview .................................................................................................. 38

4.2.1 Sandia National Laboratories – New Mexico................................................... 38
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4.2.2 Sandia National Laboratories – Nevada.......................................................... 39

4.3 Radionuclide Screening, SNL-New Mexico................................................................. 41

4.4 Ambient External Radiation ........................................................................................ 42

4.4.1 Sandia National Laboratories – New Mexico................................................... 42

4.4.2 Sandia National Laboratories – Nevada.......................................................... 44

4.5 Inhalation of Onsite Airborne Radionuclides ............................................................... 45

4.5.1 Sandia National Laboratories – New Mexico................................................... 45

4.5.1.1 Technical Areas I, II, and IV.............................................................. 46

4.5.1.2 Technical Area III.............................................................................. 47

4.5.1.3 Technical Area V .............................................................................. 48

4.5.2 Sandia National Laboratories – Nevada.......................................................... 49

4.6 Uncertainty ................................................................................................................. 50

5.0 Occupational Internal Dose.................................................................................................... 50

5.1 Introduction................................................................................................................. 51

5.1.1 Bioassay Results of Individuals....................................................................... 57

5.1.2 The Bioassay Program.................................................................................... 58

5.2 In Vitro Bioassay......................................................................................................... 61

5.2.1 Plutonium........................................................................................................ 62

5.2.2 Americium....................................................................................................... 67

5.2.3 Tritium............................................................................................................. 68

5.2.4 Uranium .......................................................................................................... 70

5.2.5 Fission and Activation Product Analysis.......................................................... 73

5.2.6 Accelerator Areas ........................................................................................... 74

5.2.7 Indoor Radon .................................................................................................. 74

5.2.8 Other Limited-Exposure Radionuclides ........................................................... 75

5.3 In Vivo Minimum Detectable Activities, Analytical Methods, and Reporting Protocols. 76

5.3.1 Whole-Body Counters–LANL (1955 to 1992) .................................................. 76

5.3.2 Whole-Body Counter–SNL.............................................................................. 76

5.3.3 Cesium-137 Intakes from Fallout .................................................................... 79

5.3.4 Lung Burdens ................................................................................................. 79

5.3.5 Wound Monitoring........................................................................................... 80

5.4 Air ConcentrationS ..................................................................................................... 80

5.4.1 Respiratory Protection Program ...................................................................... 80

5.4.2 Incident Intake Information.............................................................................. 81

6.0 Occupational External Dose................................................................................................... 81

6.1 Dosimeters Used........................................................................................................ 81

6.2 Recorded Dose Practices ........................................................................................... 83

6.3 Unmonitored Photon and electron Dose ..................................................................... 84

6.4 Uncertainty Factors .................................................................................................... 87

6.5 Missed Beta/Photon Dose .......................................................................................... 87

6.6 Partitioning of Beta/Photon Doses to Energy Categories ............................................ 88

6.7 Unmonitored Neutron Dose ........................................................................................ 90

6.8 Missed Neutron Dose ................................................................................................. 90

6.9 Partitioning of Neutron Doses to Energy Categories................................................... 91

6.10 Recommended Dose Conversion Factors .................................................................. 92
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7.0 Attributions and Annotations .................................................................................................. 92

References ........................................................................................................................................ 98

Glossary .......................................................................................................................................... 119

LIST OF TABLES

TABLE TITLE PAGE

2-1 Chronology of significant SNL programs and events .............................................................. 16

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

2-3 Magnitude of activity ..............................................................................................................18

3-1 Frequency and types of medical radiographs......................................................................... 23

3-2 Machine settings and incident air kerma for screening radiography, 1953 through 1977........ 25

3-3 Machine settings and incident air kerma for screening radiography, 1978 through 1995........ 26

3-4 Manual settings and incident air kerma for screening medical radiography, 1997 to 

present...................................................................................................................................26

3-5 Organ dose equivalents for chest projections for all periods................................................... 29

3-6 Skin dose guidance and skin dose equivalents for chest projections, 1953 through 

1977.......................................................................................................................................30

3-7 Skin dose guidance and skin dose equivalents for chest projections, 1978 through 

1995.......................................................................................................................................32

3-8 Skin dose guidance and skin dose equivalents for chest projections, 1996 through 

present...................................................................................................................................33

3-9 Organ dose equivalents for lumbar spine projections for all periods....................................... 35

3-10 Skin dose guidance and skin dose equivalents for lumbar spine projections, 1953 

through 1977..........................................................................................................................35

3-11 Skin dose guidance and skin dose equivalents for lumbar spine projections, 1978 

through 1985..........................................................................................................................36

4-1 Radionuclides meeting screening criteria and the principal worker exposure pathway ........... 42

4-2 External radiation dose for workers in SNL-NM TAs............................................................... 43

4-3 External radiation dose for workers in SNL-NM TAs, 2005 through 2010............................... 44

4-4 External radiation dose for workers at TTR ............................................................................ 45

4-5 TAs-I, -II, and -IV maximum annual intakes via inhalation ...................................................... 47

4-6 TA-III maximum annual intakes via inhalation ........................................................................ 47

4-7 TA-V maximum annual intakes via inhalation......................................................................... 48

4-8 TTR maximum annual intakes via inhalation .......................................................................... 50

5-1 Internal exposure potential by area ........................................................................................ 53

5-2 Facilities in TAs-III and -IV and the Explosive Component Facility ......................................... 53

5-3 Nuclear weapons and nosetip tests........................................................................................ 55

5-4 1991 Tiger Team assessment key findings related to the internal dosimetry program............ 59

5-5 Internal dosimetry requirements............................................................................................. 61

5-6 Action levels for airborne radioactivity .................................................................................... 61

5-7 Plutonium bioassay MDAs as listed in LANL procedures and reports..................................... 64

5-8 Pure 238Pu isotopic mixture..................................................................................................... 65

5-9 Activity composition of reference weapons-grade (6%) plutonium mixture ............................. 66

5-10 Activity composition of reference fuel-grade (12%) plutonium mixture.................................... 67

5-11 Americium-241 LANL bioassay techniques and MDAs........................................................... 68

5-12 Other sources of potential tritium intakes ............................................................................... 69

5-13 Tritium urine bioassay MDAs from LANL................................................................................ 70

5-14 Uranium conversion factors.................................................................................................... 71
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5-15 Natural uranium .....................................................................................................................71

5-16 Routine uranium urinalysis MDAs .......................................................................................... 72

5-17 Fission and activation product nuclides.................................................................................. 73

5-18 Specific activity of thorium isotopes........................................................................................ 75

5-19 Routine 210Po urinalysis MDAs ............................................................................................... 75

5-20 MDAs for whole-body counting, 2005..................................................................................... 76

5-21 MDAs for whole-body counting .............................................................................................. 77

5-22 Nuclides that produce the highest dose to organ of interest ................................................... 78

5-23 Accelerator nuclides in whole-body count library.................................................................... 78

5-24 Mean United States 137Cs body burdens from fallout............................................................ 79

5-25 Reported exposure incidents and results ............................................................................... 81

6-1 Beta/photon dosimeters ......................................................................................................... 82

6-2 Neutron dosimeters................................................................................................................ 82

6-3 Recorded dose practices........................................................................................................ 83

6-4 SNL-NM worker penetrating dose statistics, 1949 to 2005 ..................................................... 85

6-5 SNL-NM worker nonpenetrating dose statistics, 1987 to 2005 ............................................... 86

6-6 SNL beta/photon dosimeter period of use, type, MDL, exchange frequency, and 

potential annual missed dose................................................................................................. 87

6-7 Recommended beta and photon radiation energies and percentages for SNL-NM 

facilities ..................................................................................................................................89

6-8 Recommended distributions for neutron-to-photon ratio......................................................... 90

6-9 Neutron dosimeter period of use, type, MDL, exchange frequency, and potential 

annual missed dose ...............................................................................................................91

6-10 Recommended dose fractions and ICRP Publication 60 correction factors for SNL 

neutron sources ..................................................................................................................... 92

6-11 Recommended DCFs for SNL dose assessments.................................................................. 92

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 SNL-NM site south of Albuquerque, New Mexico, in 2005 ..................................................... 38

4-2 Layout and location of facilities operated by SNL-NV at TTR in 2004..................................... 40
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ACRONYMS AND ABBREVIATIONS

ABS acrylonitrile-butadienestyrene

ACPR Annular Core Pulse Reactor (later Annular Core Research Reactor)

ACRR Annular Core Research Reactor

AEC U.S. Atomic Energy Commission

AFB Air Force Base

ALARA as low as reasonably achievable

AMAD activity median aerodynamic diameter

ANSI American National Standards Institute

ANT Advanced Nosetip Test

AP anterior-posterior

AT&T American Telephone and Telegraph

BB bronchial region

BBbas bronchial region, basal cells

bbsec bronchiolar region, secretory cells

Bq becquerel (1 disintegration per second)

BZA breathing-zone air 

C coulomb

CAM continuous air monitor

CEDE Committed Effective Dose Equivalent

CEP Controls for Environmental Pollution

CFR Code of Federal Regulations

cGy centigray

CHP Certified Health Physicist

Ci curie

cm centimeter

cpm counts per minute

d day

DAC derived air concentration

dc direct current

DCF dose conversion factor

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

ECF Explosive Component Facility

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ENSD entrance skin dose (from X-ray procedures)

EPA U.S. Environmental Protection Agency

ES&H Environment, Safety and Health

ESE entrance skin exposure

ET extrathoracic 

EXSD exit skin dose (from X-ray procedures)

ft foot

FP fission product

g gram

GIF Gamma Irradiation Facility

GM geometric mean
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GSD geometric standard deviation

HCF Hot Cell Facility

HP health physics

HPGe hyperpure germanium

hr hour

HT elemental tritium (tritiated gas)

HTO tritium oxide (water or water vapor)

HVL half-value layer

ICP-MS inductively coupled plasma mass spectroscopy

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

Ka,i incident air kerma

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

KPA kinetic phosphorescence analysis

kVp kilovolts peak

kW kilowatt

L liter

LANL Los Alamos National Laboratory

LASL Los Alamos Scientific Laboratory

LAT lateral

LC decision level

LICA Low Irradiation Calibrator Apparatus

LLI lower large intestine

LLNL Lawrence Livermore National Laboratory

LN lymph nodes

LSC liquid scintillation counter

m meter

mA milliampere

mAs milliampere-second

mCi millicurie

MDA minimum detectable activity

MDL minimum detection limit

MeV megaelectron-volt, 1 million electron-volts

mg milligram

mL milliliter

mm millimeter

MOU memorandum of understanding

MPC maximum permissible concentration

mR milliroentgen

mrad millirad

mrem millirem

MT metal tritide (tritium bound to metallic compounds, such as hafnium) 

MW megawatt

MWL Mixed Waste Landfill
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n neutron

nCi nanocurie 

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NTA nuclear track emulsion, type A

NTS Nevada Test Site

NTTR Nevada Test and Training Range

NU natural uranium

OBT organically bound tritium

PA posterior-anterior

PBFA Particle Beam Fusion Accelerator

pCi picocurie

PER Program Evaluation Report

PFG photofluorography

PHA pulse height analysis

POC probability of causation

R roentgen

RAO right anterior oblique (X-ray position) 

RAS radiometric alpha spectroscopy

RBM red bone marrow

REBA Relativistic Electron Beam Accelerator

RPID Radiation Protection Internal Dosimetry

RWP radiological work permit

s second

SABRE Sandia Accelerator and Beam Research Experiment

SEC Special Exposure Cohort

SER Sandia Engineering Reactor

SERF Sandia Engineering Reactor Facility

SI small intestine

SID source-to-image distance

SMT stable metal tritide

SNL Sandia National Laboratories

SNL-NM SNL Albuquerque, New Mexico

SNL-NV SNL Tonopah Test Range in Nevada

SPHINX Short Pulse High Intensity Nanosecond X-Radiator

SPR Sandia Pulse Reactor

SRDB Ref ID Site Research Database Reference Identification

SSD source-to-skin distance

START Sandia Tomography and Radionuclide Transport (Laboratory)

TA Technical Area

TBD technical basis document

TH thoracic

TLD thermoluminescent dosimeter

TRIGA Training, Research, Isotopes General Atomics

TTR Tonopah Test Range

ULI upper large intestine

U.S.C. United States Code
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wk week

yr year

γ gamma particle or ray

μCi microcurie

μg microgram

μm micrometer

§ 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 is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 13 of 120

1.1 PURPOSE

This site profile for Sandia National Laboratories (SNL) provides background information in relation to 

dose reconstruction for SNL workers at the Albuquerque, New Mexico (SNL-NM), site and the 

Tonopah, Nevada (SNL-NV), site on the Tonopah Test Range (TTR). The site profile also provides 

details about past and current SNL practices used to assess radiation exposures and environmental 

radiation levels at the facilities. In this document, SNL refers either to the Laboratories as a whole or 

to the Albuquerque site; SNL-NV refers to the TTR site.

1.2 SCOPE

Section 2.0 provides an overview of historical operations at SNL-NM and SNL-NV that have involved 

potential external or internal radiation exposures to workers.

Section 3.0 provides information about the doses that individual workers could have received from 

medical X-rays that were required for screening and as a condition of employment. These X-rays 

included preemployment and periodic chest X-rays during required physical exams.

Section 4.0 presents environmental dose information for workers who received doses from inhalation 

of radioactive materials in the air, direct radiation from plumes (immersion dose from radioactivity in 

the air), contact with radioactive particles on the skin, and direct exposure to radionuclides in use, 

stored, and incorporated in the soil when working outside of monitored radiologically controlled areas 

on site.

Section 5.0 discusses the internal dosimetry program at SNL, including minimum detectable activities 

(MDAs). Details of the monitoring techniques and programs are also presented. 

Section 6.0 discusses the program for measuring shallow and penetrating external doses to the 

workers. The methods for evaluating external doses to workers have evolved over the years as new 

techniques and equipment have been developed.

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 INFORMATION FOR SNL-NM

Classes Added to the Special Exposure 

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

contractors and subcontractors who worked in any area at Sandia National 

Laboratories in Albuquerque, New Mexico, from January 1, 1949, through 

December 31, 1962, 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 2011).

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

contractors and subcontractors who worked in any area at Sandia National 

Laboratories in Albuquerque, New Mexico, from January 1, 1963, 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 2012).
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NIOSH has determined that monitoring data, process information, and monitoring program information 

are insufficient to support bounding internal doses for the evaluated classes. Therefore, NIOSH 

concluded it cannot bound internal doses for the period from January 1, 1949, through December 31, 

1994.

Dose reconstruction guidance in this site profile for periods between January 1, 1949 and December 

31, 1994 is presented to provide a technical basis for partial dose reconstructions for nonpresumptive 

cancers not covered in the Special Exposure Cohort (SEC) class.. Although NIOSH found that it is 

not possible to reconstruct internal radiation doses completely for the proposed class, it intends to use 

internal and external monitoring data that may be 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 but who do not qualify for inclusion in the SEC, can 

be performed using these data as appropriate.

The above SEC discussion is not applicable to the reconstruction of TTR occupational radiation dose, 

as the designated SEC classes do not apply to TTR. 

2.0 SITE DESCRIPTION

2.1 PURPOSE AND SCOPE

This section provides a site description that contains technical basis information for the evaluation of 

the total occupational dose for EEOICPA claimants. It describes SNL facilities and processes and 

details the historical information related to worker internal and external exposures.

2.2 INTRODUCTION

SNL had its origin as a satellite support site for the Los Alamos Scientific Laboratory [LASL; later 

renamed Los Alamos National Laboratory (LANL)]. The LASL Z-Division was established in July 

1945 to handle weapons development, testing, and bomb assembly for the Manhattan Engineer 

District (Ullrich 1998). In late fall of 1945, some units of Z-Division moved to the current site near 

Albuquerque. In 1948, Z-Division became a separate branch of LASL, and a year later it was 

renamed Sandia Corporation. Its functions included engineering research, development, and 

mechanical production of specialized and strategic systems components to support U.S. national 

security. The site became Sandia Laboratories in 1971 and underwent another name change in 1979 

to Sandia National Laboratories [1]. The SNL-NM facility includes five technical areas (TAs) south of 

Albuquerque, New Mexico, on the Kirtland Air Force Base (AFB). 

2.3 SITE ACTIVITIES AND PROCESSES

The site mission transitioned from its start in 1945 to performing primary weapons functions by 1949 

(Johnson 1997; Ullrich 1998, 1999). Starting in 1945, the site was first occupied by Z-Division with 

the primary projects being planning, engineering, storage of nonnuclear materials for weapons, and 

construction. During 1946, 1947, and much of 1948, work activities consisted primarily of 

engineering, testing, nonnuclear assembly, storage, and support of testing elsewhere. Occuring at

the site initially developed, this area became known as Tech Area 1 (TA-1). In 1948, TA-II became 

operational, supporting assembly work on a larger scale. Design, assembly, and testing functions 

grew with the Laboratory assuming a more principal national role and, in 1948, occupational 

radiological exposure started at the site.

From 1948 to 1959, three national laboratories were coordinated in nuclear weapons design and 

construction activities: LANL, SNL, and Lawrence Livermore National Laboratory (LLNL). The three 

laboratories were part of the larger network of U.S. Atomic Energy Commission (AEC)-controlled sites 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 15 of 120

that completed the manufacture, testing, and storage of nuclear weapons for the national stockpile. 

As described by Johnson (1997):

Los Alamos and Lawrence Livermore design the high explosive and nuclear system 

package, while Sandia designs the rest of the nuclear bomb or warhead, including the 

arming, fusing, and firing systems along with other essential components. In essence, 

Sandia “weaponizes” the nuclear systems designed at its partner laboratories.

Records of external radiation dose monitoring, along with positive results, began in 1949 (Widner 

2008a,b,c). Most radioactive components were handled away from the SNL-NM site in this early era. 

Activities included extensive field testing of weapons components until 1958, which began again in 

1961. The first nuclear reactor at SNL, the Sandia Engineering Reactor Facility (SERF), also referred 

to as the Sandia Engineering Reactor (SER) in historical records, went on line in 1957 and was used 

for material testing. 

In the early 1960s, several high-energy accelerators were added at SNL to facilitate research on the 

effects of pulsed radiation on a variety of materials. The interest in the effects of pulses of higher 

intensity radiation (higher than could be sustained continuously with either the SERF reactor or the 

accelerators of the era) led to the installation and startup of the Sandia Pulse Reactor (SPR) in 1961. 

This was a GODIVA-type reactor design that allowed running the core up to relatively high energy 

levels for very short periods to investigate material damage under controlled conditions (Burnett et al. 

1962). Over the succeeding decades, this design was upgraded to increasingly higher energy-level 

capacities to determine when exposed target materials fail or decompose. 

Over the years, the Laboratory mission expanded and new activities included working on technologies 

to monitor nuclear testing after the treaty of 1963, working with the National Aeronautics and Space 

Administration to enhance the safety of aerospace nuclear power systems, and developing 

conventional weapons and intrusion sensors for use in the Vietnam War. National and international 

events, including the energy crisis and the terrorist acts at the Munich Olympics of the early 1970s, 

caused SNL to become involved in new areas of energy research and in physical security and 

safeguards for facilities. However, SNL continued to have responsibilities in developing new nuclear 

and other weapons as well as maintaining the safety and reliability of the existing nuclear stockpile 

(Johnson 1997). 

Although international arms control efforts increased throughout the 1970s and 1980s, concerns 

remained about the impact of pulses or bursts of high radiation on both materials and functionality of 

electromagnetic and electronic systems. The relatively small energy level flash X-ray and heavy ion 

accelerators of the early 1970s were being rapidly augmented by 1- and 2-MeV machines. The 

continuing demand in the 1980s and 1990s for larger capacity and higher energy accelerators for 

testing and stressing electronic components with exposure to pulsed electron and ion accelerators 

and flash X-ray systems led to continual development changes and additions of higher energy 

accelerators at SNL during that period. 

Table 2-1 provides a chronology of SNL programs and events.

2.4 MAJOR FACILITIES AND ACTIVITIES

Early on, it became apparent that SNL would need to grow to house the increasing staff necessary to 

meet scientific demands. SNL currently consists of five TAs and several test areas. Each TA has its 

own distinctive operations, but the operations of some groups at SNL can span more than one TA. A 

description of each area is given below. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 16 of 120

Table 2-1. Chronology of significant SNL programs and events.

Year Activity

1945 Z-Division created at LANL to assemble stockpile materials.

1946 Z-Division moved to Albuquerque. Building constructed in TA-I for testing weapons components.

1948 TA-II, high-explosive final assembly area for nonnuclear components constructed; assembly 

continued for many years.

1949 Name change to Sandia Laboratory, separate from LANL.

1953 Large centrifuge, rocket sled track, vibration facility completed at TA-III.

1959 MWL (1959 to 1988) at TA-III for disposal of low-level mixed waste. Contained uranium (depleted, 

natural, and enriched), thorium, tritium beds, liquid scintillation cocktails, plutonium wastes.

1957 SNL given delivery responsibility for all training units (with high explosive), inert training units (no 

high explosive), and inert samples of weapons in early design stages until 1959. Assembly moved 

out of TA-II.

1960 TA-II converted to an explosive devices research and development area.

1960–1980s TTR used to test nonnuclear systems and components and underground nuclear experiments.

1960–1980s Kauai Test Facility, Hawaii, used for nonnuclear weapons testing. No radionuclides used.

1964 Component test facility for neutron generators added to TA-II. Potential radiation exposures 

included neutrons and tritium.

1967 Additional building added to TA-II for neutron generator (containing tritium) testing. 

1980s Operations commenced in TA-IV with pulsed power accelerators.

1985 Added a TA-V irradiation facility; north and south cells: gamma sources. 

1996 ACRR and SPR facilities with highly enriched uranium fuel online in TA-V. 

1996 HCF established to handle and examine radioactive materials (activation and fission products) 

from the SNL reactors and experiments. 

Table 2-2 lists each area and the notable facilities along with major radionuclides and years of 

operation. 

Table 2-2. Area information and parameters.

a

Area Process

Years of 

operation

Radionuclides

M = major, L = likely (particle 

AMAD = default 5 μm unless givenb)

Radiation types

B = beam types

TA-I Electron/ion beam accelerators

Cockroft-Walton(s)

(electron accelerator)

Late 1950s–

present

L = target dependent (e.g., H-3)

L = target dependent (e.g., H-3, DU, Be)

B = electron, X-ray

Van de Graaff(s)

(electron accelerator)

1958–present L = target dependent (e.g., H-3) B = electron, X-ray 

(400 keV–2 MeV)

Heavy Ion/Proton Accelerators 

~(100 kV)

1968+ NA B = protons, ions, 

and X-rays

Faxitron 1970–1973+ NA B = flash X-ray

Nerues 1971+ NA B = X-ray, gamma, 

bremsstrahlung

TA-I Manufacturing facilities

Chemistry laboratories 1949–present H-3 NA

LICA >2003–2005 M = Co-60, Cs-137 Gamma, beta

Radiation Standards Calibration 

Facility

1998–present M = Co-60, Cs-137 Gamma, beta

A machine shop ~1959–1994 M = DU, Pu NA

6-MeV ion generator 1984–1996 B = ion beam, 

X-rays

Sealed sources in microelectronics developmentSingle incident L= wide variety, but sealed. NA

Kaman Neutron Generators Mfg. 1996- Present Various neutron generator radionuclides Neutrons

TA-I Medical services

Medical X-ray (Picker GX-325-

PX350 tube) 300 mA @ 125 kVp

1949–present 

(registered in 

1978) 

NA X-ray
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Area Process

Years of 

operation

Radionuclides

M = major, L = likely (particle 

AMAD = default 5 μm unless givenb)

Radiation types

B = beam types

TA-II Repair/test services

TA-III Waste destruction, disposal, transfer

“Leaking Cask” Single incident Sr-90, Cs-137

Classified Destruction Facility 1957–1988 Transuranic elements, others

Radiological and mixed waste 

landfill

1949–1986 Cs-137, DU

TA-IV Electron/ion/X-ray beam accelerators

PBFA-I 1987–1995 N-13, O-15 gases B = electrons, 

gamma; 

target/neutrons 

Saturn (succeeded PBFA-I, 

preceded PBFA-II)

1967–1996+ N-13, O-15 gases B = electrons, 

gamma; 

target/neutrons 

PBFA-II 

(<30 MeV)

1993–present N-13, O-15 gases

(<0.042 Ci/yr N-13, 0.005 Ci/yr O-15)

B = electrons, 

gamma; 

target/neutrons 

Z-machine (modified from PBFA-II) 1996–present N-13, O-15 gases B = gamma, 

target/neutrons 

Mite pulsed X-ray (6 MeV), 4 units Present NA B = X-ray, gamma

Hermes III (higher energy version 

of Hermes II) (< 20 MeV)

~1988–1998 N-13, O-15 gases

(<2.3 & 0.03 Ci/yr)

B = gamma, 

target/neutrons 

SABRE; 6-12 MeV 1998–present N-13, O-15 gases

(<0.0058 Ci/yr, N-13.)

B = gamma, 

target/neutrons 

SPEED (1.0 MeV) 1983–1986 N-13, O-15 gases B = gamma, 

target/neutrons 

SPHINX 1992–present N-13, O-15 gases B = gamma, 

target/neutrons 

TESLA 1998– present NA B = X-rays

Neutron generator 1959–1997+ Various neutron generator radionuclides B= neutrons

TA-V Electron/ion beam accelerators

Febetrons 

(Flash X-ray system-< 2 MeV)

1967+ B = Flash X-ray 

(~2 MeV)

Proto – I (first-generation, highpowered, short-pulse accelerator)1972–-1976 Target dependent (e.g., H-3) B = electrons, 

X-rays

Proto – II (second-generation, 

high-powered, short-pulse 

accelerator)

1976–1998 Target dependent (e.g., H-3) B = electrons, 

X-rays

Pelletron Facility – variableenergy, high-stability dc electron 

beam generator (1 MeV)

1968–1991 Target dependent (e.g., H-3) B = electrons, 

X-rays

Hermes I, II (field emission 

electron beam or bremsstrahlung 

X-ray accelerator)

1968–1988 Target dependent (e.g., H-3) B = electrons, ions, 

bremsstrahlung 

X-ray

REBA = Z-machine (3.2 MV) 1994–present Target dependent (e.g., H-3) B = electrons, 

bremsstrahlung 

X-ray

Hydramite I & II – dual 

transmission line for high-energy 

short-pulse electron, 

bremsstrahlung X-ray unit (1 MeV)

1977, 1984+ Target dependent (e.g., H-3) B = electrons, ions, 

bremsstrahlung 

X-ray

REHYD = Heavy Ion Accelerator 

variable energy, positive ion 

combination of REBA and 

Hydramite (1.3 MeV)

1988–1998+ Target dependent (e.g., H-3) B = electrons, ions, 

bremsstrahlung 

X-ray

TA-V Reactors

SER (5-MW) Facility (~NPR@INL) 1958–1979 M = Ar-41;

L = All reactor-produced nuclides

B = prompt 

gamma, beta

ACRR (600 kW) also operated as 

Pulsed Reactor (ACPR) (<15,000 

MW) [TRIGA-type]

1968–1998 M = Ar-41

L = All reactor-produced nuclides

B = prompt 

gamma, beta
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 18 of 120

Area Process

Years of 

operation

Radionuclides

M = major, L = likely (particle 

AMAD = default 5 μm unless givenb)

Radiation types

B = beam types

ACRR = Mo-99 production 

(600 kW)

1998–present M = Ar-41, Tc-Mo-99

L = All reactor-produced nuclides

B = prompt 

gamma, 

SPR-I [GODIVA-type] 1961–1975 M = Ar-41

L = All reactor-produced nuclides

B = prompt 

gamma, beta

SPR-II [GODIVA-type]

(<130,000 MW peak)

>1962–

present

M = Ar-41

L = All reactor-produced nuclides

B = prompt 

gamma, beta

SPR-III [GODIVA-type] (< 170,000 

MW peak)

2003–present M = Ar-41

L = All reactor-produced nuclides

B = prompt 

gamma, beta

TA-V Irradiation and calibration support laboratories

HCF 1988–present Target dependent (e.g., metals from 

reactor, such as Cs-137, Ce-144, and 

H-3)

NA

GIF (~ 900 Ci Co-60 and similar 

Cs-137 high-activity sources for 

irradiation experiments)

1962–1998 M = Co-60 and Cs-137 NA

NEW GIF – combined with LICA = 

more flexible sources (Co-60 and 

Cs-137 high-activity sources for 

irradiation experiments)

1998?–

present 

M =Co-60 and Cs-137 NA

Coyote 

Test 

Facility

Thunder Range – source metals 

after explosion and fire tests of 

weapons, containers

1969–1994 M = DU NA

a. NA = not available.

b. AMAD = activity median aerodynamic diameter.

Table 2-3 lists specific areas of radionuclide use or radiation generation and the activity or power 

level.

Table 2-3. Magnitude of activitya

.

Area/facility Radionuclide/beam Activity/power level

TA-I

Applicable buildings Po-Be, Am-Be, Pu-Be, Co-60, 

DU

NA

Cockroft-Walton 100 keV +

Van de Graaff 400 keV–2 MeV

Heavy Ion Accelerators 100 kV

Faxitron X-ray NA

Nerues X-ray, gamma, bremsstrahlung NA

TA-II

Applicable buildings Explosives; nonnuclear NA

TA-III

Waste Facility NA

TA-IV

EBFA/PBFA I Prompt gamma, O-15, N-13 NA

PBFA II Prompt gamma, O-15, N-13 Maximum of 30 MeV with 5 shots/ wk; 

0.000168 Ci N-13 and 0.00002 Ci O-15 per 

shot (<0.042 Ci/yr N-13, 0.005 Ci/yr O-15).

Repetitive High Energy Pulsed 

Power I

X-rays

Hermes III Prompt gamma, O-15, N-13 2.32 Ci/yr of N-13 and 0.030 Ci/yr for O-15 

based on stack emissions. 20-MeV 

endpoint voltage

SABRE Prompt gamma, N-13 1.16E-05 Ci Ni-13 per shot/12 MeV 

maximum

Sub-Systems Test Facility Prompt gamma, O-15, N-13 NA

ALIAS Prompt gamma, O-15, N-13 NA
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 19 of 120

Area/facility Radionuclide/beam Activity/power level

TROLL Prompt gamma, O-15, N-13 NA

PI-112 Prompt gamma, O-15, N-13 NA

Saturn Prompt gamma, O-15, N-13, H-3 4.75 μCi H-3 between 10/29 and 11/09, 

1990. No H-3 between 1990 and 1996.

SPHINX Prompt gamma NA

Proto II (after late 1980s) NA

Mite X-ray, gamma 6 MeV

TA-V

Reactors Ar-41, Tc-99, Mo-99, reactorproduced nuclides600 kW (ACRR); <15,000 MW (ACPR); 

<130,000 MW (SPR-I); <170,000 MW 

(SPR-II)

HCF Co-60, Cs-137, mixed fission 

products

NA

GIF Co-60, Cs-137 NA

Hermes I, II Electrons, bremsstrahlung NA

REBA Electrons, bremsstrahlung 3.2 MV

Febetrons Flash X-ray < 2 MeV

Proto I, II Electrons, X-rays NA

Pelletron Electrons, X-rays 1 MeV

Hydramite Electrons, bremsstrahlung 1 MV

Rehyd Electrons, bremsstrahlung 1.3 MV

GIF Gamma sources Up to 900 Ci

TTR

Missile testing Related contamination NA

a. NA = not available or applicable.

2.4.1 TA-I

TA-I operations have been dedicated primarily to three activities: (1) the design, research, and 

development of weapon systems; (2) limited production of weapon system components; and 

(3) energy research programs. TA-I facilities include a Cockroft-Walton electron accelerator and a 

Van de Graaff generator as well as heavy ion/proton accelerators, a fexitron (an early flash X-ray 

machine), and another higher energy X-ray system with bremsstrahlung beam to irradiate electronic 

components. These systems often served as pilot versions for larger machines constructed in more 

remote areas (TAs-IV and -V).

Radiological Locations

Because TA-I was the area initially developed, it contains a large number of buildings of which at least 

75 have been routinely included in contamination survey programs over the last 20 years. Many of 

the buildings have laboratories that incidentally use small radionuclide quantities or that have ion 

beam or X-ray machines that support electronic and mechanical component manufacture. A few 

buildings have been devoted to neutron source manufacturing over a long period. 

A Van de Graaff generator and a Cockroft-Walton electron accelerator have been housed in the area 

from the late 1950s to the present. These generators were used to expose materials to large 

exposures of gamma and electron beam radiation during controlled experiments. The area also

housed the primary radiation calibration facilities used for primary calibration of all dosimetry meters 

on the SNL site over the last 50 years. These uses have continued to the present in TA-I. However, 

as noted above, the need for larger radiation sources such as research reactors, large accelerators, 

and hot cell facilities to support them for a much larger scale of radiation exposure experiments 

required the development of facilities in TAs-IV and -V.
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2.4.2 TA-II

TA-II is a 45-acre (180,000 m2) facility south of TA-I, established in 1948 for the assembly of chemical 

high-explosive main charges for nuclear weapons and later for production-scale assembly of nuclear 

weapons (Ullrich 1998). Assembly activities continued from 1952 to 1957, when the work shifted to 

other sites. Test devices and weapon prototypes continued to be assembled in TA-II for many years. 

In 1960, TA-II was converted to an explosive devices research and development area, which 

continued until 1995 when all explosive devices were removed from the area. 

Radiological Locations

In 1964, a component test facility was added to develop and test neutron generators, which continued 

over the years. In 1967, an explosive device quality laboratory was added. In general and based on 

review of the historical radiation protection program database, TA-II has had little use of radioactive 

materials.

2.4.3 TA-III

Remote facilities were built in TA-III 7 miles south of TA-II for full-scale testing of weapons with and 

without explosives. TA-III contains extensive design-test facilities such as rocket sled tracks, 

centrifuges, and a radiant heat facility (Fink 1999, pp. 45-46). Other facilities include a paper 

destructor, the Melting and Solidification Laboratory, the radioactive and low-level waste landfills, and 

the mixed waste landfill (MWL) that operated from 1959 to 1988 (Peace, Goering, and McVay 1996).

Radiological Locations

Other than small buildings housing the operators and screening detectors to manage the waste 

disposal facilities in TA-III, it does not appear from the reviewed records that any buildings in this area 

had significant sources of radioactivity present for normal operations. 

2.4.4 TA-IV

TA-IV was opened to provide specialized remote research areas for pulsed-power and high-energy 

experiments (SNL 1998a). It consists of several inertial-confinement fusion research and pulsedpower research facilities, including the High Energy Radiation Megavolt Electron Source (Hemres-III), 

the Z Facility, the Short Pulsed High Intensity Nanosecond X-Radiator (SPHINX) Facility, and the 

Saturn Accelerator (Ullrich 1998). 

Radiological Locations

There are several principal buildings that house the largest and most powerful accelerators at the SNL 

site. The machines housed in these buildings are identified in Table 2-2.

2.4.5 TA-V

TA-V is a highly secure remote research area housing experimental and engineering nuclear reactors 

and several electron beam accelerators. It contains the Annular Core Research Reactor (ACRR) and 

the SPR (in two reactor facilities), an intense Gamma Irradiation Facility (GIF) using gamma sources, 

and the Hot Cell Facility (HCF) (DOE 1996). The ACRR was built to enable operation in steady-state 

and pulsed modes; it was designated as the Annular Core Pulsed Reactor (ACPR) during the pulsedmode operational periods (SL 1979). After two decades of operation as a research reactor, the ACRR 

was transformed to a molybdenum-technetium radiopharmaceutical production unit for several years; 

it was idled in 2003. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 21 of 120

Radiological Locations

There are more than a dozen buildings in TA-V, which is primarily known for its research reactor and 

hot cell complex. Buildings housing the reactors, sources, and the HCF are the radiological facilities 

in this area.

2.4.6 Other Areas

SNL has test areas other than the five TAs listed above. These test areas, including Thunder Range 

in the canyons on the west side of the Manzanita Mountains, are known collectively as Coyote Test 

Field, which is southeast of TA-III. 

The Nevada Test Site (now the Nevada National Security Site; for convenience, this site profile refers 

to it as NTS) was used in 1957 for testing nuclear systems and bomb components in a limited series 

of aboveground tests (DOE 1997). After Russia ended the initial agreement on a nuclear test ban in 

1961, the United States reinitiated testing. All further testing occurred underground. A series of

underground nuclear tests was conducted between 1962 and 1973. Event-associated services were 

provided by the Environmental Sciences Department of Reynolds Electrical & Engineering Company. 

Health physics support was supplied by SNL (REECo 1974).

The TTR in Nevada was established in 1957 for testing nonnuclear systems and bomb components. 

Many of these tests involved components containing depleted uranium (DU); there is evidence of DU 

contamination at the site.

SNL conducted nonnuclear weapons testing in Hawaii at the naval facility on Kauai. The facility was 

used to launch missiles carrying experimental nonnuclear payloads (DOE 1997).

2.5 MAJOR SITE INCIDENTS

A single incident of “serious overexposure” at SNL occurred in 1960. The event involved misuse of

one of the new research Van de Graaff accelerators. A single employee received a life-threatening 

exposure and serious hand injuries when the interlock protocol intended to prevent such situations 

was contravened. Additional monitoring equipment was installed and revised protocols were created; 

no similar events have occurred. 

No other major site incidents were recorded at SNL. 

2.6 HEALTH PROTECTION PRACTICES

2.6.1 Personnel Monitoring

2.6.1.1 Badging

From 1948 to 1958, the exposure monitoring program used simple single-area film badges. Batches 

of these badges were periodically tested with a gridded table with pegs at specified distances from a 

central 226Ra source (traceable to the National Bureau of Standards). The data from readings of 

badges was manually recorded on 5- by 8-in. file cards. The format for recording data was eventually 

changed to punch cards so the data could be entered into a computerized database. In 1959, 

Eberline automated the readers and four-spot film badges like those used at Oak Ridge National 

Laboratory were introduced. About the same time, Victoreen R Chambers were introduced into the 

calibration process to allow the precision of the calibration exposures to be more readily determined. 

Before 1949, dosimetry was provided by LASL. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 22 of 120

In May 1971, thermoluminescent dosimeters (TLDs) replaced all film badges. Before then, several 

radiation field-mapping studies for new source areas continued to use both film badges and TLDs to 

compare the familiar form of exposure data with data from the more sensitive TLDs.

Section 6.0 provides further discussion of site dosimetry.

2.6.1.2 Area Monitoring

In 1960, area monitors were in use in some areas, but had not yet been fully installed in the work area 

adjacent to the Van de Graaff generator, where the serious overexposure incident occurred. After 

that incident, the system was widely expanded to serve as a warning system to supplement the badge 

program.

2.6.2 Access Control

SNL has changed over time from a laboratory where mostly classified activities took place to a

relatively open campus where outside researchers from industry, academia, and other national 

laboratories can gain access to the site’s facilities. In the early years, the majority of work was 

classified and access to facilities was strictly controlled by physical barriers. 

3.0 OCCUPATIONAL MEDICAL DOSE

3.1 INTRODUCTION

3.1.1 Purpose

This section presents information that can be used by a dose reconstructor to estimate the dose to a 

worker from occupational medical X-rays administered for screening and as a condition of 

employment at SNL.

3.1.2 Scope

This section provides specific information on documentation of historical medical X-ray practices at 

SNL.

3.2 EXAMINATION TYPES AND FREQUENCIES

As already mentioned in section 2.2, SNL had its origin as a satellite support site for the Los Alamos 

Scientific Laboratory [LASL; later renamed Los Alamos National Laboratory (LANL)] beginning in 

1945. It is reasonable to assume that the earliest Sandia workers may have been employed by LANL, 

and that their X-ray records reside there. Therefore, dose reconstructors should submit a request for 

additional medical records from LANL when SNL workers have verified employment before about 

1953, the earliest known X-ray equipment at SNL. If it appears from the requested records that the 

early X-rays were taken at LANL (before the worker reported for duty at Sandia), dose reconstructors 

should use the LANL doses. If it appears that the early X-rays were taken at Sandia (even though the 

worker was a LANL employee) then dose reconstructors should use the Sandia doses.

Detailed documentation of X-ray examination protocols is not available for SNL. Based on interviews 

with two former X-ray technologists and on review of employee medical records, a reasonably clear 

picture of the frequencies and types of radiographs taken at SNL over the years can be constructed 

(Stout 2005). Table 3-1 summarizes this information.

The table summarizes the current knowledge about the frequencies of the X-ray screening 

procedures from historical documents, interviews, and claim file records. Dose reconstructors should 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 23 of 120

consider the frequencies listed in Table 3-1 to be default values unless other documented data in the 

energy employee files are available.

From 1953 to 1985, SNL routinely performed preemployment or new-hire chest X-ray examinations, 

but it appears that not all new hires received these chest X-rays because the individual could choose 

to forego this portion of the examination. Further, employees transferring to SNL from other 

government agencies or contract facilities often brought their chest X-rays with them, so SNL did not 

repeat this procedure as a part of the new-hire examination (KES 2005). If X-ray examinations were 

performed in non-EEOICPA-covered facilities, the related dose cannot be included in dose 

reconstructions (ORAUT 2011a).

Table 3-1. Frequency and types of medical radiographs.

Period Type Frequency Comments

1943–1953 

(from LANL)

Singleprojection 

PFG 

(Shipman 

1955,1958)

Preemployment/new hire All workers.

Semiannual Workers handling uranium, plutonium, polonium, 

and beryllium oxide, and glass blowers (Hempleman 

1944; Grier and Hardy 1949)

Annual Workers exposed to radiation (Hempelmann 1944)

Termination All workers.

1953–1956 

(KES 2005)

Chest PFG 

and/or 14-

by 17-in. PA 

chest

Preemployment/new hire 

+ annual

Selected individuals could have received standard 

14- by 17-in. PA chest radiographs in addition to 

PFG. Employee could choose to forego a chest 

X-ray. Annual 14- by 17-in. chest X-ray was 

required for respirator users. X-ray records in the 

claim file should be used in determining the type and 

frequency of X-ray procedures. 

1957–1966 PA chest, 

14 by 17 in.

Preemployment/new hire 

+ annual

Employee could choose to forego chest X-ray. 

Annual 14- by 17-in. chest X-ray for respirator users.

1967–1974 PA chest, 

14 by 17 in.

Preemployment/new hire 

+ biennial; some annual

Employee could choose to forego chest X-ray. 

Annual 14- by 17-in. chest X-ray for respirator users.

1975–present PA chest, 

14 by 17 in.

Preemployment/new hire 

+ every 5 years; some 

annual

Employee could choose to forego chest X-ray. 

Annual 14- by 17-in. chest X-ray for respirator users.

1953–1985 Lumbar 

spine

Preemployment/new hire Four projections: AP; LAT; AP angle at S1 level, 

LAT spot. Performed on small percentage of 

workers, probably according to job classification.

From a review of employee X-ray folders, the standard procedure through 1956 was to use 4- by 5-in. 

chest photofluorography (PFG), with some employees receiving 14- by 17-in. posterior-anterior (PA) 

projections, perhaps in addition to the PFG (KES 2005). However, if an energy employee’s record 

shows a distance of 72 in., the exam was not a PFG and, therefore, a dose associated with a 14- by 

17-in. examination should be assigned. Selected employees also received lumbar spine radiographs. 

The lumbar spine radiographs are assumed to consist of four projections: an anterior-posterior (AP), 

a lateral (LAT), an AP angle at the S1 level, and an L5-S1 LAT spot film (KES 2005; ORAUT 2011b). 

Employee medical records indicate that SNL took PA chest films for most but not all employees on an 

annual or biennial frequency through 1966. In 1967, the frequency changed to biennial or triennial; to 

be favorable to claimants, this analysis assumed a biennial frequency. In 1975, the chest X-ray 

frequency changed to a 5-year cycle with the exception of a relatively small number of employees in 

special surveillance programs, and SNL discontinued lumbar spine radiographs for most job 

classifications. Employees with job classifications with higher risk for back injury, such as 

maintenance personnel and heavy-equipment operators, received lumbar spine X-rays and 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 24 of 120

employees whose job duties required qualification to wear a respirator received chest X-rays on an 

annual basis.

By about 1983, the number of chest X-rays taken at SNL was reduced by about 50%; they were 

limited based on medical history and job classification. Screening lumbar spine X-rays were 

discontinued in the mid-1980s. To be favorable to claimants, this analysis extended the date of this 

change through 1985 (as indicated in Table 3-1). In 1989, SNL again revised the physical 

examination protocol. Radiographs were limited to PA projections of the chest requested by a 

physician based on worker medical history and job classification (Clevenger 1990).

3.3 TECHNIQUE FACTORS AND INCIDENT AIR KERMA

3.3.1 Photofluorography, 1943-1956 (at LANL)

No documentation about PFG apparatus or techniques has been found for SNL. However, this 

imaging method was extensively used at LANL (ORAUT 2010d). X-ray records for early SNL 

employment should be requested from LANL. It appears that both the PFG and the 14- by 17-in.

chest films consisted of a single posterior-anterior (PA) projection in the earliest period at LANL 

(Shipman 1948). Dose reconstructors should be able to identify the PFG examinations in the claim 

file records when the examination occurred between 1943 and 1956, and when a “film number” is 

recorded, indicating the use of roll rather than sheet film. Dose reconstructors should assume a twoexposure stereo PFG only when the words “serial 1 and 2” appear on the record or radiologist’s 

interpretation. PFG at LANL ended in 1956 (Shipman 1958, p. 35).

Dose values for PFG at LANL are based on the default values for PFG in Dose Reconstruction from 

Occupational Medical X-Ray Procedures (ORAUT 2011b), except the doses are calculated for a 

single-exposure PFG (not stereo exposures) and reproduced in Table 3-5. A default incident air 

kerma value of 1.14 cGy for the single projection PFG (i.e. non-stereo) is used as the basis for the 

organ dose in this site profile (ORAUT 2011b).

3.3.2 Radiography, 1943 Through 1952 (at LANL)

Dose values for conventional 14- by 17-in. chest radiography for LANL from 1943 through 1952 are 

also based on the default values for chest radiography in Dose Reconstruction from Occupational 

Medical X-Ray Procedures (ORAUT 2011b).

3.3.3 Radiography, 1953 Through 1977 (at SNL)

The type of apparatus used before 1978 is undocumented. The official SNL history indicates that at 

least one radiographic X-ray machine went into service by 1953 (Johnson 1997). A 1977 radiological 

protection survey of the unit did not record machine details, referring to it only as “the medical X-ray 

unit” (Burnett 1977). The survey documented compliance with NCRP Report 33, which recommends 

a minimum total filter of 2.5 mm Al for operation above 70 kVp (NCRP 1968); the machine was 

presumably so equipped. The survey report noted a single deficiency (the light field projected by the 

adjustable collimator was not congruent with the actual X-ray field defined by the collimator) but did 

not mention the degree of incongruence. The report noted that this condition was found previously 

and, because “... no lasting remedy has been found,” recommended continued use of the collimator 

“because replacement of the complete unit is imminent”; the unit was replaced the following year.

Burnett (1977) included a tabulation of entrance skin exposure (ESE) for various types of 

examinations, including PA chest radiography and AP and LAT lumbar spine. Although stating these 

values were “... several percent low because no phantom was used in the measurements,” the survey 

report did not specify measurement technique and instrumentation; therefore, it is not known how the 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 25 of 120

measurements were made and it is not possible to determine if there were errors in measurement. 

However, the reported technique factors can be used to determine the incident air kerma. For a PA 

chest radiograph, the ESE is given as 5 mR based on a 70-kVp beam and an exposure of 10 mAs. 

For these technique factors, and assuming a total of 2.5-mm Al equivalent filtration, Table B.3 of 

NCRP Report 102 indicates an average output of 0.11 cGy/100 mAs for a 70-kVp beam at the sourceto-image distance (SID) of 183 cm (NCRP 1997). For a 10-mAs exposure, this equates to 0.011/

100 mAs at 183 cm. The incident air kerma is given by the equation

2 (0.11cGy/100mAs)(10mAs)(183/155) 0.0153cGy Ka,i = = (3-1)

where Ka,i is the incident air kerma in centigrays and (183/155)2 is the inverse square factor to convert 

intensity at the SID of 183 cm to intensity at the source-to-skin distance (SSD) of 155 cm (ORAUT 

2011b). 

The incident air kerma for the LAT chest is 2.5 times that of the PA chest (ORAUT 2011b), or 

0.0383 cGy.

For the AP lumbar spine projection, Burnett (1977) reports an ESE of 125 mR for an exposure of 

40 mAs at 75 kVp and a total filtration of 2.5 mm Al. The average output from Table B.3 of NCRP 

Report 102 is estimated at 0.45 cGy/100 mAs at 102 cm (NCRP 1997). Thus, as described above, 

the incident air kerma for the AP lumbar spine projection, based on a 75-kVp beam and an exposure 

of 40 mAs, assuming a total of 2.5-mm Al equivalent filtration, is 

2 (0.45 cGy/100 mAs 40 )( )( mAs 100/74) , or 0.329 cGy ,

where the SID is 102 cm and the SSD is 74 cm (ORAUT 2011b). For a LAT lumbar spine projection, 

Burnett (1977) indicated an ESE of 600 mR for an exposure of 150 mAs at 85 kVp. The average 

output from Table B.3 of NCRP Report 102 is estimated at 0.55 cGy/100 mAs at 100 cm (NCRP 

1997). Thus, as described above, the incident air kerma for the LAT lumbar spine projection, based 

on a 85-kVp beam and an exposure of 150 mAs, assuming a total of 2.5-mm Al equivalent filtration is 

(0.55 cGy/100 mAs)(150 mAs)(100/63)2

, or 2.07 cGy. 

AP and LAT lumbar spine spot film data are not available; the calculated incident air kerma values in 

Table 3-2 are estimates based on the AP and LAT lumbar spine projections (ORAUT 2011b).

Table 3-2. Machine settings and incident air kerma for screening 

radiography,a 1953 through 1977.

Procedure kVp mAs

Calculated incident 

air kerma (cGy)

PA chest 70 10 0.0153

LAT chest 70 10 0.0383

AP lumbar spine 75 40 0.328

LAT lumbar spine 85 150 2.08

AP angle lumbar spine (b) (b) 0.328

Lumbar spine spot film, LAT (b) (b) 2.08

a. Apply procedures per the guidance of Section 3.2.

b. No data are available and the values are estimated according to ORAUT 2011b.

3.3.4 Radiography, 1978 Through 1995

In 1978, the SNL Medical Department installed a new Picker GX 325 X-ray unit. This unit was 

equipped with a phototimer that was used for most chest X-rays and automatic collimation to limit the 

beam to the appropriate film size. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 26 of 120

A January 1991 study of skin doses on a phantom for various X-ray procedures reported results for 

shallow and deep doses (Sanderville 1991). Shallow dose (at 1 cm) was corrected to skin dose at the 

surface, including backscatter, using depth dose factors from NCRP Report 102 (NCRP 1989, Table 

B.8). This value was converted to incident air kerma using backscatter factors from NCRP Report 

102 (Table B.8). For the mid-PA chest, taken at 120 kVp with an exposure of 3.3 mAs, a shallow 

dose of 21 mrem was reported. Using a backscatter factor of 1.4 interpolated from NCRP Report 102, 

assuming 1 mrem = 1 mrad and using units of cGy (1 cGy = 1,000 mrad), results in an incident air 

kerma of 0.016 cGy. Table 3-3 lists similar values calculated for the LAT chest and lumbar spine 

projections using a backscatter factor of 1.35 along with the measured shallow dose values.

Table 3-3. Machine settings and incident air kerma for screening radiography,a 1978 through 1995.

Procedure kVp

HVL

(mm AL) mAs

Measured shallow 

doseb (mrem)

Calculated incident air 

kerma (cGy)

PA chest 120 3.5 3.3 21 0.016

LAT chest 120 3.5 10 64 0.048

AP lumbar spine 75 2.5 50 385 0.319

LAT lumbar spine 80 2.5 80 754 0.625c

AP lumbar spine spot (d) 2.5 (d) 385 0.319

LAT lumbar spine spot 90 2.5 80 954 0.790c

a. Apply procedures per the guidance of Section 3.2.

b. Includes backscatter.

c. The average of the LAT and the LAT spot were used in the dose calculations .

d. No data are available; same as AP lumbar spine (ORAUT 2011b).

3.3.5 Radiography, 1996 to Present

In 1996, SNL changed the X-ray apparatus and installed a Picker High Frequency (HF) unit. This 

machine was equipped with automatic exposure at 3 mAs that was used for most chest radiographs 

and automatic collimation to limit the beam size to that of the film used. Most exposures were made 

with the phototimer. Table 3-4 lists manual machine settings and corresponding calculated incident 

air kerma values.

Table 3-4. Manual settings and incident air kerma for screening medical radiography,a 1997 

to present.

Procedure KVp

HVL

(mm AL)

Measured 

ESE (mR)b

Exposure

(mAs)

SID 

(in.)

Calculated incident 

air kerma (cGy)

PA chest 80 4 13 3.3 72 0.0114

LAT chestc --- --- --- --- --- 0.0285

a. Apply procedures per the guidance of Section 3.2.

b. From Antonsen (1997).

c. ORAUT (2011b).

On March 13, 1997, this apparatus was evaluated as part of a comprehensive radiation protection 

survey performed by a radiation specialist with the U.S. Food and Drug Administration (Antonsen 

1997). According to the report, the X-ray unit met all applicable standards and specifications except 

that the acceptable standard total percent difference (length and width) between the X-ray field and 

light field alignment for the wall cassette of 4% was slightly exceeded (4.4%) when the automatic 

collimation was evaluated (Antonsen 1997). Correction was made 11 days after the survey report 

was received by SNL (Stout 1997). This small deviation should have had no significant impact on 

organ doses.

An in-depth radiographic unit performance summary by a Board-certified medical physicist, on May 

20, 2005, found that the unit was in compliance with the New Mexico Radiation Protection Code and 

met current performance recommendations of the American Association of Physicists in Medicine 

(Heintz 2005). Beam quality at 80 kVp was determined and a calculated half-value layer (HVL) of 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 27 of 120

3.71 mm Al was obtained. Thus, for dose reconstruction, an HVL of 4 mm Al was assumed in order to 

account for higher kVp techniques. Table 3-4 summarizes these for radiographic procedures and 

manual technique factors.

The incident air kerma is calculated from (ORAUT 2011b):

1 1 K (R)(2.58E-4C/kg R )(33.97J/C)(100cGy/Gy)( a,i 1Gy/1J kg ) − − = (3-2)

where 

Ka,i is the incident air kerma to be used in organ dose calculations in units of cGy in air; 

R is the exposure in roentgens at the skin entrance plane (i.e., ESE) 

2.58E-4 C/kg R-1 is the conversion factor from R to coulombs per kilogram

3.4 X-RAY DOSES TO WORKERS

3.4.1 Conversion of ESE to Dose

Organ and skin doses were determined according to methods described in ORAUT (2011b). 

3.4.2 Organ Doses from Chest Photofluorography

No documentation about PFG apparatus or techniques has been found for SNL. However, this 

imaging method was extensively used at LANL (ORAUT 2010) where early SNL workers were likely 

X-rayed. .Because no measurements or other information on PFG apparatus have been found, organ 

and skin doses were determined according to methods described in ORAUT (2011b). Because LANL 

primarily used the single (non-stereo) projection for PFG, the incident air kerma value used for PFG in 

this site profile is half that listed in ORAUT (2011b). Table 3-5 lists calculated organ doses for chest 

PFG and Table 3-6 lists calculated skin doses for chest PFG. 

3.4.3 Organ Doses from 14- by 17-in. PA Chest Radiography

Organ and skin doses were determined according to methods described in ORAUT (2011b). 

Table 3-5 provides organ dose equivalents for chest projections for all periods. Tables 3-6, 3-7, and 

3-8 provide skin dose guidance and skin dose equivalents for chest projections. In addition to PA and 

LAT projection doses, oblique projection and right anterior oblique (RAO) projection doses are 

provided. Further detail on the various X-ray projections and derivation of doses is provided in 

ORAUT (2011b). 

Entrance skin dose (ENSD) is determined by multiplying the incident air kerma by the backscatter 

factors of 1.35 and 1.4 for HVL of 2.5 and 3.5 mm Al, respectively, from NCRP Report 102 (NCRP 

1997, Table B.8). Skin doses for all areas of skin are provided in Tables 3-5 through 3-11.

3.4.4 Organ Doses from Lumbar Spine Radiography

Organ and skin doses were determined according to methods described in ORAUT (2011b). 

Table 3-9 provides organ dose equivalents for lumbar spine projections through 1985. Tables 3-10 

and 3-11 provide skin dose guidance and skin dose equivalents for lumbar spine projections. Further 

detail on the various X-ray projections and derivation of doses is provided in ORAUT (2011b).
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 28 of 120

3.5 UNCERTAINTY ANALYSIS

ORAUT-OTIB-0006 (ORAUT 2011b) lists the major sources of uncertainty in X-ray output intensity 

and subsequent effect on dose to the worker. The five sources of uncertainty are: 

1. X-ray beam measurement error (±2%),

2. Variation in peak kilovoltage (±9%),

3. Variation in X-ray beam current (±5%),

4. Variation in exposure time (±25%), and

5. Variation in SSD as a result of worker size (±10%). 

The 10% uncertainty in output intensity as a result of worker size was based on an inverse square 

correction of output intensity changes from differences of standard chest thickness of ±7.5 cm.

These uncertainties are assumed to be random; therefore, the combined statistical uncertainty was 

calculated as the square root of the sum of the squares of all the uncertainties, which is ±28.9%. 

Rounding this up to ±30% provides an adequate and suitably conservative indication of uncertainty. 

Therefore, for a derived dose equivalent to an individual organ, a total combined standard uncertainty 

of ±30% can be assumed. Dose reconstructors should, therefore, input the organ dose equivalent as 

the mean of a normal distribution with a standard uncertainty of ±30%.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 29 of 120

Table 3-5. Organ dose equivalents (rem) for chest projections for all periods.

Organ Projection

PFG (at 

LANL),

1943–1952a

14- by 17-in.

(at LANL),

1943–1952

14- by 17-in.,

1953–1977

14- by 17-in.,

1978–1995

14- by 17-in.,

1996–

present

Thyroid PA 1.97E–01 3.48E-02 2.66E-03 9.92E-04 8.89E-04

LAT/OBL -- -- 5.25E-03 7.25E-03 4.67E-03

Eye/brain PA 3.63E–02 6.40E-03 4.90E-4 9.92E-04 8.89E-04

LAT/OBL -- -- 5.25E-03 7.25E-03 4.67E-03

Ovaries PA 2.50E–02 2.50E-02 2.57E-03b 5.12E-05 5.93E-05

LAT/OBL -- -- 2.18E-03b 7.68E-05 7.13E-05

Urinary

bladder/prostate

PA 2.50E–02 2.50E-02 2.57E-03b 5.12E-05 5.93E-05

LAT/OBL -- -- 2.18E-03b 7.68E-05 7.13E-05

Colon/rectum PA 2.50E–02 2.50E-02 2.57E-03b 5.12E-05 5.93E-05

LAT/OBL -- -- 2.18E-03b 7.68E-05 7.13E-05

Testes PA 5.00E–03 5.00E-03 2.14E-04b 1.60E-07 1.14E-07

LAT/OBL -- -- 1.15E-04b 4.80E-06 2.85E-06

Lungs (male) PA 4.75E–01 8.38E-02 6.41E-03 9.04E-03 7.16E-03

LAT/OBL -- -- 7.39E-03 1.32E-02 8.92E-03

Lungs (female) PA 5.11E-01 9.02E-02 6.90E-03 9.76E-03 7.68E-03

LAT/OBL -- -- 8.43E-03 1.49E-02 1.00E-02

Thymus PA 5.11E-01 9.02E-02 6.90E-03 9.76E-03 7.68E-03

LAT/OBL -- -- 8.43E-03 1.49E-02 1.00E-02

Esophagus PA 5.11E-01 9.02E-02 6.90E-03 9.76E-03 7.68E-03

LAT/OBL -- -- 8.43E-03 1.49E-02 1.00E-02

Stomach PA 5.11E-01 9.02E-02 6.90E-03 9.76E-03 7.68E-03

LAT/OBL -- -- 8.43E-03 1.49E-02 1.00E-02

Bone surface PA 5.11E-01 9.02E-02 6.90E-03 9.76E-03 7.68E-03

LAT/OBL -- -- 8.43E-03 1.49E-02 1.00E-02

Liver/gall 

bladder/spleen/

pancreas

PA 5.11E-01 9.02E-02 6.90E-03 9.76E-03 7.68E-03

LAT/OBL -- -- 8.43E-03 1.49E-02 1.00E-02

Remainder 

organs

PA 5.11E-01 9.02E-02 6.90E-03 9.76E-03 7.68E-03

LAT/OBL -- -- 8.43E-03 1.49E-02 1.00E-02

Breast PA 5.56E–02 9.80E-03 7.50E-04 1.46E-03 1.32E-03

LAT/OBL -- -- 9.77E-03 1.52E-02 9.78E-03

Uterus PA 2.50E–02 2.50E-02 2.28E-03b 4.80E-05 5.93E-05

LAT/OBL -- -- 1.65E-03b 6.72E-05 5.99E-05

Bone marrow 

(male)

PA 1.04E–01 1.84E-02 1.41E-03 2.34E-03 2.03E-03

LAT/OBL -- -- 1.42E-03 2.93E-03 2.17E-03

Bone marrow 

(female)

PA 9.75E–02 1.72E-02 1.32E-03 2.26E-03 1.96E-03

LAT/OBL -- -- 1.11E-03 2.30E-03 1.68E-03

Entrance skinc PA 1.53E+00 2.70E-01 2.07E-02 2.24E-02 1.60E-02

LAT/OBL -- -- 5.17E-02 6.72E-02 3.99E-02

a. PFG presumed to have been done at LANL. Doses for are from ORAUT-OTIB-0006 (ORAUT 2011b) and are for 

non-stereo projections.

b. DCF for the PA abdomen was used to determine dose.

c. Using method described in ORAUT-OTIB-0006 (ORAUT 2011b). 
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Table 3-6. Skin dose guidance and skin dose equivalents (rem) for chest projections, 1953 through 1977.

Area of skin

PFG 1943–1952 PA chest LAT chest 1953–1977 RAO chest 1953–1977

Guidance Dose Guidance

1943–1952 

Dose

1953–1977

Dose Guidance Dose Guidance Dose

Right front shoulder EXSD 3.36E–02 EXSD 5.9E-03 5.E-04 ENSD 5.17E-02 EXSD 2.E-04

Right back shoulder ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 ENSD 5.17E-02 ENSD 5.17E-02

Left front shoulder EXSD 3.36E–02 EXSD 5.9E-03 5.E-04 EXSD 2.E-04 EXSD 2.E-04

Left back shoulder ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 EXSD 2.E-04 ENSD 5.17E-02

Right upper arm to 

elbow

10% ENSD 1.53E–01 ENSD 2.70E-01 2.07E-02 ENSD 5.17E-02 ENSD 5.17E-02

Left upper arm to 

elbow

10% ENSD 1.53E–01 ENSD 2.70E-01 2.07E-02 EXSD 2.E-04 ENSD 5.17E-02

Left hand ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 10% ENSD 5.2E-03 10% ENSD 5.2E-03

Right hand ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 10% ENSD 5.2E-03 10% ENSD 5.2E-03

Left elbow, forearm, 

wrist

10% ENSD 1.53E–01 ENSD 2.70E-01 2.07E-02 10% ENSD 5.2E-03 10% ENSD 5.2E-03

Right elbow, forearm, 

wrist

10% ENSD 1.53E–01 ENSD 2.70E-01 2.07E-02 10% ENSD 5.2E-03 10% ENSD 5.2E-03

Right side of head 

(including ear and 

temple)

10% ENSD 1.53E–01 10% ENSD 2.70E-02 2.1E-03 Eye/brain 5.2E-03 10% EXSD 2.E-05

Left side of head 

(including ear and 

temple)

10% ENSD 1.53E–01 10% ENSD 2.70E-02 2.1E-03 Eye/brain 5.2E-03 10% ENSD 5.2E-03

Front left thigh RSD (0.52 m) 4.E–04 RSD (0.52 m) 8.E-05 6.E-06 RSD (0.52 m) 8.E-06 RSD (0.52 m) 8.E-06

Back left thigh RSD (0.52 m) 4.E–04 RSD (0.52 m) 8.E-05 6.E-06 RSD (0.52 m) 8.E-06 RSD (0.52 m) 8.E-06

Front right thigh RSD (0.52 m) 4.E–04 RSD (0.52 m) 8.E-05 6.E-06 RSD (0.52 m) 8.E-06 RSD (0.52 m) 8.E-06

Back right thigh RSD (0.52 m) 4.E–04 RSD (0.52 m) 8.E-05 6.E-06 RSD (0.52 m) 8.E-06 RSD (0.52 m) 8.E-06

Left knee and below RSD (0.86 m) 2.E–04 RSD (0.86 m) 3.E-05 2.E-06 RSD (0.86 m) 3.E-06 RSD (0.86 m) 3.E-06

Right knee and below RSD (0.86 m) 2.E–04 RSD (0.86 m) 3.E-05 2.E-06 RSD (0.86 m) 3.E-06 RSD (0.86 m) 3.E-06

Left side of face Eye/brain 3.63E–02 Eye/brain 6.4E-03 5.E-04 Eye/brain 5.2E-03 ENSD 5.17E-02

Right side of face Eye/brain 3.63E–02 Eye/brain 6.4E-03 5.E-04 Eye/brain 5.2E-03 EXSD 2.E-04

Left side of neck 10% ENSD 1.53E–01 ENSD 2.70E-01 2.07E-02 Eye/brain 5.2E-03 ENSD 5.17E-02

Right side of neck 10% ENSD 1.53E–01 ENSD 2.70E-01 2.07E-02 Eye/brain 5.2E-03 EXSD 2.E-04

Back of head 10% ENSD 1.53E–01 10% ENSD 2.70E-02 2.1E-03 Eye/brain 5.2E-03 10% ENSD 5.2E-03

Front of neck Eye/brain 3.63E–02 Eye/brain 6.4E-03 5.E-04 Eye/brain 5.2E-03 Eye/Brain 5.2E-03

Back of neck 10% ENSD 1.53E–01 ENSD 2.70E-01 2.07E-02 Eye/brain 5.2E-03 ENSD 5.17E-02

Front torso: base of 

neck to end of 

sternum

EXSD 3.36E–02 EXSD 5.9E-03 5.E-04 Lung 8.4E-03 EXSD 2.E-04
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Area of skin

PFG 1943–1952 PA chest LAT chest 1953–1977 RAO chest 1953–1977

Guidance Dose Guidance

1943–1952 

Dose

1953–1977

Dose Guidance Dose Guidance Dose

Front torso: end of 

sternum to lowest rib

EXSD 3.36E–02 EXSD 5.9E-03 5.E-04 Lung 8.4E-03 EXSD 2.E-04

Front torso: lowest 

rib to iliac crest

EXSD 3.36E–02 EXSD 5.9E-03 5.E-04 Lung 8.4E-03 EXSD 2.E-04

Front torso: iliac 

crest to pubis

10% EXSD 3.4E–03 10% EXSD 6.E-04 5.E-05 10% lung 8.E-04 10% EXSD 2.E-05

Back torso: base of 

neck to mid-back

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 Lung 8.4E-03 ENSD 5.17E-02

Back torso: midback to lowest rib

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 Lung 8.4E-03 ENSD 5.17E-02

Back torso: lowest 

rib to iliac crest

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 Lung 8.4E-03 ENSD 5.17E-02

Back torso: buttocks 

(Iliac crest and below)

10% ENSD 1.53E–01 10% ENSD 2.70E-02 2.1E-03 10% Lung 8.E-04 10% ENSD 5.2E-03

Right torso: base of 

neck to end of 

sternum

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 ENSD 5.17E-02 EXSD 2.E-04

Right torso: end of 

sternum to lowest rib

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 ENSD 5.17E-02 EXSD 2.E-04

Right torso: lowest 

rib to iliac crest

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 ENSD 5.17E-02 EXSD 2.E-04

Right torso: iliac 

crest to pubis (right 

hip)

10% ENSD 1.53E–01 10% ENSD 2.70E-02 2.1E-03 10% ENSD 5.2E-03 10% EXSD 2.E-05

Left torso: base of 

neck to end of 

sternum

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 EXSD 2.E-04 ENSD 5.17E-02

Left torso: end of 

sternum to lowest rib

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 EXSD 2.E-04 ENSD 5.17E-02

Left torso: lowest rib 

to iliac crest

ENSD 1.53E+00 ENSD 2.70E-01 2.07E-02 EXSD 2.E-04 ENSD 5.17E-02

Left torso: iliac crest 

to pubis (left hip)

10% ENSD 1.53E–01 10% ENSD 2.70E-02 2.1E-03 10% EXSD 2.E-05 10% ENSD 5.2E-03
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Table 3-7. Skin dose guidance and skin dose equivalents (rem) for chest projections, 1978 through 1995.

Area of skin

PA chest, 1978–1995 LAT chest, 1978–1995 RAO chest, 1978–1995

Guidance Dose Guidance Dose Guidance Dose

Right front shoulder EXSD 6.E-04 ENSD 6.72E-02 EXSD 4.E-04

Right back shoulder ENSD 2.24E-02 ENSD 6.72E-02 ENSD 6.72E-02

Left front shoulder EXSD 6.E-04 EXSD 4.E-04 EXSD 4.E-04

Left back shoulder ENSD 2.24E-02 EXSD 4.E-04 ENSD 6.72E-02

Right upper arm to elbow 10% ENSD 2.2E-03 ENSD 6.72E-02 10% ENSD 6.7E-03

Left upper arm to elbow 10% ENSD 2.2E-03 EXSD 4.E-04 10% ENSD 6.7E-03

Left hand 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Right hand 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Left elbow, forearm, wrist 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Right elbow, forearm, wrist 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Right side of head (including ear and temple) 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% EXSD 4.E-05

Left side of head (including ear and temple) 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Front left thigh RSD (0.52 m) 8.E-06 RSD (0.52 m) 1.E-05 RSD (0.52 m) 1.E-05

Back left thigh RSD (0.52 m) 8.E-06 RSD (0.52 m) 1.E-05 RSD (0.52 m) 1.E-05

Front right thigh RSD (0.52 m) 8.E-06 RSD (0.52 m) 1.E-05 RSD (0.52 m) 1.E-05

Back right thigh RSD (0.52 m) 8.E-06 RSD (0.52 m) 1.E-05 RSD (0.52 m) 1.E-05

Left knee and below RSD (0.86 m) 3.E-06 RSD (0.86 m) 4.E-06 RSD (0.86 m) 4.E-06

Right knee and below RSD (0.86 m) 3.E-06 RSD (0.86 m) 4.E-06 RSD (0.86 m) 4.E-06

Left side of face Eye/brain 9.9E-04 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Right side of face Eye/brain 9.9E-04 10% ENSD 6.7E-03 10% EXSD 4.E-05

Left side of neck 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Right side of neck 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% EXSD 4.E-05

Back of head 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Front of neck Thyroid 9.9E-04 10% ENSD 6.7E-03 Thyroid 7.25E-03

Back of neck 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% ENSD 6.7E-03

Front torso: base of neck to end of sternum EXSD 6.E-04 Lung 1.49E-02 EXSD 4.E-04

Front torso: end of sternum to lowest rib EXSD 6.E-04 Lung 1.49E-02 EXSD 4.E-04

Front torso: lowest rib to iliac crest 10% EXSD 6.E-05 10% Lung 1.5E-03 10% EXSD 4.E-05

Front torso: iliac crest to pubis 10% EXSD 6.E-05 10% Lung 1.5E-03 10% EXSD 4.E-05

Back torso: base of neck to mid-back ENSD 2.24E-02 Lung 1.49E-02 ENSD 6.72E-02

Back torso: mid-back to lowest rib ENSD 2.24E-02 Lung 1.49E-02 ENSD 6.72E-02

Back torso: lowest rib to iliac crest 10% ENSD 2.2E-03 10% lung 1.5E-03 10% ENSD 6.7E-03

Back torso: buttocks (Iliac crest and below) 10% ENSD 2.2E-03 10% lung 1.5E-03 10% ENSD 6.7E-03

Right torso: base of neck to end of sternum ENSD 2.24E-02 ENSD 6.72E-02 EXSD 4.E-04

Right torso: end of sternum to lowest rib ENSD 2.24E-02 ENSD 6.72E-02 EXSD 4.E-04

Right torso: lowest rib to iliac crest 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% EXSD 4.E-05

Right torso: iliac crest to pubis (right hip) 10% ENSD 2.2E-03 10% ENSD 6.7E-03 10% EXSD 4.E-05

Left torso: base of neck to end of sternum ENSD 2.24E-02 EXSD 4.E-04 ENSD 6.72E-02
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Area of skin

PA chest, 1978–1995 LAT chest, 1978–1995 RAO chest, 1978–1995

Guidance Dose Guidance Dose Guidance Dose

Left torso: end of sternum to lowest rib ENSD 2.24E-02 EXSD 4.E-04 ENSD 6.72E-02

Left torso: lowest rib to iliac crest 10% ENSD 2.2E-03 10% EXSD 4.E-05 10% ENSD 6.7E-03

Left torso: iliac crest to pubis (left hip) 10% ENSD 2.2E-03 10% EXSD 4.E-05 10% ENSD 6.7E-03

Table 3-8. Skin dose guidance and skin dose equivalents (rem) for chest projections, 1996 through present.

Area of skin

PA chest, 1996–present LAT chest, 1996–present RAO chest, 1996–

present

Guidance Dose Guidance Dose Guidance Dose

Right front shoulder EXSD 5.E-04 ENSD 3.99E-02 EXSD 3.E-04

Right back shoulder ENSD 1.60E-02 ENSD 3.99E-02 ENSD 3.99E-02

Left front shoulder EXSD 5.E-04 EXSD 3.E-04 EXSD 3.E-04

Left back shoulder ENSD 1.60E-02 EXSD 3.E-04 ENSD 3.99E-02

Right upper arm to elbow 10% ENSD 1.6E-03 ENSD 3.99E-02 10% ENSD 4.0E-03

Left upper arm to elbow 10% ENSD 1.6E-03 EXSD 3.E-04 10% ENSD 4.0E-03

Left hand 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Right hand 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Left elbow, forearm, wrist 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Right elbow, forearm, wrist 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Right side of head (including ear and temple) 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% EXSD 3.E-05

Left side of head (including ear and temple) 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Front left thigh RSD (0.52 m) 6.E-06 RSD (0.52 m) 8.E-06 RSD (0.52 m) 8.E-06

Back left thigh RSD (0.52 m) 6.E-06 RSD (0.52 m) 8.E-06 RSD (0.52 m) 8.E-06

Front right thigh RSD (0.52 m) 6.E-06 RSD (0.52 m) 8.E-06 RSD (0.52 m) 8.E-06

Back right thigh RSD (0.52 m) 6.E-06 RSD (0.52 m) 8.E-06 RSD (0.52 m) 8.E-06

Left knee and below RSD (0.86 m) 2.E-06 RSD (0.86 m) 3.E-06 RSD (0.86 m) 3.E-06

Right knee and below RSD (0.86 m) 2.E-06 RSD (0.86 m) 3.E-06 RSD (0.86 m) 3.E-06

Left side of face Eye/brain 9.E-04 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Right side of face Eye/brain 9.E-04 10% ENSD 4.0E-03 10% EXSD 3.E-05

Left side of neck 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Right side of neck 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% EXSD 3.E-05

Back of head 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Front of neck Thyroid 9.E-04 10% ENSD 4.0E-03 Thyroid 4.7E-03

Back of neck 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% ENSD 4.0E-03

Front torso: base of neck to end of sternum EXSD 5.E-04 Lung 1.00E-02 EXSD 3.E-04

Front torso: end of sternum to lowest rib EXSD 5.E-04 Lung 1.00E-02 EXSD 3.E-04

Front torso: lowest rib to iliac crest 10% EXSD 5.E-05 10% lung 1.0E-03 10% EXSD 3.E-05

Front torso: iliac crest to pubis 10% EXSD 5.E-05 10% lung 1.0E-03 10% EXSD 3.E-05

Back torso: base of neck to mid-back ENSD 1.60E-02 Lung 1.00E-02 ENSD 3.99E-02

Back torso: mid-back to lowest rib ENSD 1.60E-02 Lung 1.00E-02 ENSD 3.99E-02
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Area of skin

PA chest, 1996–present LAT chest, 1996–present RAO chest, 1996–

present

Guidance Dose Guidance Dose Guidance Dose

Back torso: lowest rib to iliac crest 10% ENSD 1.6E-03 10% lung 1.0E-03 10% ENSD 4.0E-03

Back torso: buttocks (Iliac crest and below) 10% ENSD 1.6E-03 10% lung 1.0E-03 10% ENSD 4.0E-03

Right torso: base of neck to end of sternum ENSD 1.60E-02 ENSD 3.99E-02 EXSD 3.E-04

Right torso: end of sternum to lowest rib ENSD 1.60E-02 ENSD 3.99E-02 EXSD 3.E-04

Right torso: lowest rib to iliac crest 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% EXSD 3.E-05

Right torso: iliac crest to pubis (right hip) 10% ENSD 1.6E-03 10% ENSD 4.0E-03 10% EXSD 3.E-05

Left torso: base of neck to end of sternum ENSD 1.60E-02 EXSD 3.E-04 ENSD 3.99E-02

Left torso: end of sternum to lowest rib ENSD 1.60E-02 EXSD 3.E-04 ENSD 3.99E-02

Left torso: lowest rib to iliac crest 10% ENSD 1.6E-03 10% EXSD 3.E-05 10% ENSD 4.0E-03

Left torso: iliac crest to pubis (left hip) 10% ENSD 1.6E-03 10% EXSD 3.E-05 10% ENSD 4.0E-03
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 35 of 120

Table 3-9. Organ dose equivalents (rem) for lumbar spine projections for all periods.

Organ

AP and AP spot 

lumbar spine 

projections,

1953–1977a

LAT and LAT 

spot lumbar 

spine projections,

1953–1977b

AP and AP spot 

lumbar spine 

projections,

1978–1985a

LAT and LAT 

spot lumbar 

spine projections,

1978–1985b

Thyroid 1.97E-04 4.16E-05 1.91E-04 1.42E-05

Eye/brain 1.97E-04 4.16E-05 1.91E-04 1.42E-05

Ovaries 1.42E-01 1.95E-01 1.38E-01 6.66E-02

Urinary/bladder/prostate 1.42E-01 1.95E-01 1.38E-01 6.66E-02

Colon/rectum 1.42E-01 1.95E-01 1.38E-01 6.66E-02

Testes 2.76E-03 3.33E-03 2.68E-03 1.13E-03

Lungs male 5.19E-02 5.82E-02 5.04E-02 1.98E-02

Lungs female 5.19E-02 5.82E-02 5.04E-02 1.98E-02

Thymus 5.19E-02 5.82E-02 5.04E-02 1.98E-02

Esophagus 5.19E-02 5.82E-02 5.04E-02 1.98E-02

Stomach 1.42E-01 1.95E-01 1.38E-01 6.66E-02

Bone surfaces 1.42E-01 1.95E-01 1.38E-01 6.66E-02

Liver/gall bladder/spleen 1.42E-01 1.95E-01 1.38E-01 6.66E-02

Remainder 1.42E-01 1.95E-01 1.38E-01 6.66E-02

Breast 1.11E-03 4.22E-03 5.08E-04 6.38E-04

Uterus 1.89E-01 1.29E-01 1.83E-01 4.39E-02

Bone marrow male 2.43E-02 9.15E-02 2.36E-02 3.12E-02

Bone marrow female 2.43E-02 9.15E-02 2.36E-02 3.12E-02

Entrance skinc 8.87E-01 5.61E+00 8.61E-01 1.71E+00

a. The dose value is for two AP projections, the AP and AP spot for S-1.

b. The dose value is for two LAT projections, the LAT and LAT spot for L-4/5.

c. Using method described in ORAUT-OTIB-0006 (ORAUT 2011b).

Table 3-10. Skin dose guidance and skin dose equivalents (rem) for lumbar spine projections, 1953 

through 1977.

Area of skin

AP and AP spot lumbar 

spinea, 1953–1977

LAT and LAT spot lumbar 

spinea, 1953–1977

Guidance Dose Guidance Dose

Right front shoulder 10% ENSD 8.87E-02 10% ENSD 5.61E-01

Right back shoulder 10% EXSD 1.9E-03 10% ENSD 5.61E-01

Left front shoulder 10% ENSD 8.87E-02 10% EXSD 2.5E-03

Left back shoulder 10% EXSD 1.9E-03 10% EXSD 2.5E-03

Right upper arm to elbow 10% ENSD 8.87E-02 10% ENSD 5.61E-01

Left upper arm to elbow 10% ENSD 8.87E-02 10% EXSD 2.5E-03

Left hand ENSD 8.87E-01 10% EXSD 2.5E-03

Right hand ENSD 8.87E-01 10% ENSD 5.61E-01

Left elbow, forearm, wrist ENSD 8.87E-01 10% EXSD 2.5E-03

Right elbow, forearm, wrist ENSD 8.87E-01 10% ENSD 5.61E-01

Right side of head (including ear and temple) Eye/brain 2.E-04 Eye/brain 4.E-05

Left side of head (including ear and temple) Eye/brain 2.E-04 Eye/brain 4.E-05

Front left thigh 10% ENSD 8.87E-02 10% EXSD 2.5E-03

Back left thigh 10% EXSD 1.9E-03 10% EXSD 2.5E-03

Front right thigh 10% ENSD 8.87E-02 10% ENSD 5.61E-01

Back right thigh 10% EXSD 1.9E-03 10% ENSD 5.61E-01

Left knee and below RSD (0.60 m) 2.E-04 RSD (0.60 m) 6.E-04

Right knee and below RSD (0.60 m) 2.E-04 RSD (0.60 m) 6.E-04

Left side of face Eye/brain 2.E-04 Eye/brain 4.E-05

Right side of face Eye/brain 2.E-04 Eye/brain 4.E-05

Left side of neck Eye/brain 2.E-04 Eye/brain 4.E-05

Right side of neck Eye/brain 2.E-04 Eye/brain 4.E-05

Back of head Eye/brain 2.E-04 Eye/brain 4.E-05
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 36 of 120

Area of skin

AP and AP spot lumbar 

spinea, 1953–1977

LAT and LAT spot lumbar 

spinea, 1953–1977

Guidance Dose Guidance Dose

Front of neck Eye/brain 2.E-04 Eye/brain 4.E-05

Back of neck Eye/brain 2.E-04 Eye/brain 4.E-05

Front torso: base of neck to end of sternum 10% ENSD 8.87E-02 Lung 5.82E-02

Front torso: end of sternum to lowest rib ENSD 8.87E-01 Lung 5.82E-02

Front torso: lowest rib to iliac crest ENSD 8.87E-01 Lung 5.82E-02

Front torso: iliac crest to pubis ENSD 8.87E-01 Lung 5.82E-02

Back torso: base of neck to mid-back 10% EXSD 1.9E-03 Lung 5.82E-02

Back torso: mid-back to lowest rib EXSD 1.93E-02 Lung 5.82E-02

Back torso: lowest rib to iliac crest EXSD 1.93E-02 Lung 5.82E-02

Back torso: buttocks (Iliac crest and below) EXSD 1.93E-02 Lung 5.82E-02

Right torso: base of neck to end of sternum 10% ENSD 8.87E-02 10% ENSD 5.61E-01

Right torso: end of sternum to lowest rib ENSD 8.87E-01 ENSD 5.61E+00

Right torso: lowest rib to iliac crest ENSD 8.87E-01 ENSD 5.61E+00

Right torso: iliac crest to pubis (right hip) ENSD 8.87E-01 ENSD 5.61E+00

Left torso: base of neck to end of sternum 10% ENSD 8.87E-02 10% EXSD 2.5E-03

Left torso: end of sternum to lowest rib ENSD 8.87E-01 EXSD 2.46E-02

Left torso: lowest rib to iliac crest ENSD 8.87E-01 EXSD 2.46E-02

Left torso: iliac crest to pubis (left hip) ENSD 8.87E-01 EXSD 2.46E-02

a. The organ doses listed are for two projections summed together, the AP and AP spot, and the LAT and LAT spot.

Table 3-11. Skin dose guidance and skin dose equivalents (rem) for lumbar spine projections, 1978 

through 1985.

Area of skin

AP and AP spot lumbar 

spinea, 1978–1985

LAT and LAT spot lumbar 

spinea, 1978–1985

Guidance Dose Guidance Dose

Right front shoulder 10% ENSD 8.61E-02 10% ENSD 1.71E-01

Right back shoulder 10% EXSD 1.9E-03 10% ENSD 1.71E-01

Left front shoulder 10% ENSD 8.61E-02 10% EXSD 8.E-04

Left back shoulder 10% EXSD 1.9E-03 10% EXSD 8.E-04

Right upper arm to elbow 10% ENSD 8.61E-02 10% ENSD 1.71E-01

Left upper arm to elbow 10% ENSD 8.61E-02 10% EXSD 8.E-04

Left hand 10% ENSD 8.61E-02 10% EXSD 8.E-04

Right hand 10% ENSD 8.61E-02 10% ENSD 1.71E-01

Left elbow, forearm, wrist 10% ENSD 8.61E-02 10% EXSD 8.E-04

Right elbow, forearm, wrist 10% ENSD 8.61E-02 10% ENSD 1.71E-01

Right side of head (including ear and temple) Eye/brain 2.E-04 Eye/brain 1.E-05

Left side of head (including ear and temple) Eye/brain 2.E-04 Eye/brain 1.E-05

Front left thigh 10% ENSD 8.61E-02 10% EXSD 8.E-04

Back left thigh 10% EXSD 1.9E-03 10% EXSD 8.E-04

Front right thigh 10% ENSD 8.61E-02 10% ENSD 1.71E-01

Back right thigh 10% EXSD 1.9E-03 10% ENSD 1.71E-01

Left knee and below RSD (0.60 m) 2.E-04 RSD (0.60 m) 2.E-04

Right knee and below RSD (0.60 m) 2.E-04 RSD (0.60 m) 2.E-04

Left side of face Eye/brain 2.E-04 Eye/brain 1.E-05

Right side of face Eye/brain 2.E-04 Eye/brain 1.E-05

Left side of neck Eye/brain 2.E-04 Eye/brain 1.E-05

Right side of neck Eye/brain 2.E-04 Eye/brain 1.E-05

Back of head Eye/brain 2.E-04 Eye/brain 1.E-05

Front of neck Eye/brain 2.E-04 Eye/brain 1.E-05

Back of neck Eye/brain 2.E-04 Eye/brain 1.E-05

Front torso: base of neck to end of sternum 10% ENSD 8.61E-02 Lung 1.98E-02

Front torso: end of sternum to lowest rib ENSD 8.61E-01 Lung 1.98E-02

Front torso: lowest rib to iliac crest ENSD 8.61E-01 Lung 1.98E-02
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 37 of 120

Area of skin

AP and AP spot lumbar 

spinea, 1978–1985

LAT and LAT spot lumbar 

spinea, 1978–1985

Guidance Dose Guidance Dose

Front torso: iliac crest to pubis ENSD 8.61E-01 Lung 1.98E-02

Back torso: base of neck to mid-back 10% EXSD 1.9E-03 Lung 1.98E-02

Back torso: mid-back to lowest rib EXSD 1.88E-02 Lung 1.98E-02

Back torso: lowest rib to iliac crest EXSD 1.88E-02 Lung 1.98E-02

Back torso: buttocks (Iliac crest and below) EXSD 1.88E-02 Lung 1.98E-02

Right torso: base of neck to end of sternum 10% ENSD 8.61E-02 10% ENSD 1.71E-01

Right torso: end of sternum to lowest rib ENSD 8.61E-01 ENSD 1.71E+00

Right torso: lowest rib to iliac crest ENSD 8.61E-01 ENSD 1.71E+00

Right torso: iliac crest to pubis (right hip) ENSD 8.61E-01 ENSD 1.71E+00

Left torso: base of neck to end of sternum 10% ENSD 8.61E-02 10% EXSD 8.E-04

Left torso: end of sternum to lowest rib ENSD 8.61E-01 EXSD 7.5E-03

Left torso: lowest rib to iliac crest ENSD 8.61E-01 EXSD 7.5E-03

Left torso: iliac crest to pubis (left hip) ENSD 8.61E-01 EXSD 7.5E-03

a. The organ doses listed are for two projections summed together, the AP and AP spot, and the LAT and LAT spot.

4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

NIOSH has determined that it is not feasible to reconstruct internal dose from January 1, 1949 

through December 31, 1994 for the SNL-NM facility based on insufficient monitoring data, process 

information, and monitoring program information (NIOSH 2011, 2012). 

Dose reconstruction guidance in this site profile for periods included in the SEC is presented to 

provide a technical basis for partial dose reconstructions for nonpresumptive cancers not covered in 

the Special Exposure Cohort (SEC) class (January 1, 1949 through December 31, 1994). Although 

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

class, it intends to use internal and external monitoring data that may 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 but who do not qualify 

for inclusion in the SEC, can be performed using these data as appropriate.

Environmental intake information is determined to be insufficient for dose reconstruction of 

environmental internal doses prior to 1973. Therefore, per the above, such dose reconstruction may 

not be performed unless information in addition to that available in this document becomes available. 

Environmental intake information for employment during 1973 and later is sufficient for dose 

reconstruction of environmental internal doses.

4.1 INTRODUCTION

4.1.1 Purpose

This section provides a technical basis for evaluating the occupational environmental dose for 

EEOICPA claimants who were employed at SNL-NM and the TTR in Nevada. Occupational 

environmental dose refers to radiation exposures received by workers while on the site but outside the

SNL radiological facilities from facility discharges to the atmosphere, from ambient external radiation 

originating in the facilities, and from inadvertent inhalation of site-generated radionuclides. The 

receptors of concern are SNL employees who did not wear external dosimetry or who were not 

monitored for internal exposures. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 38 of 120

4.1.2 Scope

This section describes environmental occupational exposures at SNL facilities but outside the 

occupational workplace. It includes estimated annual intakes of radionuclides from inhalation for the 

period of 1973 to 2010 and the estimated radiation dose from ambient external exposures for the 

period of 1948 to 2010. Results for 2011 and later will be incorporated if necessary at a later date. 

Estimates for different facilities vary by year and by type of exposure because sources of 

radionuclides and radiation did not exist at all facilities for all years during this period.

4.2 OPERATIONS OVERVIEW 

4.2.1 Sandia National Laboratories – New Mexico

SNL-NM began as the Albuquerque branch of LASL and operated from 1945 to 1949. Sandia 

Corporation was formed in 1949; the name changed to Sandia Laboratories in 1971 and to Sandia 

National Laboratories in 1979. Although a single comprehensive history of Sandia facility radiological 

operations has not been identified, information on construction and operation of new facilities can be 

gathered from annual environmental monitoring reports and a few historical summaries (Ulrich 1998; 

SNL 2006). The layout of the SNL-NM site as of 2005 is shown in Figure 4-1.

Figure 4-1. SNL-NM site south of Albuquerque, New Mexico, in 2005.

TA-I began operations in 1946 with one primary building and several temporary buildings until 

construction of more permanent buildings for an expanded facility began in 1948. The Albuquerque 

Laboratory site concentrated primarily on development and testing of nonnuclear components. 

Activities included the operation of a Cockroft-Walton accelerator and a Van de Graaff generator, as 

well as chemistry laboratories that used radiochemicals to support research needs. Environmental 

monitoring reports have noted the potential for release of tritium from TA-I (Millard, Gray, and O’Neal 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 39 of 120

1984), but often note that no releases occurred. A Neutron Generator Facility was completed in 1996 

and has been the source of larger releases of tritium.

Early manufacturing activities at Sandia Laboratories included the assembly of weapons. This 

resulted in the development of TA-II about a half mile south of TA-I for the handling and incorporation

of explosives into the weapons. Construction of TA-II, which paralleled the development of TA-I, was 

initiated in 1948. The resultant construction included two identical assembly buildings and a control 

building completed in 1949. Assembly activities continued at TA-II until 1957. The records note the 

potential for the release of small amounts of 85Kr vented from the earth by explosive testing during 

fracture permeation tests (Millard, Gray, and O’Neal 1984). 

The increasing use of missiles as delivery vehicles led to the full-scale environmental testing of 

weapons with and without explosives. As a result, there was a need for complex equipment and 

specialized engineers to analyze the test results. A decision was made to centralize a group of test 

devices in TA-III, approximately 7 miles south of TA-I. Planning for this area began in 1952; the first 

group of facilities, which consisted of a centrifuge, a rocket sled, a vibration testing facility, and an 

instrument control center, was completed in 1953. The MWL operated at TA-III from 1959 to 1988 as 

a disposal site for low-level radioactive and mixed waste. The Radioactive and Mixed Waste 

Management Facility was completed in 1995 for repackaging low-level radioactive and transuranic 

wastes from current and legacy activities at SNL and its predecessors. 

In 1957, a nuclear reactor facility was proposed for what was to become TA-V. TA-V is in the extreme 

northeast corner of TA-III and is remote from TAs-I, -II, and -IV as well as from most of the facilities in 

TA-III. The SPR facility was started in May 1961. The SPR is an unreflected, cylindrical, enricheduranium assembly. Small amounts of fission product gases and air activation products (primarily 41Ar) 

are produced and released during operations. The SER was a 5-MW heterogeneous reactor fueled 

with aluminum-clad 235U-enriched U-Al alloy elements. The reactor was cooled and moderated by 

light water and was operational from October 1962 to June 1969. Air activation products (primarily 

41Ar) were released through the stack. The ACPR (now the ACRR) is a modified Training, Research, 

Isotopes General Atomics (TRIGA) reactor that became operational in June 1967. Air activation 

products (primarily 41Ar) were released through the stack (Brewer 1973). The HCF began operation in 

1979 (Simmons 1980). Other facilities in TA-V have included electron beam accelerators, a GIF, and 

a neutron irradiation facility. Tritium could have been released from the accelerators during 

operations and other radionuclides could have been released from HCF operations.

Beginning in 1979, TA-IV was developed as a location for inertial confinement fusion research and 

pulsed power research. The first facility was the Particle Beam Fusion Accelerator (PBFA)-I 

(originally to be called the Electron Beam Fusion Facility), which was completed in 1980 with an 

upgrade to PBFA-II completed in 1985. After additional upgrades, this accelerator is now called the Z 

Pinch Machine. Gaseous tritium (HT) effluents were generated from TA-IV activities during some 

years (Millard, Gray, and O’Neal 1984), but not during others. More accelerator facilities were 

developed in TA-IV, including Hermes and others. These accelerators have or had the potential to 

generate short-lived activation product radionuclides as well as external radiation. 

4.2.2 Sandia National Laboratories – Nevada

The TTR is about 140 miles northwest of Las Vegas, Nevada, and 32 miles southeast of Tonopah, 

Nevada. It is on approximately 526 square miles at the northern boundary of the Nevada Test and 

Training Range (NTTR; formerly Nellis Air Force Range), which was established in 1940 by President 

Roosevelt as the Las Vegas Bombing and Gunnery Range. The topography at TTR is characterized 

by a broad flat valley bordered by two north- and south-trending mountain ranges: the Cactus Range 

to the west (occurring mostly within the boundaries of TTR) and the Kawich Range to the east. 

Cactus Flat is the valley floor where the main operational area of TTR is located. An area of low hills 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 40 of 120

outcrops in the south. Elevations within TTR range from 1,630 m (5,347 ft) at the valley floor to 

2,279 m (7,482 ft) at Cactus Peak. The elevation in the town of Tonopah is at 1,837 m (6,030 ft) 

(Sanchez, Hamilton, and Mayeux 2001a). Figure 4-2 shows the layout of SNL-NV at TTR in 2004.

The TTR was eventually selected as a bombing range after similar facilities at the Salton Sea Test 

Base in California, as well as Yucca Flat on the NTS, became inadequate. The TTR site was well 

suited because it had immense areas of flat terrain needed for the increasing use of rockets and lowaltitude, high-speed aircraft operations (Sanchez, Hamilton, and Mayeux 2001a). 

The TTR area was withdrawn in 1956 and Sandia began activities in 1957 to operate and test new 

weapon systems. In the years following World War II, facilities that were built at TTR were originally 

designed and equipped to gather data on aircraft-delivered inert test vehicles under AEC (now DOE) 

cognizance. Over the years, the facilities and capabilities at TTR were expanded to accommodate 

tests related to the DOE Weapons Ordnance Program (Sanchez, Hamilton, and Mayeux 2001a).

The Main Compound in Area 3 is the heart of the test range activities. The Operations Control Center 

controls and coordinates all test functions and affords a 360° view of the site. During test operations, 

the test director, range safety officer, test project engineer, camera controller, and range 

communicator operate the consoles in the Operations Control Center to control and coordinate all test 

functions. Another important location at the range is Area 9, which has weapons storage facilities and 

is used to conduct ground-to-air rocket launching tests (Sanchez, Hamilton, and Mayeux 2001a).

Principal DOE activities at TTR include stockpile reliability testing; research and development testing 

support of structural development; arming, fusing, and firing systems testing; and testing nuclear 

weapon delivery systems. However, no nuclear devices are tested at TTR. TTR has a wide array of 

signal-tracking equipment including video; high-speed cameras; radar tracking devices used to 

characterize ballistics, aerodynamics, and parachute performance on artillery shells; bomb drops; 

missiles; and rockets (Sanchez, Hamilton, and Mayeux 2001a).

Figure 4-2. Layout and location of facilities operated by SNL-NV at TTR in 2004.

In recent years, specific test activities at TTR have consisted of the following:
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• Air drops (trajectory studies of simulated weapons)

• Gun firings

• Ground-launched rockets (study of aeroballistics and material properties)

• Air-launched rockets (deployed from aircraft)

• Explosive testing (e.g., shipping and storage containers)

• Static rocket tests (related to the Trident Submarine Program)

• Ground penetrator tests

These activities require a remote range both for public safety and to maintain national security. The 

majority of test activities at TTR occur in Cactus Flat, a valley with almost no topographical relief 

flanked by mountains and hills (Sanchez, Hamilton, and Mayeux 2001a).

Clean Slate and Double Tracks Sites

In May and June 1963, Project Roller Coaster included a series of four nuclear weapons destruction 

tests that resulted in plutonium dispersal in the surrounding soils. These tests were conducted to 

study plutonium dispersal from accidental nonnuclear explosions of plutonium-bearing weapons 

(Millard and Lathrop 1982). Three of these tests were conducted within the boundaries of TTR; the 

fourth was conducted on the NTTR just west of TTR. The three Project Roller Coaster test sites at 

TTR are referred to as Clean Slates 1, 2, and 3 (shown in Figure 4-2). The fourth test site at NTTR is 

referred to as Double Tracks. In 1996, the Double Tracks was closed after remediation of soil 

contamination to a level of no more than 200 pCi/g of transuranic radionuclides (Sanchez, Hamilton, 

and Mayeux 2001a). 

The initial cleanup of each Clean Slate site was conducted shortly after each test. Test-related debris 

was bladed into a hole at ground zero and backfilled. An initial fence was built around each test area 

where the soil contamination was set at approximately 1,000 μg/m2 of plutonium. The soil survey was 

conducted on 61-m grids with a hand-held survey meter or field instrument for the detection of lowenergy radiation. In 1973, additional outer fences were set at 40 pCi/g of plutonium in soil, also using 

the hand-held meter method. Soil sampling is conducted periodically at these sites and the areas are 

visually inspected twice a year to determine if fence repairs are required. Horses that wander inside 

the fenced areas are promptly removed (Sanchez, Hamilton, and Mayeux 2001a). 

The Clean Slate sites are the only known sources of potential occupational environmental 

radionuclide exposure at TTR. 

4.3 RADIONUCLIDE SCREENING, SNL-NEW MEXICO 

SNL-NM’s principal research activities have mainly been nonradiological; these activities have never 

been large producers of radionuclides. Radiation and radionuclides have been used to support 

research activities; their use has generally increased since the early days. Many of the research 

activities have been episodic, with durations of several years before being suspended or changed.

Radioactive effluent monitoring information is available for 1973 to 2010 and includes listed airborne 

emissions of 69 radionuclides. Many of these radionuclides were released in very small quantities 

that were negligible contributors to worker dose. Therefore, radionuclide screening was conducted to 

identify those radionuclides that were important contributors to dose. The parameters for the 

screening were as follows:

• The maximum yearly release for each radionuclide was evaluated.

• The screening threshold was set at 0.01 mrem/yr for the maximum release.
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• The exposure pathways of inhalation and external dose from plume passage were considered, 

using the air dose screening factors from NCRP Report 123 (NCRP 1996).

• The radionuclides were assumed to be released only during the normal working day (over a 

period of 2,000 hours rather than continuously over 8,760 hours); workers were assumed to be 

exposed for 2,000 hr/yr.

• Tritium intakes were applied at 150 percent of the inhalation rate to account for additional dose 

due to absorption of tritium.

• The atmospheric dispersion factor was assumed to be 0.001 s/m3

, a value approximately 

3 times higher than the highest atmospheric dispersion factor for a ground-level release at 

100 m calculated using Albuquerque-specific meteorology from the CAP88PC Version 3.0 

computer program (TEA 2007). 

The screening indicated 22 radionuclides with the potential to exceed the 0.01-mrem/yr threshold for 

the maximum yearly release. These radionuclides and their exposure pathways are listed in 

Table 4-1.

Table 4-1. Radionuclides meeting screening criteria and the principal worker exposure pathway.

Inhalation External plume Both pathways

Am-241, H-3, I-129, I-131, I-133, Pu-239, 

U-238, Th-232, Co-60, Sr-90

Ar-41, Kr-85m, Kr-87, N-13, O-15, 

Xe-133, Xe-133m, Xe-135, Xe-135m

I-135, Rb-88, Kr-88

4.4 AMBIENT EXTERNAL RADIATION

In applying the ambient doses described for SNL-NM and the TTR, a maximizing approach can be 

used when appropriate, assuming the employee worked continuously in the area with the highest 

external onsite ambient dose rate for 50 hours per week for 52 weeks per year, or 2,600 hours per 

year (ORAUT 2006a). Thus, the ambient doses in Tables 4-2, 4-3, and 4-4 will be adjusted for 

2,600 hours of exposure for the maximizing approach. 

If a best estimate of onsite ambient dose is performed, it should be assumed that the energy 

employee worked 50 hours per week for 50 weeks per year, for a total of 2,500 work-hours per year. 

The onsite ambient dose for actual work areas, if known, shall be applied for best estimates. If a 

particular work area with recorded onsite ambient doses cannot be identified, the site maximum value 

must be applied for a best estimate. 

4.4.1 Sandia National Laboratories – New Mexico

Ambient external radiation dose information for SNL-NM is available for 1980 to 2010 (Simmons 

1980; Millard et al. 1981, 1982, 1983, 1987, 1988, 1989a; Millard, Gray, and O’Neal 1984; Millard, 

Gray, and Thompson 1985, 1986; Hwang et al. 1990a, 1991a; Culp et al. 1992, 1993, 1994; Shyr et 

al. 1995; Shyr, Duncan, and Sanchez 1996; Fink and Duncan 1997; Fink, Duncan, and Sanchez 

1998; Fink 1999; Duncan and Sanchez 2001; Sanchez, Hamilton, and Mayeux 2001b; Sanchez et al. 

2002a, 2008a, 2009a, 2010a, 2011a; Sneddon et al. 2007a; Wagner et al. 2003a, 2004a, 2005a, 

2006a). Sources of ambient external radiation could have been direct external radiation produced by 

reactors in TA-V and accelerators (principally those in TA-IV) and exposure to external radiation from 

plumes of short-lived activation products released from accelerator operations. Evaluation of the 

potential external doses from these plumes showed potential doses to be very low in most cases, on 

the order of a few microrem per year using the screening criteria in Section 4.3.
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Starting in 1980, TLDs were used to monitor the environmental ionizing radiation background levels in 

areas surrounding the research facilities at SNL, in areas along the perimeter of Kirtland AFB and 

within and surrounding the City of Albuquerque. Field TLDs were replaced and measured quarterly. 

TLDs were prepared and evaluated by the SNL staff. The results of the TLD measurements were 

reported in annual environmental monitoring reports to the DOE. 

The locations of the environmental TLD stations were specified by SNL technical staff and are 

described in the annual environmental monitoring reports. From 1980 to 1985 and from 1993 to 1999, 

only the arithmetic mean plus one standard deviation was reported. The TLD data for each station 

were reported for 1986 to 1992 and for 2000 to 2010. 

Examination of TLD results shows little difference between onsite and offsite locations except that, for 

measurement data before 2005, there are infrequent instances of unusually high quarterly results for 

TA-V. During these periods, activities in TA-V were likely to have resulted in higher external doses to 

area workers. An assumption favorable to claimants was made that workers in this area could have 

received external ambient doses of 10 mrem/yr during the years the area was active. It was assumed 

that workers in TA-III adjacent to TA-V could have received some lower exposure from TA-V activities. 

Accelerator activities in TA-IV were assumed to have the potential to result in higher external doses. 

It was also assumed that workers in TAs-I and -II could have received some lower exposure from TAIV activities. The estimates of external onsite ambient dose to unmonitored workers in the different 

TAs were based on a working year of 2,000 hours and are listed in Table 4-2 for 1948 through 2004. 

For measurement data of 2005 and later, the actual TA results, including any general site 

measurements in immediate proximity to a TA, are provided in Table 4-3 as ambient doses with 

correction for mean offsite background measurements. The maximum onsite ambient dose, which 

may have been recorded outside of a TA, corrected for mean offsite measurements, is also provided. 

Table 4-2. External radiation dose for workers in SNL-NM TAs (mrem/yr).a

Year TA-I TA-II TA-III TA-IV TA-V Year TA-I TA-II TA-III TA-IV TA-V

1948–1952 < 1 < 1 N/A N/A N/A 1979 < 1 < 1 5 N/A 10

1953 < 1 < 1 < 1 N/A N/A 1980 5 5 5 10 10

1954 < 1 < 1 < 1 N/A N/A 1981 5 5 5 10 10

1955 < 1 < 1 < 1 N/A N/A 1982 5 5 5 10 10

1956 < 1 < 1 < 1 N/A N/A 1983 5 5 5 10 10

1957 < 1 < 1 < 1 N/A N/A 1984 5 5 5 10 10

1958 < 1 < 1 < 1 N/A N/A 1985 5 5 5 10 10

1959 < 1 < 1 < 1 N/A N/A 1986 5 5 5 10 10

1960 < 1 < 1 < 1 N/A N/A 1987 5 5 5 10 10

1961 < 1 < 1 5 N/A 10 1988 5 5 5 10 10

1962 < 1 < 1 5 N/A 10 1989 5 5 5 10 10

1963 < 1 < 1 5 N/A 10 1990 5 5 5 10 10

1964 < 1 < 1 5 N/A 10 1991 5 5 5 10 10

1965 < 1 < 1 5 N/A 10 1992 5 5 5 10 10

1966 < 1 < 1 5 N/A 10 1993 5 5 5 10 10

1967 < 1 < 1 5 N/A 10 1994 5 5 5 10 10

1968 < 1 < 1 5 N/A 10 1995 5 5 5 10 10

1969 < 1 < 1 5 N/A 10 1996 5 5 5 10 10

1970 < 1 < 1 5 N/A 10 1997 5 5 5 10 10

1971 < 1 < 1 5 N/A 10 1998 5 5 5 10 10

1972 < 1 < 1 5 N/A 10 1999 5 5 5 10 10

1973 < 1 < 1 5 N/A 10 2000 5 5 5 10 10

1974 < 1 < 1 5 N/A 10 2001 5 5 5 10 10

1975 < 1 < 1 5 N/A 10 2002 5 5 5 10 10
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Year TA-I TA-II TA-III TA-IV TA-V Year TA-I TA-II TA-III TA-IV TA-V

1976 < 1 < 1 5 N/A 10 2003 5 5 5 10 10

1977 < 1 < 1 5 N/A 10 2004 5 5 5 10 10

1978 < 1 < 1 5 N/A 10

a. Based on a working year of 2,000 hours.

Table 4-3. External radiation dose for workers in SNL-NM TAs 

(mrem/yr), 2005 through 2010.a

Year TA-I TA-II TA-III TA-IV TA-V Maximum site

2005 < 1 1 4 2 1 6

2006 < 1 < 1 < 1 < 1 < 1 3

2007 < 1 < 1 < 1 < 1 < 1 3

2008 < 1 1 1 < 1 < 1 4

2009 < 1 1 < 1 1 1 5

2010 < 1 1 1 < 1 < 1 4

a. Based on a working year of 2,000 hours.

As demonstrated by Tables 4-2 and 4-3, if onsite ambient dose values are required for dose

reconstruction of employment beyond 2010, respective TA doses and maximum doses will be 

moderately overestimated in recent years by applying 2004 data. Therefore, it is favorable to 

claimants and a reasonable practice to apply 2004 onsite ambient doses to any evaluated 

employment during 2011 and later if the employee was not monitored by external dosimetry.

4.4.2 Sandia National Laboratories – Nevada

The U.S. Environmental Protection Agency (EPA) has monitored external radiation at TTR and at 

Goldfield and Tonopah, Nevada, since at least 1971 (EPA 1972) as part of overall NTS offsite 

environmental monitoring. The early results are typically a single dosimeter reading for the three 

locations. EPA results are reported in annual environmental monitoring reports for TTR (Millard and 

Lathrop 1982, 1984, 1985; Millard 1986; Millard and West 1987, 1988; Millard et al. 1989b; Hwang et 

al. 1990b, 1991a; Howard and Culp 1992; Culp and Howard 1993; Culp, Howard and McClellan 1994; 

Culp and Forston 1995, 1996, 1997, 1998; Duncan and Sanchez 1999; Forston 2000; Sanchez, 

Hamilton, and Mayeux 2001b; Sanchez et al. 2002b, 2008b, 2009b, 2010b, 2011b; Sneddon et al. 

2007b; Wagner et al. 2003b; 2004b, 2005b, 2006b) as well as in the EPA offsite environmental 

monitoring reports. These results generally do not show significant differences in external radiation 

dose between the NTS offsite locations and TTR.

A separate long-term environmental external radiation monitoring network using TLDs at TTR was 

established by SNL-NV in January 1994. Environmental TLDs were placed at locations off the site, at 

the site perimeter, and on the site to measure gamma radiation (Culp and Forston 1995). This 

program significantly expanded the level of external radiation monitoring at TTR. One dosimeter 

location (T-13) at the northeast corner of the Operations Center perimeter fence had elevated results 

for 1995, 1996, and 1997. During these years, the average annual dose at T-13 was 210 ±26, 

260 ±42, and 218 ±52 mrem, respectively, while the site annual averages including these results were 

144 ±22, 159 ±17, and 150 ±21 mrem, respectively. Results were reported for only two of three 

periods in 1994, but the third period record was 108.1 ±9.9 mrem, which was higher than would be 

expected and consistent with higher results of the following years. Soil sampling (OC-1) was 

conducted near T-13 and the results did not indicate radionuclide contamination. Gamma surveys 

were performed using hand-held radiation survey equipment; the surveys found no unusual results. 

Reviews of the historic TLD data for T-13 showed the elevated results did not occur every monitoring 

period but were more cyclic or intermittent in nature (Culp and Forston 1998). 

The possible source of these higher T-13 TLD results was not identified in the annual monitoring 

reports. However, a teleconference with an author of the reports indicated he later gained information 
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that these higher results could have related to radiography of test assemblies in the Operations area 

(Ikenberry 2006a). A teleconference interview with an SNL Radiation Protection Dosimetry employee

indicated that all workers had personal TLDs and external dosimetry records were available for all 

workers since 1990 (Ikenberry 2006b). Therefore, the potential for radiation exposure from 

radiography operations would be included in the workers’ personal external dosimetry records. 

Further, the employee interviewed was of the opinion that this trend was established well before 1990, 

but that only the records for 1990 and later were readily and electronically available from the 

dosimetry database. Previous radiation doses for each individual were compiled in the database as a 

total external dose record for each worker, and pre-1990 yearly individual hard-copy records should 

be available. For measurement data of 2005 and later, the actual environmental monitoring results 

were determined with correction for mean offsite background measurements. 

As noted above, it is unlikely that onsite ambient dose reconstruction will be necessary since external 

doses would have been monitored. However, the data are provided in the event that reconstruction of 

such dose is warranted. Table 4-4 provides the estimates of external ambient dose from TTR 

operations to unmonitored workers.

Table 4-4. External radiation dose for workers 

at TTR (mrem/yr).

Years Annual external radiation dose

1957–2004 0

2005 20

2006 17

2007 7

2008 12

2009 7

2010 8

Based on contemporary TTR site radiological characteristics, routine site activities (Sanchez et al. 

2011b), and data listed in Table 4-4, if onsite ambient dose values are required for dose 

reconstruction of employment beyond 2010, TTR onsite ambient dose can be estimated as 0.020 rem 

per year for 2011 and later.

4.5 INHALATION OF ONSITE AIRBORNE RADIONUCLIDES 

4.5.1 Sandia National Laboratories – New Mexico

The environmental monitoring program at SNL-NM was initiated in December 1958 to establish 

background radiation levels before the startup of the reactors in TA-V (Burnett et al. 1961). There is 

no indication of radionuclide release before TA-V reactor activities; however, assumptions favorable to 

claimants about earlier releases can be made for the TAs. 

Radionuclide air emission data are available in the annual environmental monitoring reports for SNLNM from 1973 to 2010 (Brewer 1974; Holley 1975; Holley and Simmons 1976; Simmons 1977, 1978, 

1980; Millard et al. 1981, 1982, 1983; Millard, Gray, and O’Neal 1984; Millard, Gray, and Thompson 

1985, 1986; Millard et al. 1987, 1988, 1989a; Hwang et al. 1990a, 1991b; Culp et al. 1992, 1993, 

1994; Shyr et al. 1995; Shyr, Duncan, and Sanchez 1996; Fink and Duncan 1997; Fink, Duncan, and 

Sanchez 1998; Fink 1999; Duncan and Sanchez 2001; Sanchez, Hamilton, and Mayeux 2001b; 

Sanchez et al. 2002a, 2008a, 2009a, 2010a, 2011a; Sneddon et al. 2007a; Wagner et al. 2003a, 

2004a, 2005a, 2006a). The 1973 report stated that because SNL released only small amounts of 41Ar 

from TA-V and tritium from TA-I as a result of operations, dose estimates were based on effluent 

release data rather than environmental monitoring data. X/Q values were calculated for the various 

distances from the release points to the site boundaries. In 1979, 85Kr was included as an effluent 
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from TA-II. In 1989, 133Xe, 13N, 15O, and 238U were added to the list of radioactive air emissions. In 

1990, 129I was reported as a release from the HCF at TA-V.

Three findings from the 1991 Tiger Team Assessment of SNL-NM questioned the adequacy of 

airborne radionuclide emission monitoring. SNL responded (SNL 1995a) that although it did not have 

any point source with the potential for doses to the public greater than 0.1 mrem/yr that would require 

continuous monitoring, it had installed stack monitors capable of continuous monitoring at two 

facilities in TA-V: the ACRR and the HCF. As a result of this assessment, the number of reported 

effluents from SNL increased in 1991 to 32 radionuclides released from 13 facilities. The practice of 

reporting radionuclides from more facilities has continued.

Reported facility radionuclide emissions were the basis for estimating potential worker environmental 

inhalation intakes. Years with data vary by facility, but the most complete information is available for 

1991 and later. For years and areas in which there are no emission data, environmental inhalation 

intakes will not be calculated. Identification of specific radionuclides released from facilities in stack 

emissions, which are available in data from 1973 to 2010, were used to characterize radionuclide 

emissions for all years. There are no radionuclide-specific data from earlier years. 

Estimates of inhalation intakes were made using the exposure criteria that were included in the initial 

radionuclide screening. These criteria are: 

• The radionuclides were assumed to be released only during the normal working day (over a 

period of 2,000 hours rather than continuously over 8,760 hours), which maximized potential 

worker exposure; workers were assumed to be exposed for 2,000 hr/yr.

• Any uranium and thorium detected is assumed to originate from facility activities, even though 

it could be naturally occurring.

• Tritium intakes were applied at 150 percent of the inhalation rate to account for additional dose 

due to absorption of tritium.

• The atmospheric dispersion factor was assumed to be 0.001 s/m3

, a value approximately 

3 times higher than the highest atmospheric dispersion factor for a ground-level release at 

100 m calculated using Albuquerque-specific meteorology from the CAP88PC Version 3.0 

computer program (TEA 2007). The assumed worker breathing rate for light activity was 1.2

m3

/hr (3.3 × 10-4 m3/s) (ICRP 1994a).

Use of these parameters results in estimated radionuclide inhalation intakes that are favorable to 

claimants. 

4.5.1.1 Technical Areas I, II, and IV

No emission monitoring data were available from before 1973 for TAs-I and -II. TA-IV was not created 

until 1979. Environmental monitoring data indicate that tritium was the principal radionuclide 

released; most available data was from 1991 and later. A TA-I radiation laboratory was the source of 

small releases of 241Am and 238U with data available for a limited number of years.

The only radionuclide released in TA-IV that could contribute to significant inhalation dose was tritium. 

Inhalation of tritium was assumed to be the same as inhalation in TAs-I and -II from 1979 and later. 

This is also an assumption favorable to claimants that overestimates inhalation of tritium in TA-IV

based on the information contained in the environmental monitoring reports (Sanchez 2001a, 2010a, 

2009a, 2008a, Sneddon 2007a and Wagner 2006a). In 2009, it was reported that the annual release 

from the TA-I Sandia Tomography and Radionuclide Transport (START) Laboratory included 
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sufficient 239Pu to warrant dose reconstruction for resulting environmental intakes (Sanchez et al. 

2010a). Thus, 239Pu annual intake rates are included for TA-1 for 2009, but are not significant for 

other TAs or other years. The START Laboratory also contributed to a significantly increased 241Am 

TA-I environmental intake in 2010 (Sanchez et al. 2011a), as listed in Table 4-5. 

The estimated inhalation intakes of radionuclides for TAs-I, -II, and -IV are listed in Table 4-5. The 

material form and solubility selected for each radionuclide should be the values most favorable to the 

claimant.

Table 4-5. TAs-I, -II, and -IV maximum annual intakes (Bq/yr) via inhalation.

a

Year

Radionuclide

Year

Radionuclide

TA-I and TA-II TA-IV TA-I and TA-II TA-IV

H-3 Am-241 U-238 H-3 H-3 Am-241 U-238 Pu-239 H-3

1973 2E+03 2E-07 5E-08 N/A 1992 2E+00 2E-07 5E-08 N/A 2E+00

1974 6E+02 2E-07 5E-08 N/A 1993 2E+00 2E-07 5E-08 N/A 2E+00

1975 6E+02 2E-07 5E-08 N/A 1994 2E+00 2E-07 5E-08 N/A 2E+00

1976 6E+02 2E-07 5E-08 N/A 1995 6E+00 2E-07 5E-08 N/A 6E+00

1977 2E+02 2E-07 5E-08 N/A 1996 2E+03 6E-09 5E-08 N/A 2E+03

1978 5E+04 2E-07 5E-08 N/A 1997 4E+04 6E-09 5E-08 N/A 4E+04

1979 6E+04 2E-07 5E-08 N/A 1998 2E+05 6E-09 5E-08 N/A 2E+05

1980 8E-01 2E-07 5E-08 1E+00 1999 5E+04 6E-09 5E-08 N/A 5E+04

1981 8E-01 2E-07 5E-08 1E+00 2000 6E+04 6E-09 5E-08 N/A 6E+04

1982 8E-01 2E-07 5E-08 8E-01 2001 8E+04 6E-09 5E-08 N/A 8E+04

1983 8E-01 2E-07 5E-08 8E-01 2002 6E+03 6E-09 5E-08 N/A 6E+03

1984 8E-01 2E-07 5E-08 8E-01 2003 8E+03 6E-09 5E-08 N/A 8E+03

1985 8E-01 2E-07 5E-08 8E-01 2004 2E+03 6E-09 5E-08 N/A 2E+03

1986 8E-01 2E-07 5E-08 8E-01 2005 1E+04 6E-09 3E-06 N/A 1E+04

1987 8E-01 2E-07 5E-08 8E-01 2006 6E+03 6E-09 3E-06 N/A 6E+03

1988 8E-01 2E-07 5E-08 8E-01 2007 2E+05 3E-04 2E-08 N/A 2E+05

1989 8E-01 2E-07 5E-08 8E-01 2008 7E+05 4E-04 2E-08 N/A 7E+05

1990 3E+00 2E-07 5E-08 3E+00 2009 1E+05 4E-04 2E-04 8E-03 1E+05

1991 8E-01 2E-07 5E-08 8E-01 2010 1E+05 6E-01 2E-04 N/A 1E+05

a. Radionuclide solubility class should be selected to be favorable to the claimant.

4.5.1.2 Technical Area III

TA-III activities began in 1953; it is unclear if radionuclides were involved during the earliest years. 

Environmental monitoring data indicate that tritium was the principal radionuclide released, mainly 

from the diffuse source MWL with data available for 1993 and later. The area has included a landfill 

that has been a diffuse source of small quantities of 60Co, 232Th, and 238U; a facility that was a small 

source of 241Am in the mid-1990s; and a mixed waste facility that was a comparatively large source of 

241Am and 90Sr from 2002 to 2004. Release and inhalation of 241Am and 90Sr from TA-III was 

assumed to occur only during the years in which these radionuclides were identified (2002 to 2010). 

The estimated inhalation intakes of radionuclides for TA-III are listed in Table 4-6. The material form 

and solubility selected for each radionuclide should be the values most favorable to the claimant.

Table 4-6. TA-III maximum annual intakes (Bq/yr) via inhalation.

a

Year

Radionuclide

H-3 Am-241 U-238 Th-232 Co-60 Sr-90

1993 3E+04 N/A 6E-02 2E-04 3E+00 N/A

1994 6E+03 N/A 6E-02 2E-04 3E+00 N/A

1995 6E+03 N/A 6E-02 2E-04 3E+00 N/A

1996 8E+04 N/A 6E-02 2E-04 3E+00 N/A

1997 5E+04 N/A 6E-02 2E-04 3E+00 N/A

1998 2E+04 N/A 2E-03 2E-04 3E+00 N/A
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Year

Radionuclide

H-3 Am-241 U-238 Th-232 Co-60 Sr-90

1999 2E+04 N/A 6E-02 2E-04 3E+00 N/A

2000 2E+04 N/A 6E-02 2E-04 8E+00 N/A

2001 6E+03 N/A 6E-02 2E-04 2E-04 N/A

2002 8E+03 3E-03 6E-02 2E-04 2E-01 5E-03

2003 3E+05 2E-01 6E-02 2E-04 3E+00 2E-01

2004 2E+04 2E-01 6E-02 2E-04 3E+00 5E-03

2005 2E+05 2E-01 6E-02 2E-04 3E+00 4E-03

2006 2E+05 2E-01 6E-02 2E-04 3E+00 4E-03

2007 2E+05 2E-01 6E-02 2E-04 3E+00 4E-03

2008 1E+06 2E-01 6E-02 2E-04 3E+00 6E-03

2009 5E+04 2E-01 6E-02 2E-04 3E+00 5E-03

2010 2E+05 5E-03 6E-02 2E-04 3E+00 2E-03

a. Radionuclide solubility class should be selected to be favorable to the claimant.

4.5.1.3 Technical Area V

TA-V activities began in 1961 with operation of the SPR. Early environmental monitoring reports 

indicate that most TA-V releases were noble gases. There is some indication of tritium releases from 

TA-V in the late 1970s and for small releases of radioiodines and other radionuclides from the ACRR 

and HCF in the mid-1990s. 

The estimated inhalation intakes of radionuclides for TA-V are listed in Table 4-7. The material form 

and solubility selected for each radionuclide should be the values most favorable to the claimant. The 

intake data include no results after 2004 because monitoring data from 2005 and later show no TA-V 

releases that would result in significant internal doses.

Table 4-7. TA-V maximum annual intakes (Bq/yr) via inhalation.

a

Year

Radionuclide

H-3 I-129 I-131 I-133 I-135 Kr-88 Rb-88

1973 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1974 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1975 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1976 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1977 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1978 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1979 6E+04 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1980 2E+05 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1981 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1982 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1983 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1984 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1985 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1986 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1987 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1988 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1989 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1990 4E+02 2E-01 2E+01 2E+02 8E+02 2E+03 2E+03

1991 4E+02 2E+00 2E+01 2E+02 8E+02 2E+01 3E+01

1992 4E+02 3E+00 2E+00 3E+02 2E+03 2E+03 3E+03

1993 4E+02 2E+00 2E+01 2E+02 8E+02 5E+03 5E+03

1994 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+02

1995 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 5E+00

1996 4E+02 2E+00 2E+01 8E+01 2E+01 6E+03 2E+03
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Year

Radionuclide

H-3 I-129 I-131 I-133 I-135 Kr-88 Rb-88

1997 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1998 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

1999 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

2000 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

2001 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

2002 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

2003 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

2004 4E+02 2E+00 2E+01 2E+02 8E+02 2E+03 2E+03

a. Radionuclide solubility class should be selected to be favorable to the claimant.

4.5.2 Sandia National Laboratories – Nevada

The only identified source of potential radionuclide inhalation at TTR is resuspension of radionuclides 

from the Clean Slate sites (Millard and Lathrop 1982, 1984, 1985; Millard 1986; Millard and West 

1987, 1988; Millard et al. 1989b; Hwang et al. 1990b, 1991a; Howard and Culp 1992; Culp and 

Howard 1993; Culp, Howard and McClellan 1994; Culp and Forston 1995, 1996, 1997, 1998; Duncan 

and Sanchez 1999; Forston 2000; Sanchez, Hamilton, and Mayeux 2001b; Sanchez et al. 2002b, 

2008b, 2009b, 2010b, 2011b; Sneddon et al. 2007b; Wagner et al. 2003b; 2004b, 2005b, 2006b). No 

other areas have been determined to be contaminated with site-originated radionuclides. 

As noted in Section 4.2.2, the highest areas of contamination at the Clean Slate sites were initially 

buried and fenced off after the events in 1963 and a second, larger perimeter fence was set 10 years 

later. Lower levels of transuranic radionuclide contamination have been identified south of the original 

sites and away from the principal occupied areas of the range. These levels are due to the prevailing 

winds at the time the events occurred (Ikenberry 2006a). Soil sampling is conducted each year to 

assess the potential for dispersion of transuranic radionuclide contamination. No contamination has 

been detected except in the areas south of the Clean Slate sites. These sites are remote from the 

Operations Center, and the annual site environmental reports emphasize that test activity is planned 

to avoid disturbing these areas. 

SNL instituted continuous air monitoring at the TTR airport for a 1-year period from February 22, 

1996, to February 25, 1997, to assess potential radiation dose from resuspension of radionuclides 

from the Clean Slate sites as part of an investigation in relation to the National Emission Standards for 

Hazardous Air Pollutants (Duncan and Sanchez 1999). The TTR airport was determined to be the 

location for the highest calculated dose to an onsite maximally exposed individual and so is likely to 

be a location favorable to claimants for estimating exposure of unmonitored SNL-NV workers. The 

average annual air concentrations were measured as follows: 241Am at 4.1 × 10-18 Ci/m3

, 238Pu at 

1.6 × 10-18 Ci/m3

, and 239/240Pu at 9.5 × 10-18 Ci/m3 (Duncan and Sanchez 1999). Based on a 

2,000-hr/yr exposure to these concentrations with a light activity breathing rate of 3.3 × 10-4 m3

/s, the 

activity inhaled annually is 4.6 × 10-4 Bq 241Am, 1.8 × 10-4 Bq 238Pu, and 1.0 × 10-5 Bq 239/240Pu.

Because of the early efforts to bury soil with high contamination levels near ground zero, the fencing 

of the Clean Slate contaminated areas, and the operational avoidance of the Clean Site areas, the air 

concentrations at the TTR airport in 1996 and 1997 are an estimate of radionuclide inhalation for 

unmonitored workers that is favorable to claimants. The estimated inhalation intakes of radionuclides 

for TTR are listed in Table 4-8. The material form and solubility selected for each radionuclide should 

be the values most favorable to the claimant.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 50 of 120

Table 4-8. TTR maximum annual intakes (Bq/yr) via inhalation.

a

Year

Radionuclide

Am-241 Pu-238 Pu-239/240

1957–1962 0 0 0

1963–2010b 4E-04 2E-04 8E-05

a. Radionuclide solubility class should be selected to be favorable to the claimant.

b. Potential inhalation after 1963 is estimated on dispersion of contaminants from 

the Clean Slate sites (see Section 4.2.2).

4.6 UNCERTAINTY

There are no documents available from which quantitative estimates of uncertainty can be made, so 

qualitative estimates are provided. 

External Dose

For maximized onsite ambient dose reconstruction that is favorable to claimants, the dose values 

should be adjusted by a multiplication factor of 1.3 to account for potential under-response by the 

ambient radiation measurement dosimetry. The adjusted doses should be assigned in the Interactive 

RadioEpidemiological Program (IREP) as a constant value. 

For best estimate onsite ambient doses, a qualitative estimate of external dose uncertainty is based 

on a normal distribution with a standard deviation of 30% for all exposure periods and TAs. 

Inhalation Intakes

There are no available documents with estimates of uncertainty for SNL-NM stack releases. A 

qualitative estimate of inhalation intakes assumes lognormally distributed chronic intakes, with 

uncertainty of a geometric standard deviation (GSD) of 2 for all exposure periods. A qualitative 

estimate of inhalation intake uncertainty at SNL-NV is based on a lognormal distribution with a GSD of 

2 for all exposure periods.

The GSDs reported above consider the uncertainty associated with the determination of intake activity 

but do not include the biological variation associated with the conversion to dose. Therefore, a GSD 

of 3 should be applied to internal doses based on environmental intakes to account for biological 

variation and uncertainty in the models (ORAUT 2012). 

5.0 OCCUPATIONAL INTERNAL DOSE

NIOSH has determined that it is not feasible to reconstruct internal dose from January 1, 1949

through December 31, 1994 for the SNL-NM facility based on insufficient monitoring data, process 

information, and monitoring program information (NIOSH 2011, 2012). As such, this time period has 

been included in the EEOICPA Special Exposure Cohort (SEC). The EEOICPA covered period for 

SNL- NM is from January 1st, 1945 to present. Dose reconstruction guidance in this site profile for the 

period between January 1st, 1949 and December 31st, 1994 is presented to provide a technical basis 

for partial dose reconstructions for nonpresumptive cancers not covered in the Special Exposure 

Cohort (SEC) class. 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 may 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 from January 1, 1949, through December 31, 1994, but who do not qualify for inclusion in the 

SEC, can be performed using these data as appropriate.

Occupational internal dose is the dose received by an individual from an intake of radioactive material 

while performing tasks in buildings and structures at SNL or from activities outside the buildings, such 

as burial of waste and monitoring of tests, where occupational intakes of radioactive material could 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 51 of 120

occur. This document contains information for reconstruction of occupational internal doses at SNL 

facilities throughout its history. 

5.1 INTRODUCTION

Operations began at SNL (then Z-Division of LASL, Albuquerque Branch) in July 1945. SNL was 

originally created to perform ordnance engineering and assembly aspects of LASL design work. In 

essence, Z-Division was responsible for all nonnuclear components of nuclear weapons, whether 

through internal assembly or procurement. In late 1945, LASL began transferring its field testing and 

engineering organization, known as Z-Division, to Sandia Base near Albuquerque. Personnel from 

the Army Air Corps 509th Composite Group at Wendover Air Base in Utah joined the original group to 

assemble weapons. This organization formed the nucleus of Sandia Laboratory, created in 1948 as a 

separate branch of LASL. 

The following year, the laboratory formally separated from LASL when the University of California, the 

Los Alamos managing contractor, asked to be relieved of the responsibility. American Telephone and 

Telegraph (AT&T), at the request of President Truman, agreed to take over management of the 

facility, and Sandia Corporation, a wholly owned subsidiary of Western Electric and AT&T’s 

production arm, was formed to serve as the managing contractor [2]. 

By 1952, the weapons production complex was in place. SNL focused on weapons development and 

expanded its engineering staff to accommodate the growing number of weapons projects. In addition 

to design and production coordination, SNL undertook extensive field testing of components and 

supported the atmospheric tests sponsored by its partner laboratories. The bulk of the work at SNL 

and its properties has been related to the nonnuclear aspects of nuclear weapons design. This work 

includes weapons design and testing, production engineering, stockpile maintenance, and stockpile 

surveillance (Ullrich 1998).

From 1945 to the present, SNL employees have been involved in operations at a number of Sandia 

Corporation sites. These sites include Livermore, California; Hattiesburg, Mississippi; NTS; 

Clarksville, Tennessee; and Salton Bay Station, California. In addition, SNL employees spent time at 

other DOE facilities. Since 1949, when weapons storage sites were opened, until 1967, SNL 

stationed personnel at the storage sites to monitor, maintain, and assemble weapons [3]. Dosimetry 

monitoring records from these offsite activities may or may not be available in the SNL dosimetry 

records [4]. 

SNL operates major research nuclear reactors and electron/ion accelerators for the DOE Office of 

Military Applications. Research and development activities are conducted in relation to nuclear 

weapons systems, nonnuclear weapons systems, advanced nuclear reactors, simulation source 

development, and other basic and applied research areas at each facility [5].

The potential for chronic intakes at SNL is far less than at DOE production sites because of the nature 

of the tasks performed. Certain areas of the site are nonnuclear. In addition, the nature of the 

research environment at the SNL site results in intake potentials that are often unique and of short 

duration. Nevertheless, the potential for monitored and unmonitored intakes has existed throughout 

the history of the site [6]. 

Assembly

Assembly is used in a variety of contexts and conveys a variety of meanings, especially in the early 

history of SNL (Ullrich 1999). This can make it difficult to determine the exposure potential of a 

worker associated with statements of “assembly” of weapons in either the telephone interview or the 

other work histories [7]. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 52 of 120

Although the Little Boy gun-type nuclear weapon received some research and development attention 

after World War II, the majority of the early postwar nuclear weapons were implosion weapons based 

on the Manhattan Project Fat Man design. In this design, a sphere of high explosive surrounded a 

central core of nuclear material. Detonation of the high explosive compressed the central core of 

nuclear material to a supercritical mass. Introduction of neutrons to the highly compressed core 

initiated the nuclear chain reaction. To use a weapon, it had to be partially disassembled, the nuclear 

core inserted, and the weapon reassembled. Sealed-pit weapons with their nuclear components 

installed during assembly did not enter the stockpile until 1957. 

Due to the nature of these weapons, individuals using the term assembly for work during the 1940s 

and 1950s could mean a number of things. Assembly could mean the final assembly before a 

weapon was used or tested (i.e., inserting the nuclear core into the weapon). The latter use segued 

into the description of putting together inert mockups or prototypes of weapons (that is, assembling a 

weapon with neither high-explosive nor nuclear core). The term assembly was also used to refer to 

putting appropriate pieces together into components, leading up to final assembly of one of the other 

sorts. Therefore, terms similar to weapons assembly in the telephone interview or work history do not 

necessarily indicate potential exposure to radionuclides. 

Certain TA-I buildings, such as the first permanent facility built on the site, were not used for highexplosive lens assembly or nuclear core insertion; other facilities were available for that purpose. All 

evidence indicates that the first permanent TA-1 facility was designed and used as a mechanical test 

laboratory. As such, it represents the core of SNL’s early ordnance design mission (Ulrich 1999).

Nuclides

Nuclides with the widest historical and current application throughout the SNL facilities are:

• Tritium (3

H)

• Uranium (238U, 234U, 235U)

• Fission and activation products (e.g., 90Sr, 137Cs, 65Zn, 60Co, 182Ta)

and, to a lesser degree,

• Plutonium (238Pu, 239Pu, 240Pu, 241Pu)

• Americium (241Am) [8].

These radionuclides of primary concern are listed in SNL reports from 1945 to the present. The 

potential for exposure to nuclides existed in TAs and test facilities (Mallon 1995), as listed in 

Table 5-1. Additional descriptions of the TAs and potential hazards are in Section 2.0. If the specific 

work areas are known, the dose reconstructor should review Section 2.0, Tables 5-1 and 5-2, and 

available dose reconstruction references to determine the potential for exposure to internal hazards. 

Many of the buildings did not have a potential for routine intakes of radioactive material or were 

completely devoted to nonradioactive functions. Table 5-2 lists the significant potential worker 

exposure to radioactive materials in TA-III and TA-IV and the Explosive Component Facility (ECF). 

Many of the exposure histories and work records are not specific about the work areas to which 

individuals were assigned. However, if information about the work location is available, Table 5-1 can 

be used to determine the probable exposures.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 53 of 120

Table 5-1. Internal exposure potential by area (Mallon 1995) [9].

TA Uranium Tritiuma Plutonium Americiumb Other (Sr-90, Th, tracers, etc.)

TA-I

c DU Various 

buildings; 

erbium 

tritided in one 

building

Sealed 

Pu-238 heat 

sources, 

(nuclear 

materials 

receiving)

Environmental 

restoration 

sites

Environmental releases of 

accelerator activation products 

(O-15, N-13, Ar-41), machining 

of activated stainless-steel and 

fission products in one shop 

(VanDevender 1984), induced 

activity in one building, sealed 

sources in other buildings.

TA-IIe

(1948–

1960s)

Decontamination 

facility, melt 

facility, and 

waste facility

HT or HTO FP in a decontamination facility

TA-III DU, entire areaf Entire areaf Pu in 

contaminated 

wastesg

FP, Co-60, thoriumg in a mixed 

waste facility

TA-IVh Accelerator 

targetsi

Fe-59, Fe-55 activation 

products, aerosol when targets 

are machined.

TA-V Reactors and hot 

cellsj

Reactors and 

hot cellsj

Reactors and 

hot cellsj

Reactors and 

hot cellsj

FP from reactors

Coyote 

Test Field

DU, entire areaf

TTR DU, entire areaf,k Entire areaf,k Entire areab,f

a. Some facilities use tritiated solvents that may require special biokinetics.

b. As a portion of the plutonium mixture and as pure Am-241 (Stanley 1993).

c. Nonnuclear components were assembled by the Road group in TA-I for shipment to TA-II to complete subassembly..

d. Research suggests that the retention time of metal tritides (MTs) is longer than that of HTO (ORAUT 2007).

e. 1948 to 1952 primary site for weapons assembly (subassembly level, no nuclear pits in the weapon). Facilities for small 

explosive component research and testing in late 1950s to early 1960s. After 1960, the primary purpose of TA-II was 

research and testing of high-explosive components. Tritium was contained in test components, although it is unclear if 

this was the condition during the entire history of the site.

f. Also at environmental restoration sites and a mixed waste management facility.

g. From Section 2.0.

h. See Table 5-2 for details of specific areas in TA-IV.

i. An incident involving a leaking neutron calibration device in April 1991 produced tritium intakes (Burnett 1991).

j. Includes the hot cell laboratory, the Glovebox Laboratory (10 high-purity gloveboxes), and the analytical laboratory.

k. No bioassay is routinely done for uranium at TTR because workers are not likely to encounter 2 μCi of plutonium or 

10 μCi of DU. Baselines may have been performed (Potter 1993a).

Table 5-2. Facilities in TAs-III and -IV and the Explosive Component Facility.

Technical area Facility Radiation Remarks

TA-III For available facility 

details, see SNL 

(1998b).

No radioactive material is treated

For available facility

details, see SNL 

(1998c).

Low-level waste, mixed waste 1957–1988

Melt furnace contaminated with uranium 

(Tucker 1977)

Reactor experiment, activated materials 

(Tucker 1977)

Building with uranium stored (Tucker 1977)

For available facility 

details, see SNL 

(1998c).

Low-level waste, transuranic waste, mixed 

waste, sealed sources (stored or in stack 

monitors).

1988–present

treatment and 

packaging
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Technical area Facility Radiation Remarks

TA-IV For available facility 

details, see SNL 

(1998d).

External and O-15 and N-13 only, shortlived activation products (Zn-65, Co-57, 

Na-22, Mn-54)

40 wk/yr, 9-week 

operational setup 

and maintenance

For available facility 

details, see SNL 

(1998d).

External and activation products require a 

30-minute delay before entry and handling, 

maximum a few days for reuse. Sealed 

source Cm-244, Na-22

Items stored for 

decay before 

cleaning and reuse.

For available facility 

details, see SNL 

(1998e).

Some activation of Be-Cu materials and 

external.

For available facility 

details, see SNL 

(1998f).

External, no activation products are made.

For available facility 

details, see SNL 

(1998g).

External

For available facility 

details, see SNL 

(1998h).

External

For available facility 

details, see Jow (1991).

Ni-63 (<10 μCi each), Am-241 calibration 

sources, 

For available facility 

details, see Jow (1991).

Mixed activation materials and parts from 

NTS, storage for units containing <100 μCi 

Kr-85, <10 μCi Ni-63, and 1,000 μCi H-3.

For available facility 

details, see SNL 

(1998i).

Tritium; low level of water concentration. 

Nuclear targets; tritium, Pu-239, DU

Activated hardware, 50,000 kg

Neutron shots

200 mg Pu and DU

1,000 Ci tritium

For available facility 

details, see SNL 

(1998j).

No radioactive materials

For available facility 

details, see 

VanDevender (1984).

Activated stainless steel Aerosols generated 

during machining

Outside secured 

area

For available facility 

details, see SNL 

(1998k).

DU, C-14

Sealed sources

Millicurie quantities 

ECF ECF See SNL (1998l).

Many SNL workers received exposure and intake monitoring at sites other than SNL. Those 

attending nuclear tests and, to a lesser degree, nosetip tests may have had a potential for intakes. 

The names and dates of selected nuclear weapons tests (shots) and nosetip tests are listed in 

Table 5-3 to assist the dose reconstructor in identifying terms that may be referenced in personnel 

records or the telephone interview [10]. SNL employees visiting or temporarily assigned to other 

facilities, such as other laboratories, processing facilities and production facilities may also have had 

the potential for intakes. Offsite dosimetry records may or may not be in the SNL records, but are 

supplied directly from the appropriate sites [11]. The dose reconstructor is referred to the technical 

basis documents for the specific sites when it is necessary to evaluate potential exposure at sites 

other than SNL-NM or TTR.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 55 of 120

Table 5-3. Nuclear weapons and nosetipa tests (not all-inclusive) (SNL 1993a).

Test name Date Details

Crossroads Summer 

1946

Two test shots at Bikini Atoll to assess the effects of nuclear weapons on 

ships at sea. Shot Able was an air drop. Shot Baker was an underwater 

detonation.

Sandstone 1948

Hard Hat 02/16/1962 NTS, seepage through stemming

Marshmallow 06/28/1962 NTS, massive stemming failure, release continued for several days

Shoal 10/26/1963 Fallon, release of 110 Ci of Xe-131m, <1 Ci I-131.

Gumdrop 04/21/1965 NTS, controlled ventilation, primarily Xe-135

Diluted Waters 06/16/1965 NTS, massive stemming failure, gross fission products released

Tiny Tot 06/17/1965 NTS, seepage through shaft, Xe-138, Xe-135, Kr-87, Kr-88, 20 Ci of 

radioiodines. 

Red Hot 03/05/1966 NTS, release of noble gasses, I-135 (2,000 Ci), I-133 (500 Ci), I-131 

(20 Ci)

Pinstripe 04/25/1966 NTS, gross fission product release

Piledriver 06/02/1966 NTS, only Xe-135 detected in release

Double Play 06/15/1966 NTS, primarily noble gasses released

Derringer 09/12/1966 NTS, noble gasses, I-135 (152 Ci), I-133 (41 Ci), I-131 (1.5 Ci)

Newpoint 12/13/1966 NTS, primarily noble gasses

Midi Mist 06/26/1967 NTS, noble gasses and radioiodines in release

Door Mist 08/31/1967 NTS, noble gasses with radioiodines (I-135, I-133, I-131) and ruthenium 

(Ru-103 and Ru/Rh-106), primarily

Dorsal Fin 02/29/1968 NTS, no release

Milkshake 03/25/1968 NTS, primarily Xe-138

Diana Moon 08/27/1968 NTS, seepage, Xe-138 and radioiodines (I-135, I-133, I-131); later release 

of Xe-135, I-135 (3.6 Ci), I-133 (2.1 Ci), and I-131 (0.1 Ci)

Hudson Seal 09/24/1968 NTS, no release

Ming Vase 11/20/1968 NTS, no release

Cypress 02/12/1969 NTS, no release

Minute Steak 09/12/1969 NTS, release of Xe-138, Xe-135, Xe-133,I-135 (34.4 Ci), I-133 (3.4 Ci), 

and I-131 (0.05 Ci)

Diesel Train 12/08/1969 NTS, no release

Diana Mist 02/11/1970 NTS, no release

Mint Leaf 05/05/1970 NTS, release initially primarily of Xe-135, later of Xe-133m and Xe-133 and 

some radioiodines

Diamond Duet 05/12/1970 NTS, release of fission gasses, Xe-133m and Xe-133

Hudson Moon 05/26/1970 NTS, release of Xe-135, Kr-88, and Kr-85m

Diamond Mine 07/01/1971 NTS, no release

Grommet 07/1971–

06/1972

NTS

Diagonal Line 11/24/1971 NTS, primarily Xe-135, lesser amounts of Kr-85m, Kr-87, Kr-88, Xe-131m, 

Xe-132, Xe-133, Xe-133m, and trace I-131, I-132, I-133 and I-135 

Misty North 05/02/1972 NTS, no release

Toggle 07/1972–

06/1973

NTS

Diamond Sculls 07/20/1972 NTS, no release

Dido Queen 08/05/1973 NTS, no release

Husky Ace 10/12/1973 NTS, no release

Ming Blade 06/19/1974 NTS, no release

See SL (1977). 1974–1977 Contained Sr-90/Se-75/Ta-182 with external Ta-182 surface contamination 

(SL 1977)

Hybla Fair 10/28/1974–

01/06/1975

NTS, Xe-133 and Xe-133m

Dining Car 04/05/1975 NTS, no release

Husky Pup 10/24/1975 NTS, no release
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Test name Date Details

Mighty Epic 05/12/1976 NTS, no release

See SL (1975b) 1976 Contained Se-75, sealed Co-57 sources, and DU (ballast).

Hybla Gold 11/01/1977 NTS, no release

See Miller (1978) 1977–1980 External Ta-182 surface contamination (Miller 1978)

Diablo Hawk 09/13/1978 NTS, no release

Huron King 06/24/1980 NTS, no release

Miners Iron 10/31/1980 NTS, no release

Huron Landing 09/23/1982 NTS, Xe-133, Xe-133m, Xe-135, Kr-85m, Kr-88

Mini Jade 05/26/1983 NTS, Xe-133 and Xe-133m

Tomme/Midnight 

Zephyr

09/21/1983 NTS, no release

Midas Myth 02/15/1984 NTS, no release

Midas Rain 04/06/1985 NTS, Xe-133, Xe-133m, Xe-135

Mill Yard 10/09/1985 NTS, Xe-133, Xe-133m, Xe-135

Diamond Beach 10/09/1985 NTS, Xe-133, Xe-133m, Xe-135

Mighty Oak 04/10/1986 NTS, Xe-133 and 2.4 Ci I-131

Middle Note 06/20/1987 NTS, no release

Mission Ghost 06/20/1987 NTS, Kr-85

Mission Cyber 12/02/1987 NTS, no release

Misty Echo 12/10/1988 NTS, no release

Disko Elm 09/14/1989 NTS, Xe-133, Xe-133m, Xe-135

Mineral Quarry 07/25/1990 NTS, no release

Distant Zenith 09/19/1991 NTS, no release

Diamond Fortune 04/20/1992 NTS, Xe-133 and traces of I-131

Hunters Trophy 09/19/1992 NTS, no release

Nosetip tests had an intake potential different from that of weapons tests. Nosetips containing TaC 

rods were irradiated to produce 182Ta. The most significant hazard in these tests was external 

radiation, although there was some potential for intakes from surface contamination caused by tramp 

elements on surfaces before irradiation. Some instances of surface contamination caused by the 

irradiation of materials on the nosetip or heat shield have occurred during tests. The intake potential 

would be to a limited number of nuclides (182Ta is the most significant with the possibility of 75Se). The 

nosetips in these tests were typically prepared at SNL and shipped to Vandenberg AFB. A health 

physicist from SNL would travel to Vandenberg to perform the surveys on the shipment when it 

arrived and during the time the shipment was prepared for the test. The health physicist would take 

contamination surveys and external dose readings (Miller 1978). 

General Internal Dosimetry History

The SNL mission was primarily research, mostly focused on the effects of environmental factors and 

external radiation exposure on weapons materials and components. No potential for internal 

exposure was anticipated in many of the areas on the site [12].

The routine bioassay program was not formalized at SNL until after 1992. There is very little 

documentation about the internal dosimetry program or formal information about bioassay techniques 

or frequency before 1992. Some bioassay for tritium and uranium was performed beginning in 1949 

[13] and some air-sampling limits were found from as early as 1964. The Industrial Hygiene 

Department was responsible for bioassay from at least 1959 through 1991 (Potter 1993b), and 

possibly as early as November 1949 (Rarrick 2007; Thompson 2007). Tritium and possibly some 

uranium and plutonium bioassays were performed [14]. Fewer than 100 workers participated in the 

bioassay program each year from the beginning of the program through 1991. The bioassay records 

were transferred to the corporate archives as microfilm or microfiche and dose summary results were 

entered into an electronic database beginning in 1967 (Thompson 2007). Much of the in vivo
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 57 of 120

bioassay was outsourced until the SNL whole-body counter became operational in 1993. Although 

the number of individuals monitored has increased, not all individuals working at SNL are currently 

monitored.

Air sampling occurred infrequently, although health physics technician logs mention continuous air 

monitor (CAM) alarms and malfunctions. Records indicate that CAMs were in the Glovebox 

Laboratory of the SER facility (Author unknown, date unknown).

In 1991, a DOE Tiger Team assessment team identified several issues with the internal dosimetry 

program at SNL (SNL 1995a,b). The results of bioassay samples analyzed between August 1992 and 

April 1994 by Controls for Environmental Pollution (CEP, an offsite commercial laboratory) were 

invalidated based on the results of spike samples and the integrity of the data (DOE 1994; Hartman 

2007). Records were found for uranium lung counting performed for workers in 1989 and 1990 by 

Helgeson Scientific Services. A positive bias, based on results of unexposed workers, is indicated for 

these counts (Brake 1989) [15].

Tritium, gamma isotopic, and uranium in vitro bioassays are currently being performed on site, as is

whole-body counting. Monitored work areas include the plutonium areas (239Pu and heat source 

238Pu), uranium areas, tritium facilities, laboratory facilities, reactors, accelerators, and environmental 

restoration sites. 

Section 5.2 discusses in vitro methods for specific radionuclides. Section 5.3 discusses in vivo

bioassay methods used currently and historically at SNL. 

5.1.1 Bioassay Results of Individuals

There are few pre-1989 results of bioassay of individual SNL workers in the electronic records. An 

interview with a retiree who was involved in the archiving of early dosimetry records indicated that 

records of bioassay beginning in 1949 were entered into an electronic database when the bioassay 

results were being microfilmed [16]. Of the claimant records reviewed, most of the bioassay results in 

the records are from temporary assignments at sites other than SNL. Bioassay records before 1989 

have been found in paper and microfiche format in the SNL archives, but retrieval is labor-intensive. 

These records contain uranium, plutonium, and tritium bioassay results. In addition, it is likely the 

bioassay records for workers employed by SNL before November 1, 1949 [17], when the Z-Division 

officially separated from LANL, may only be in the LANL bioassay database. If employment dates 

before November 1, 1949, are listed for a claimant, dosimetry records should be requested from LANL 

if none are in the worker’s file. 

A 1989 appraisal of the Health Physics and Industrial Hygiene Departments indicates that there were 

no procedures for the Internal Dosimetry Program detailing employees on bioassay, frequency of 

sampling, and responsibility for results (Hyde 1989). The report indicated that there was good 

radiochemistry support to the bioassay analysis using documented procedures for nonquantitative 

uranium and possibly tritium (Preston 2005; Buvinghausen 2005; SNL 2005a,b). There is evidence 

that some workers may have been monitored by the Industrial Hygiene Department using breathingzone air (BZA) samples (Pigg 1993). However, it is not known if the results of the BZA samples are 

included with the individual dosimetry records provided by the DOE.

In vivo monitoring for uranium was performed by Helgeson Scientific Services at SNL as subsequently 

addressed (HSS 1989). Brake (1989) suggests that a large number of positive 238U and 235U positive 

lung counts reported by Helgeson Scientific Services could have been attributed to the acknowledged 

positive bias exhibited by the Helgeson system. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 58 of 120

CEP provided bioassay services for SNL; however, CEP bioassay results shall not be used to 

assess reconstructed doses based on the following discussion. Bioassay samples were sent to 

CEP from August 1992 to April 1994. Poor performance on blank and spike samples was noted early 

in that contract. An investigation in 1995 invalidated all these results (Hartman 2007). Approximately 

500 bioassay results were suspect (DOE 1994). Bioassay results and exposures were evaluated and 

a small number of individuals were resampled (Jones and Schwoebel 1995a,b). However, because 

the follow-up occurred as much as 2 years after the potential exposures, the results for analytes such 

as tritium and fission products would not have been representative of potential intakes. Some 

bioassay analyses for tritium were performed by SNL Industrial Hygiene or NTS during that period. 

Those results are valid. Before the contract was awarded to CEP, samples were outsourced to AtlanTech. That contract was dropped in favor of CEP because of poor performance. Therefore, results 

from Atlan-Tech should be suspect (Hallman and Potter 1992).

Tritium results for SNL personnel on assignment at LLNL could be reported in whole-body dose rather 

than as in vitro bioassay results [18]. The bioassay results from LLNL are available for dose 

reconstruction on request. Assignment of tritium dose for these individuals should follow the protocol 

established for workers permanently assigned to LLNL. This protocol for tritium dose is discussed in 

Section 5.2.3 and in Lawrence Livermore National Laboratory – Occupational Internal Dose (ORAUT 

2010a). 

5.1.2 The Bioassay Program 

Before 1992, the Industrial Hygiene Department was “in charge” of the bioassay (Potter 1993b) and 

administered the respiratory protection program. Individuals were monitored following suspected 

intakes prompted by incidents, high results of general air samples, CAM alarms, or nasal swipes 

(Hasenkamp 1961). The assumption was that the potential for intakes was low (Gonzales 1985) and 

that the particle sizes of dispersed radioactive materials were generally accepted as larger than 

respirable, as described in a conference call with a retiree (Martin 2005) and discussed in a uranium 

briefing (Jow undated). Administrative limits were based on the prevailing AEC/DOE guidance [e.g., 

ICRP Publications 2 (ICRP 1959) and 30 (ICRP 1979)]. Discussions with a retiree indicate that 

bioassays were performed for uranium, plutonium, and tritium as early as 1949 (Rarrick 2007; 

Thompson 2007). Bioassay samples may have been sent offsite for analysis.

The Final Headquarters Report on the Nuclear Safety Program Appraisal of the DOE Albuquerque 

Operations Office in June 1985 described the internal dosimetry program as follows (Gonzales 1985):

The SNLA has minimal need for an internal Dosimetry program and thus the program 

is virtually nonexistent. No whole body or lung counting is routinely being done, 

however urine sampling is scheduled for a few employees. The following findings are 

provided:

Finding #15: The most recent Internal Dosimetry procedure available was dated 1977. 

Some substantive changes had since been written in but the procedure had not been 

updated.

Finding #16: The quarterly collection of urine samples which are processed for tritium 

is ineffective. If workers are to be routinely monitored for possible tritium exposure 

most facilities would collect samples for tritium analysis at least weekly.

Finding #17: The quality assurance audit program is not documented and extremely 

limited in scope.
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Routine sampling was not a general practice. Bioassay was performed in response to suspected 

intakes of radioactive materials or if the worker exceeded an administrative threshold (Hasenkamp 

1961). Baseline bioassay was required of individuals with a history of exposure to radioactive 

material. Termination bioassay was performed when an individual no longer qualified to participate in 

the program because of changes in work conditions or termination of employment. Management 

practice was to determine the need for participation in a bioassay program based on the likelihood of 

exceeding dose guidelines. A 1992 internal memorandum on the participation of the staff of 

Department 6521 in the bioassay program suggests baseline bioassay followed by SPR staff testing

once after a scheduled maintenance activity and ACRR staff testing once after the next operation 

involving high surface contamination. The criteria for participating in further bioassay would be “based 

upon the likelihood of ingesting a 100 mR/yr burden, staff desire to be tested, results from the above 

mentioned one-time tests, knowledge of program costs, DOE requirements, and health physics 

advice” (Bryson 1992).

Security inspectors were required to participate in a routine bioassay program and have annual 

whole-body counts and urinalyses. If routine bioassay of radiation workers who performed hands-on 

work indicated positive exposures, required bioassays would be expanded to include security 

inspectors who could have been present when the exposures occurred. Baseline and termination 

bioassay were required (Hallman 1992b).

The SNL bioassay program was formalized beginning in 1992 in response to an assessment by the 

DOE Tiger Team. An assessment of the program in 1995 (Skrable 1996) further clarified aspects of 

the program that required refinement. This resulted in a revision to the Technical Basis Document for 

Internal Dosimetry at Sandia National Laboratories in 1998 (Mallon 1995). The requirements of 

10 CFR Part 835, “Occupational Radiation Protection,” were incorporated into the program.

1991 Tiger Team Assessment Findings

Table 5-4 lists key findings and action item responses from the 1991 assessment (SNL 1995a,b). 

These findings indicate that the internal dosimetry program was not extensive or centralized during 

the years immediately preceding the assessment. Aspects of the program existed but may not have 

been implemented site-wide or, for other reasons, were not consistently implemented. While this 

assessment captures only the conditions that existed in 1991 [during the transition of the program to 

the new DOE Order 5480.11 (DOE 1988)], no records have been located that indicate that the 

program was more extensive in earlier years, although health physics concerns were being addressed 

in individual areas. An example is a memo of record dated April 27, 1965, that details a trip to 

investigate health physics problems related to TRIGA reactors as related to the health physics 

operations of the Sandia Pulsed Annular Core Reactor (Tucker and Tucker 1965). Several of the 

findings relate to the delay in implementing changes specific to the new system; others are more 

generic. The DOE Headquarters report discussed above and the Tiger Team findings listed in 

Table 5-4 provide a summary of the program and possible shortcomings.

 Table 5-4. 1991 Tiger Team assessment key findings related to the internal dosimetry program.

Finding 

no. Description Action plan response

Closure

date

16 KF-SH-01; lack of compliance with DOE 

Orders for Radiation Protection; 14 of 24 

areas were in noncompliance. 

Establish a program and staffing to come into 

compliance. 

05/94

379 Radiation protection activities are being 

performed without definitive written 

guidance or rigorous professional 

oversight, resulting in inconsistency and 

noncompliance with DOE requirements.

A series of procedures was available by 6/91. 

Others were scheduled to be modified.

06/98
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Finding 

no. Description Action plan response

Closure

date

387 A formal permit system for control of 

radiation work has not been 

implemented.

Five new radiation protection programs 

developed. The programs include procedures 

specifying protective clothing requirements, 

respiratory protection requirements, 

completing RWPs, HP document control, use 

and approval of RWPs, and engineering 

design in the workplace. 

06/93

393 Personnel are permitted to enter and 

work in areas with a potential for 

airborne radioactive contaminants that 

have not been monitored, as required by 

DOE O 5480.11 (DOE 1988).

A new program for monitoring workplace air 

was developed. Equipment was to be installed 

in 1996. 

01/96

397 Glovebox gloves are frequently patched 

rather than replaced as industry practice 

dictates.

Modified procedure to survey for contamination 

and replace gloves if damaged.

09/92

400 ALARA program does not fully meet the 

requirements of DOE O 5480.11 (DOE 

1988) and other DOE guidance.

Procedure was being rewritten. 12/94

401 Personnel exposure files do not contain 

all the radiological information related to 

personal exposures and radiological 

working conditions as required by DOE 

O 5480.11 (1988).

The files will be expanded to include relevant 

data.

04/96

402 Internal dosimetry records are not 

generated and maintained as required 

by ANSI N13.6 (ANSI 1972) and DOE 

O 5480.11 (DOE 1988).

The Personnel Internal Radiation Dosimetry 

program was under development. A 

commitment was made to comply. 

[Noncompliance with ANSI 13.6 (ANSI 1972) 

was cited in an assessment in 1985 (Gonzales 

1985).]

08/96

405 Management has not implemented a 

system to ensure ES&H Department 

review, approval, and control of all 

radiation protection activities. 

Implement process to ensure ES&H 

Department review of activities and RWPs. 

12/91

452 Respiratory Protection Program does not 

comply with ANSI Z88.2-1980 (ANSI 

1980), DOE O 5480.11 (DOE 1988), and 

29 CFR 1910.134.

Approved new program and began quantitative 

testing for all negative pressure respirators in 

4/1991. However full implementation was not 

scheduled until 1995.

10/95

535 Surveillance of work area conditions 

during respirator use was not adequate. 

Sufficient documentation to support 

respirator selection and use did not exist. 

All planned items to be addressed. 08/95

536 The breathing air garment developed for 

the Tritium Research Laboratory had not 

been approved by DOE.

Ceased use until approval of the garment 

could be completed.

08/90

Current Internal Dosimetry Program 

SNL relies on engineering controls to prevent intakes. Bioassay is used as a confirmation that the 

engineering controls are functioning properly. An example is the annual whole-body counts received 

by radiological control technicians and waste handlers. Waste handlers also receive routine bioassay 

for tritium. No other routine programs are in place [19]. 

Beginning in 1994, workers were required to participate in the bioassay program related to their entry 

into posted areas. Table 5-5 lists the current requirements. Table 5-6 lists historical requirements 

found in the records. Baseline bioassay is not specifically defined, but is assumed to include either 
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the full set of bioassay and whole-body counts or the bioassay related to the area and type of work to 

be performed. Potter (1994) describes baseline sampling as “bioassay procedures for new workers 

who may be exposed at SNL to significant technologically enhanced levels of naturally occurring 

radionuclides or to synthetic radionuclides to which they were exposed prior to work at SNL.” Visitors 

were not required to submit baseline samples. 

Table 5-5. Internal dosimetry requirements (current) (Potter 1994).

Posted area Internal dosimetry requirementsa

Entry into controlled area or radioactive materials area None

Entry into radiation area Baseline bioassay (radiation workers only)

Entry into high radiation area Baseline bioassay (radiation workers only)

Entry into very high radiation area Baseline bioassay (radiation workers only)

Entry into contamination area Review by RPIDb to establish bioassay

Entry into high contamination area Review by RPID to establish bioassay

Entry into soil contamination area Review by RPID to establish bioassay

Entry into airborne radioactivity area Review by RPID to establish bioassay

Entry into tritium contamination area Review by RPID to establish bioassay

a. In addition, individuals who wear respiratory protection for radiological purposes or who perform contamination or 

airborne surveys and emergency response personnel shall meet the internal dosimetry requirements. Termination 

bioassay is required of anyone participating in the bioassay program. The program includes routine and job-specific 

sampling requirements. 

b. RPID = Radiation Protection Internal Dosimetry.

Table 5-6. Action levels for airborne radioactivity (μCi/mL).

Years

Action level

Alpha Beta Tritium Gamma Monitoring requirements Remarks

1945–1958 No information found

1959–1960 >10% MPC

1961 Bioassays if there is 

suspected uptake of material 

(Hasenkamp 1961)

Follow 

AEC 

Rules

1971 2,000–

100,000 dpm

Establish a Radiation Work 

Permit Area (O’Neal and 

Burnett 1971)

1994–1996 2.00E-12 

(SNL 1994a)

2.00E-10 

(SNL 1994a)

>2% of the DAC-hours (40 

DAC-hours) or >100 mrem

10 CFR 

Part 835

1997–

present

1.00E-12 

(SNL 1999)

1.00E-10 

(SNL 1999)

Program limits after 1992 were based on ICRP Publication 30 (ICRP 1979), including derived air 

concentration hour (DAC-hr) calculations, as the requirements of 10 CFR Part 835 were implemented 

(Potter 1994, 1996).

In 1999, a memorandum-to-file described the evaluation of the requirements of the routine bioassay 

program for TA-V. The tritium being produced by the pool was known to be 200,000 times less than

the concentration that would be required to produce a 1-DAC air concentration in the ACRR Hibay; 

therefore, the routine bioassay sampling for tritium was discontinued [20]. Routine semiannual 

bioassay for uranium and annual whole-body counting continued (Culp 1999).

5.2 IN VITRO BIOASSAY

The in vitro bioassay program historically included the nuclides listed in Table 5-1. This section 

contains a detailed discussion of the primary or more significant site radionuclides. Some of the 

bioassay samples were sent to outside laboratories. Formal procedures are available for 

radiochemistry results generated on the site after 1992. In most cases, MDAs are not provided for 
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any of the results in the early years. MDAs can occasionally be inferred from the results, but these 

are inferences only [21]. A Statement of Work for the contract laboratories and a review of the 

program in 1994 list the desired radionuclides and the requested MDAs (Vosburg 1993a). 

Current MDAs are available. These MDAs can be used for the program from 1992 to the present, but 

are not necessarily applicable to the 1945-to-1991 period. Because the Z-Division was attached to 

LANL before November 1, 1949, the LANL MDAs apply to bioassay results before that date. Because 

of the bioassay agreement with LANL in the early years, LANL MDAs could be applicable for missed 

dose calculations. In addition, because techniques used in the early years of operation were similar 

to those used at LANL and in the nuclear industry in general for uranium, tritium, and plutonium, 

MDAs from LANL techniques can be assumed [22]. 

5.2.1 Plutonium

Because of the nature of the operations at SNL, plutonium was not a major component of the potential 

intakes. Potential chronic missed intakes of plutonium should not be assumed unless there is 

evidence, through bioassay results or work history, that the individual was exposed to plutonium in a 

workplace or field situations [23]. Work with plutonium was done in gloveboxes or hoods [24]. 

Plutonium bioassay using alpha spectroscopy for the determination of 238Pu and 239+240Pu (alpha 

spectroscopy cannot differentiate 239Pu and 240Pu) was requested from the contract laboratory. 

Sample results from CEP between 1992 and 1994 are considered to be invalid and these results 

should not be used for determination of intake (Hartman 2007). However, if results of samples 

analyzed by CEP are in a worker’s record, the worker should be considered as monitored for the 

applicable period.

Respiratory Tract Absorption Type 

The absorption type assigned for plutonium isotopes is dependent on the chemical process used to 

produce the plutonium mixture. However, no information is available about the potential respiratory 

tract absorption type or age of mixtures encountered at SNL. Because the lung class W (Type M) of 

239Pu and 241Am was considered to result in limiting exposures to unknown mixtures, Class W (Type 

M) has historically been considered the default absorption at SNL for dose assessment (Mallon 1995). 

In practice, much of the plutonium would be considered Type S because of the nature of the material. 

Material actually dispersed as a consequence of an explosion would have been subjected to high heat 

and, therefore, would be in the oxide form. The particle size of the dispersed material was typically a 

larger size than respirable particles (Martin 2005). Plutonium-238 could have been encountered as a 

high-fired oxide in the heat source technologies (Holley 1967). Excretion rates could differ from 

typical Type S compounds (Hammond, Lagerquist, and Mann 1968). 

No other information about absorption types assigned to specific operations has been found. 

Therefore, the dose reconstructor should apply the absorption type most favorable to the claimant 

(ICRP 1994b). When applicable, internal dose due to potential exposure to plutonium strongly 

retained in the lung (Type Super S plutonium) must be evaluated (ORAUT 2010b).

Sample Collection Procedures

Routine bioassay sample collection procedures were not well established at SNL before 1992. Most 

samples were collected on a protocol similar to the early LANL protocol that used three 1-L bottles for 

a 24-hour or simulated 24-hour sample. Spot samples may also have been taken [25]. Routine or 

special bioassay performed before 1992 was handled by the Industrial Hygiene Department rather 

than the Health Physics Department. Much of the bioassay monitoring was for nonradiological 

compounds associated with the processes at SNL; radiological bioassay was performed as part of the 

set of analyses of the routine bioassay samples submitted by workers [26]. Archives of radiological 
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bioassay have been found dating back to the early years of operations (Rarrick 2007; Thompson 

2007). The Technical Basis Document for Internal Dosimetry at Sandia National Laboratories (Mallon 

1995) stated that urine bioassay was to be used to monitor routine intakes, but details about the 

frequencies and protocols associated with sample collection were not provided. Because exposure to 

plutonium at SNL was limited, extensive routine plutonium bioassay is not likely for most workers. 

Missed Intakes

Intakes of plutonium could occur from both acute and short-term chronic exposures. Chronic 

exposures may not be identified as incidents but can still result in a measurable burden of plutonium. 

However, long-term chronic exposures are not likely because of the nature of the tasks at SNL. If 

chronic intakes favorable to the claimant are assigned in the absence of positive bioassay results, 

these intakes should be assigned only for workers in TA-V or for those participating in waste-handling 

activities specifically involving plutonium [27].

Routine Sample Frequencies

In general, the need for air sampling, BZA samples, or bioassay appears to have been based on the 

amount of material in the process at the time and the perceived possibility of an intake [28]. The 

Technical Basis Document for Internal Dosimetry at Sandia National Laboratories (Mallon 1995) 

addresses the use of bioassay to evaluate and demonstrate compliance with NRC guidance on 

internal dose and monitoring requirements. While these exact criteria were not in place early in the 

history of SNL, the philosophy of monitoring only when there was a potential for exceeding the current 

guidelines appears to have been consistent throughout the history of the program [29]. 

According to a memorandum in 1993, individuals working with 2 μCi or more of plutonium would be 

required to participate in a bioassay program as directed by Internal Dosimetry (Potter 1993b).

A special sampling program would be initiated when a radiological incident occurred or a positive 

routine bioassay sample result was obtained that indicated the possibility of an unexpected dose of 

100 mrem Committed Effective Dose Equivalent (CEDE) or more (Bryson 1992).

Sample Analysis Procedures

Nondestructive analyses are currently performed on the SNL site. However, analyses requiring 

radiochemistry (e.g., plutonium) have been sent off site since 1992 (Vosburg 1993b). However, 

according to conversations with a retiree and records in the SNL archives, beginning in 1949 samples 

were analyzed at SNL for total plutonium alpha (Rarrick 2007; Thompson 2007). These samples 

were prepared as deposited or electroplated samples in the Industrial Hygiene chemistry laboratories 

and transferred to the counting laboratory for proportional counting and, in later years, nuclear track 

emulsion, type A (NTA) or alpha spectrometry counting. In addition, arrangements were made with 

LANL and commercial laboratories to process samples (Hallman 1992c). The responsibility for 

bioassay was transferred to the Radiation Protection Department in 1992. From 1992 on, plutonium 

bioassay was performed off site. All sample results analyzed by CEP from 1992 to 1994 are 

considered invalid as discussed above. Most in vitro bioassay samples are urine. Fecal and tissue 

samples are performed only if requested in special circumstances. 

If plutonium bioassay results are found in dosimetry records, these results could be labeled either as 

total plutonium or as 239Pu (which contains 240Pu) or 238Pu. Results labeled as 239Pu before 1967, the 

year when alpha spectroscopy was readily available in the industry, should be treated as total 

plutonium alpha unless accompanied by 238Pu results [30]. 

Detection Sensitivities and Reporting Limits

MDAs are not available for older bioassay results. MDAs in the “picocurie range” were achieved 

beginning in 1960 when the electroplated disks were counted using NTA counting (Rarrick 2007). 

From 1992 to the present, the MDA has been 0.05 pCi/L for isotopic plutonium offsite analysis (Potter 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 64 of 120

2006a). A 1993 Statement of Work for analytical services listed the required isotopic plutonium MDA 

as 0.0017 Bq/L (0.05 pCi/L) (Vosburg 1993b). Results are currently reported as the calculated value, 

positive or negative. When not specified, the MDA is considered to be two times the detection level 

(or uncertainty of the blank). MDAs may be reported with sample results before 1992. After 1992, a 

decision level (DC) equal to one-half the MDA may be reported with the sample results. 

Because there was a Memorandum of Understanding (MOU) with LANL in the early years, after ZDivision was no longer directly controlled by LANL, the MDAs for plutonium analysis are included in 

Table 5-7, which lists the MDAs with reporting limits from LANL (Hallman 1992c). Bioassay 

performed for employees of the Z-Division during the period that the Z-Division was part of LANL 

would have used the applicable LANL procedures (Rarrick 2007). MDA levels for bioassay performed 

at SNL before 1992 are not generally available unless listed with specific sample results. However, 

analytical techniques and instrumentation similar to those used at LANL were used at SNL; therefore, 

the MDAs would probably have been similar (Rarrick 2007).

Table 5-7. Plutonium bioassay MDAs as listed in LANL procedures and reports.

Nuclide Year(s)

Sample 

typea Technique (“era”)

MDA level Reporting limit

Unit/24-hr sampleb Unit/24-hr sampleb

Pu-Total 

alpha

1944 U Cupferronc 0.7 pCi (Clark 2005)d >0.8 pCie

1945–1949 U Cupferronc 0.16 pCi to 0.05 pCif

1 dpm (0.45 pCi) DL 

(Moss 1990)

>0.8 pCie

1949–1/1957 U Bi-phosphate/alpha 

counting

0.20 pCi to 0.07 pCif 2 dpm or 0.9 pCi

0.4 pCie

1/1957–1965 U Aluminum nitrate/NTA 0.03 pCig 0.2 dpm or 0.09 pCi

1966 U ZnS 0.03 pCih (0.07 dpm)

Pu-239 1967–1991 U RAS or PHA only starting 

in 1971

0.03 pCii

1977–1981 F RAS (PHA) 1 nCi/sample (less if 

Am-241 ratio known)

1981–1983 F Phoswich detector, 4-cm 

sample thickness

0.4 nCi /sample or 400 pCi/ 

sample

(17-keV X-rays)

Pu-238 1967–1971 U RAS (alpha PHA) 0.03 pCii 0.2 dpm/24-hr 

investigate

1971–1976 U RAS (alpha PHA) 0.03 pCii

1977–1991 U RAS (PHA) 0.03 pCii

1977–1981 F RAS (PHA) 0.4 nCi/sample

1981–1983 F Phoswich detector, 4-cm 

sample thickness

0.2 nCi/sample, (17-keV 

X-rays)

a. U=urine, F=fecal.

b. Unless otherwise noted.

c. “A successful method of analyzing urine was developed in Jan. 1945 but could not be used as a routine test until a 

contamination-free laboratory (ML Building) was ready for use in Feb. 1945” (Hempelmann 1945).

d. Not adjusted for potential chemical recovery.

e. Results above these values were considered “high” (i.e., positive) and subject to statistical investigation. Source: 

Lawrence (1978). 

f. Background count rate 1 cpm (changed to 0.1 cpm at some unknown time before 1957), 1,000-minute count time, 50% 

efficiency, average recovery 82.3% ±19.4% (1945 to 1949) and 67% ±21% (Nov. 1949 to Jan. 1957).

g. Sources: Campbell et al. (1972); McInroy et al. (1991). 

h. Source: Moss et al. (1969).

i. Source: UC (1978).

Plutonium results from LANL will most likely be reported in activity per 24-hour sample. However, 

results provided by commercial laboratories are expected to be reported as picocuries per liter. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 65 of 120

Therefore, if volume units are not provided with the sample results and the results are not indicated as 

provided by LANL or SNL, the dose reconstructor should assume the per-liter concentration. 

Plutonium Isotopic Ratios (Mixtures)

There is no definitive historical information on the 240Pu:239Pu atom ratios of SNL sources with the 

exception of a pure 238Pu [31] source term (Holley 1967), and there is no information on how the ratios 

vary with time and location.

Exposure to heat source technology pure 238Pu (238PuO2) may be suspected if positive 238Pu results 

are encountered when the associated 239Pu results are below the level of detection. A plutonium 

mixture that can be used when pure 238Pu from heat source technology is encountered is 238Pu (81%), 

239Pu (15%), 240Pu (2.9%), 241Pu (0.8%), and 242Pu (0.1%) (Holley 1967). The disassembly of a 238Pu 

heat source in a glovebox was discussed in a SER logbook in 1972 (SL 1974). Table 5-8 lists the 

activity and weight ratios for heat technology plutonium mixture [32]. It is likely that pure 238Pu would 

be fresh during processing in the heat source technologies. However, intakes of aged pure 238Pu may 

be possible during decontamination or decommissioning activities in areas where pure 238Pu was 

processed. Table 5-8 includes the activity ratios necessary to calculate the appropriate mixture for 

aged pure 238Pu. Potential ingrowth of 241Am can be determined from the ratios in Table 5-8 (Holley 

1967). Weapons-grade (6%) and fuel-grade (12%) aged mixtures are listed in Tables 5-9 and 5-10. 

Table 5-8. Pure 238Pu isotopic mixture (Holley 1967).

Mixture designation

Fresh 5 yr 10 yr 15 yr 20 yr 30 yr

Years of aginga

0 5 10 15 20 30

Component

Weight fraction 

(fresh) Specific activity (μCi/g of mixture)

Pu-238 0.81 1.39E+07 1.33E+07 1.28E+07 1.23E+07 1.18E+07 1.09E+07

Pu-239 0.15 9.30E+03 9.30E+03 9.30E+03 9.30E+03 9.29E+03 9.29E+03

Pu-240 0.029 6.58E+03 6.58E+03 6.58E+03 6.57E+03 6.57E+03 6.56E+03

Pu-241 0.008 8.31E+05 6.53E+05 5.14E+05 4.04E+05 3.17E+05 1.96E+05

Pu-242 0.001 3.93E+00 3.93E+00 3.93E+00 3.93E+00 3.93E+00 3.93E+00

Am-241 NA 0 5.90E+03 1.05E+04 1.40E+04 1.68E+04 2.05E+04

Pu-239+

Pu-240

0.179 1.59E+03 1.59E+03 1.59E+03 1.59E+03 1.59E+03 1.59E+03

Activity ratiosb,c

Pu-238/

Pu-239

1,490 1,430 1,380 1,320 1,270 1,180

Pu-238/

Pu-240

2,100 2,020 1,950 1,870 1,800 1,670

Pu-238/

Pu-241

16.7 20.4 24.9 30.5 37.3 55.7

Pu-238/

Pu-242

3.52E+06 3.38E+06 3.25E+06 3.13E+06 3.01E+06 2.78E+06

Pu-238/

Am-241

NA 2,230 1,220 876 704 532

Pu-238/

Pu-239+Pu-240

8,710 8,370 8,050 7,740 7,440 6,870

a. Time since separation of Am-241 from the plutonium mix.

b. Calculate dose for Pu-239 and Pu-240 separately using these ratios.

c. Pure Pu-238 was not on site at SNL until after the introduction of radiometric alpha spectroscopy (RAS) for analysis of 

bioassay samples. Therefore, the activity ratios from total alpha plutonium are not applicable.

Table 5-9 lists activity and weight ratios as referenced for DOE sites for 6% weapons-grade 

plutonium. These ratios are based on the presumed age of the material for DOE sites. Table 5-10 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 66 of 120

lists the relative activities of plutonium isotopes and 241Am, which grows in from 241Pu (Gallaher and 

Efurd 2002), in 12% 240Pu mixtures (Carbaugh 2003). In these tables, aging refers to the time since 

the 241Am was separated from the plutonium.

Table 5-9. Activity composition of reference weapons-grade (6%) plutonium mixture.

Isotope

Mixture designation

Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aginga

0 5 10 15 20 25 30

Specific activity in mixture (Ci/g)

Pu-238 8.56E-03 8.23E-03 7.91E-03 7.60E-03 7.31E-03 7.03E-03 6.75E-03

Pu-239 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02

Pu-240 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02

Pu-241 8.24E-01 6.48E-01 5.09E-01 4.00E-01 3.15E-01 2.48E-01 1.95E-01

Pu-242 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06

Am-241 0 5.83E-03 1.04E-02 1.39E-02 1.66E-02 1.87E-02 2.03E-02

Pu-239+240 7.13E-02 7.13E-02 7.13E-02 7.13E-02 7.12E-02 7.12E-02 7.12E-02

Pu-alpha 7.99E-02 7.95E-02 7.92E-02 7.89E-02 7.85E-02 7.83E-02 7.80E-02

Total alpha 7.99E-02 8.53E-02 8.96E-02 9.28E-02 9.52E-02 9.70E-02 9.83E-02

Activity ratios

Pu-239+240:Am-241 N/A 12.2 6.87 5.13 4.28 3.80 3.50

Pu-239+240:Pu-238 8.33 8.67 9.01 9.38 9.74 10.1 10.5

Pu-239:Pu-240 4.24 4.24 4.24 4.24 4.24 4.24 4.24

Pu-241:Pu-239+240 11.6 9.09 7.15 5.62 4.42 3.48 2.73

Pu alpha:Pu-239+240 1.12 1.12 1.11 1.11 1.10 1.10 1.10

Pu alpha:Pu-238 9.33 9.66 10.0 10.4 10.7 11.1 11.6

Pu alpha:Am-241 NA 13.6 7.62 5.68 4.73 4.19 3.84

Pu-241:Pu alpha 10.3 8.15 6.43 5.07 4.01 3.17 2.50

a. Time since separation of Am-241 from the plutonium mix.

The nature of the process work at SNL, however, involved research-grade plutonium mixtures usually 

considered to be fresh. While the plutonium mixture encountered in these processes should be 

assumed to be fresh, aged plutonium could have been encountered during building decommissioning 

and environmental restoration and in other areas where plutonium or plutonium contamination could 

have been in place for many years. Any 241Am observed in lung counts performed years after the 

intake can be assumed to be the result of the in-growth of 241Am from the 241Pu in the mixture over 

time or the 241Am in the initial plutonium mixture unless the incident report specifically indicates 

potential exposure to pure 241Am or urine bioassay for 241Am is found for the intake period. The most 

probable type of mixture encountered by the worker should be assumed to be a 6% plutonium mixture 

unless other information is available in the dosimetry records [33].
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 67 of 120

Table 5-10. Activity composition of reference fuel-grade (12%) plutonium mixture.

Isotope

Mixture designation

Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aginga

0 5 10 15 20 25 30

Specific activity in mixture (Ci/g)

Pu-238 1.71E-02 1.64E-02 1.58E-02 1.52E-02 1.46E-02 1.40E-02 1.35E-02

Pu-239 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.25E-02

Pu-240 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.71E-02 2.71E-02

Pu-241 3.09E+00 2.43E+00 1.91E+00 1.50E+00 1.18E+00 9.29E-01 7.30E-01

Pu-242 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06

Am-241 0 2.19E-02 3.89E-02 5.22E-02 6.24E-02 7.03E-02 7.63E-02

Pu-239+240 7.98E-02 7.98E-02 7.98E-02 7.97E-02 7.97E-02 7.97E-02 7.97E-02

Pu-alpha 9.69E-02 9.62E-02 9.56E-02 9.49E-02 9.43E-02 9.37E-02 9.32E-02

Total alpha 9.69E-02 1.18E-01 1.35E-01 1.47E-01 1.57E-01 1.64E-01 1.69E-01

Activity ratios

Pu-239+240:Am-241 NA 3.64 2.05 1.53 1.28 1.13 1.04

Pu-239+240:Pu-238 4.67 4.86 5.05 5.24 5.46 5.69 5.90

Pu-239:Pu-240 1.93 1.93 1.93 1.93 1.93 1.93 1.93

Pu-241:Pu-239+240 38.7 30.5 24.0 18.8 14.8 11.7 9.16

Pu alpha:Pu-239+240 1.21 1.21 1.20 1.19 1.18 1.18 1.17

Pu alpha:Pu-238 5.67 5.87 6.05 6.24 6.46 6.69 6.90

Pu alpha:Am-241 NA 4.39 2.46 1.82 1.51 1.33 1.22

Pu-241:Pu alpha 31.9 25.3 20.0 15.8 12.5 9.91 7.83

a. Time since separation of the Am-241 from the plutonium mix.

5.2.2 Americium

At SNL, 241Am is usually encountered as a trace contaminant in plutonium. However, there is a 

potential for exposure to pure 241Am. 

Americium-241 has always been included in the bioassay profile at SNL, but the specific source term 

driving the inclusion is not known [34]. However, documents have been found indicating that a pure 

241Am source term does exist and has existed at SNL [e.g., a spill of 241Am was reported to have 

occurred on or about March 26, 1992 (SNL 1993b)]. There is an indication that workers submitted 

samples for the americium bioassay program only if there was a potential for exposure to pure 

americium. Therefore, plutonium mixtures should not be inferred from americium bioassay results if 

no plutonium results are associated with the americium bioassay. Conversely, if the selected 

plutonium mixture, based on plutonium bioassay results, indicates the presence of americium, the 

absence of americium bioassay should not preclude the calculation of the dose from the americium 

contribution to the plutonium mixture. 

Encounters with 241Am could have occurred in relation to certain experiments or the handling of 

leaking calibration sources. 

The potential for chronic exposure to pure 241Am or 241Am in a plutonium mixture at SNL is not routine. 

Potential missed chronic intakes of pure 241Am should not be assigned unless there is evidence that 

the worker was at least intermittently exposed to 241Am at SNL. This evidence would be from 

bioassay results or work histories [35]. 

Sample Frequencies

At present, no routine sampling programs are in place for 241Am. A special sampling program would 

be initiated if a radiological incident occurred or a positive routine bioassay sample result indicated the 

possibility of an unexpected dose of 100-mrem CEDE or more (Bryson 1992).
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 68 of 120

Minimum Detectable Activities 

Since 1992, bioassay for 241Am has been sent to offsite laboratories. All reported results have been 

listed as actual values, positive or negative. If not specified, 2 times the detection level is considered 

the MDA. The current MDA requested from offsite laboratories is 0.05 pCi/L (Vosburg 1993b; Potter 

2006a). No record of MDAs before 1992 is available but, because of the MOU with LANL to perform 

bioassay analysis, LANL MDAs can be used if bioassay results without MDAs are found before 1992. 

Table 5-11 lists LANL MDAs.

Bioassay results listed as having been performed by CEP between 1992 and 1994 should be 

considered invalid (DOE 1994; Hartman 2007). However, if results of samples analyzed by CEP are 

in a worker’s record, the worker should be considered as monitored for the applicable period. 

Table 5-11. Americium-241 LANL bioassay techniques and MDAs.

Sample 

type Year Method MDA

Urine 1954–1957 Unknown 9.0E-01 pCi/24 hra

Urine 1958–1982 Chemical extraction/proportional counting 2.0E-01 pCi/24 hrb

Fecal 1977 Phoswich 4.0E-02 nCi/sample

Urine 1983c–1991 Coprecipitation/alpha spectroscopyd 1.5E-02 pCi/24 hr

Fecal 1983c Am/Pu screening/Phoswich 1.0E-02 nCi/sample

a. No MDA available, use derived investigation level; tolerance level 3.1 pCi/sample.

b. Source: McClelland (1958); method can carry over thorium, plutonium, curium, actinium, and 

neptunium. Exact end date to the start of this MDA is not known.

c. Source: Gautier (1983); exact start date of the MDA is not known.

d. Source: Inkret et al. (1998). 

5.2.3 Tritium

A 1989 review of health physics and industrial hygiene functional areas listed limited tritium handling 

as one of the main potential source terms at SNL (Hyde 1989). Applying potential missed dose based 

on chronic intakes of tritium is reasonable if work may have involved tritium exposure. 

Tritium was encountered in several forms: tritium oxide (tritiated water; HTO), tritiated gas (HT), 

organically bound tritium (OBT), and metal tritide (MT). Each form has unique characteristics. If the 

form is not indicated or the work area is not specific, then, with the exception of MTs, dose 

reconstructors should assume HTO, which is the form generally encountered (SNL 1990). 

Smear surveys of a TA-I accelerator facility in 1972 indicated significant tritium contamination on tools 

and target area parts. Some smears indicated more than 5 × 106 dpm of tritium on parts and 

surfaces, especially those related to the target. Contamination was also found on the floors of the 

area. TA-I includes other potential sources of tritium intakes in areas of neutron generator work or 

known use and storage of tritium [36]. A smear survey of a TA-V reactor plenum indicated a 

maximum of 77 dpm/swipe (SNL 1989a). Tritium incidents have been recorded for accelerator areas 

[37]. 

The primary method of limiting uptakes of tritium was engineering controls, including proper 

ventilation. Smearable tritium of 2,000 dpm or less was considered a clean area. Smearable 

contamination of 2,000 dpm to 100,000 dpm required that contamination control procedures be 

observed and radiological work permit (RWP) areas established. Above 100,000 dpm, a Radiation 

Danger Zone was established (O’Neal and Burnett 1971). 

Organically Bound Tritium

The first approximation of the dose from OBT (labeled) compounds is the tritium in body water dose. 

However, the absorption, distribution, and excretion of tritium-labeled compounds are specific to the 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 69 of 120

chemical and physiological behavior of the particular compound. Specific guidance is available in 

ORAUT (2007) if the dose records indicate an exposure to tritium-labeled compounds.

Metal Tritides

Tritium exposures in the form of stable metal tritides (SMT) aerosols were possible. The compounds 

include the chemical hydrides and dihydrides of hafnium, erbium, titanium, zirconium, and other 

metals. In TA-I there was a potential for metal tritide exposure in one building, with the predominant 

form being erbium tritide. One TA-II facility also presented the potential for exposure to SMTs, which

were used in neutron generators, although the potential for intakes from these sources was limited 

(SNL 1990).

If it is indicated that an energy employee performed work with a potential for exposure to SMTs, such 

as work with neutron generators or on components in the ECF, SMT exposure should be evaluated in 

the dose reconstruction. As with other forms of tritium, specific guidance is available in ORAUT 

(2007) if the dose records indicate an exposure to SMT.

Other Sources of Potential Tritium Intakes

Noteworthy isolated tritium exposure incidents have occurred. Tritium swipe surveys have been 

performed consistently in many of the areas with a potential for tritium contamination. Table 5-12 lists 

these incidents, including surveys that identified areas of contamination, though other incidents at 

SNL are likely to have occurred.

Table 5-12. Other sources of potential tritium intakes.

Time Description Intake Comments

3/9/71 Tritium survey of TA-I neutron generator manufacturing areas

indicated measurable tritium on general surfaces.

Unknown (SL 1975a)

6/22/73 Tritium contamination on work area surfaces in TA-I neutron 

generator manufacturing areas, up to 7,142,000 dpm/smear on 

the source and 574,000 dpm/smear on a tray. 

(SL 1975a)

5/30/73 Memorandum discusses the potential for intakes while 

handling the 50 Ci of tritium in a TA-I building.

(SL 1975a)

8/7/73 Tritium contamination on work area surfaces in a TA-I neutron 

generator manufacturing area.

(SL 1975a)

1/15/76 Swipe surveys in two TA-V accelerator buildings show levels 

up to 26,000 dpm.

(SL 1974)

1989 Tritium bioassay results annual report indicate several workers 

with sample results >MDA.

(Hallman 1990)

1991 The front sights of the M-16 rifles carried by SNL security 

inspectors contained 3 to 6 mCi of tritium. It was discovered 

that several of these rifles had leaking tritium cartridges. As a 

result, several security inspectors had positive urine bioassays 

(Ball 1991).

Because of this incident 

security officers received 

tritium bioassay at SNL 

and TTR.

1994 Thermocouples contaminated with tritium; 17,500 dpm 

removable.

None Most tritium fixed on 

surface

Analytical Techniques

All results since 1992 are likely to be reported as actual values, positive, negative, or zero (SNL 

2005a,b). The reported MDA for onsite tritium bioassay analysis is 1,000 pCi/L (0.001 μCi/L). A 

Statement of Work for analysis from an outside vendor requested a desired MDA of 9,000 pCi/L 

(Vosburg 1993b). The MDA before 1992 is unknown, but a report of all tritium results for 1989 

indicates values of 0.01 μCi/L as the lowest value reported above 0.0. It appears that the analyses 

were performed at SNL (Hallman 1990). This agrees with the MDA reported by LANL at that time. 

LANL could have provided some analytical services to SNL before 1992. Typical MDAs reported by 

LANL before 1992 are listed in Table 5-13.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 70 of 120

Table 5-13. Tritium urine bioassay MDAs from LANL.

Time Detection level Reporting level Counting method

1950–1951 5.18E4 Bq/24 hra (1 μCi/L) Electroscope

1952–1953 5.18E4 Bq/24 hra (1 μCi/L) Geiger-Mueller counter

1954–1957 5.18E4 Bq/24 hr (1 μCi/L)

1958–1968 5.18E4 Bq/24 hr (1 μCi/L) Internal Geiger-Mueller

1969b–1987 1.04E3 Bq/24 hr (0.02 μCi/L)b 5.18E4 Bq/24 hr (1 μCi/L) LSC

1988–1992 5.18E2 Bq/24 hr (0.01 μCi/L) 5.18E3 Bq/24 hr (0.1 μCi/L) LSC and 1 mL raw urine

a. Expected to be the same as 1954.

b. Source: Gautier (1983).

Bioassay results listed as having been performed by CEP between 1992 and 1994 should be 

considered invalid (DOE 1994; Hartman 2007). However, if results of samples analyzed by CEP are 

in a worker’s record, the worker should be considered as monitored for the applicable period. 

Routine Sampling

No information is currently available on historical routine sampling programs. At present, only waste 

handlers are on a routine tritium bioassay program. The SNL protocol assumed that persons working 

in areas where there was a potential for tritium exposure were monitored on an as-needed basis. 

Historically, individuals likely to exceed 0.25 rem (1/20 of 5-rem annual dose from AEC requirements) 

would require further investigation and control. This would correspond to a single acute dose 

producing a urine concentration at time zero of 23 μCi/L. As an administrative control, individuals with 

urine bioassay of 20 μCi/L would be excluded from tritium work until their bioassay reached 10 μCi/L 

(O’Neal and Burnett 1971). 

A special sampling program would be initiated when a radiological incident occurred or a positive 

routine bioassay sample result was obtained that indicated the possibility of an unexpected dose of 

100-mrem CEDE or more (Bryson 1992).

Assignment of Tritium Dose from Monitoring at LLNL

Tritium monitoring results for SNL workers temporarily assigned to LLNL were typically reported in 

dose rather than as bioassay results. The assigned annual dose can be entered into the POC 

calculation directly. However, this process overestimates the total measured tritium dose because the 

historical assessment methodology would use a 12-day effective half-life for tritium and a beta linear 

energy transfer value of 1.7, both of which would lead to a higher dose than the current methodology. 

When evaluating potential LLNL exposure, use the guidance of the LLNL occupational internal dose 

TBD (ORAUT 2010a). The tritium bioassay data are available by request from LLNL [38]. 

5.2.4 Uranium

Historically, uranium at SNL was DU, natural uranium (NU), or enriched uranium. Table 5-14 lists 

generic uranium conversion factors derived from Rich et al. (1988). Table 5-15 lists conversion 

factors for NU. A DU Briefing (Jow undated) states that DU is considered ≥99.75% 238U and ≤0.25% 

235U. Conversion factors differ among DOE sites because the fractional activity of the 234U differs in 

the enrichments. Thus, 0.25% enriched DU at SNL could have a different amount of 234U than the 

same enrichment at a different DOE site, depending on the source of the DU and the processing. 

The particle size was considered to be typically nonrespirable (i.e., > 10 μm AMAD) (Jow undated). 

However, not all uranium at SNL was of nonrespirable particle size [39]. DU was machined in the 

Toxic Shop in TA-I until about 1985 and after that in another TA-I facility. The machining process is 

capable of producing respirable aerosols (VanDevender 1984). 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 71 of 120

According to Hyde (1989), DU machining operations represent one of the main potential source terms 

at SNL. Most particles are considered to be ≥1 mm, which is much larger than the 10-μm size 

considered to be respirable (Jow undated). However, oxides produced by high-energy events, such 

as penetration of hard armor, can produce some fraction of aerosolized (<3 μm) particles. Some 

explosion and implosion experiments (e.g., Nuclear Engineering Safety Technologies) resulted in 

approximately 11% aerosolized particles. Reactor experiments resulted in less than 1% aerosolized 

particles. Areas of known contamination were, and are, posted and require monitoring for 

contamination before exit or direction by radiation protection before entry into the area. 

Table 5-14. Uranium conversion factors.

Enrichment 

percent

Fraction by activity Fraction by mass Total

U-234 U-235 U-238 U-234 U-235 U-238 U pCi/μg

0.1 0.225 0.010 0.765 0.00001597 0.001 0.99898 0.438

0.17 0.260 0.010 0.730 0.00002 0.0017 0.999 0.465

0.2 0.285 0.010 0.705 0.00002178 0.002 0.997978 0.476

0.3 0.340 0.010 0.650 0.00003 0.003 0.997 0.514

0.5 0.410 0.020 0.570 0.00004 0.005 0.995 0.591

0.7 0.480 0.020 0.500 0.00005 0.007 0.993 0.668

0.711 0.486 0.022 0.492 0.00005 0.007 0.993 0.672

0.72 0.486 0.022 0.492 0.00005 0.007 0.993 0.675

3 0.745 0.045 0.210 0.00019 0.030 0.970 1.57

90 0.970 0.029 0.001 0.010 0.900 0.090 62.1

93 0.972 0.027 0.001 0.010 0.930 0.060 65.1

95 0.973 0.026 0.001 0.010 0.950 0.040 67.2

Table 5-15. Natural uranium.

Isotope U-234 U-235 U-236 U-238 Total

Weight fraction 0.0000537 0.0072 0 0.99274

Specific activity (pCi/μg) 0.33369 0.01557 0.33369 0.68296a

Fraction of total activity 0.4886 0.0228 0 0.4886

a. As listed in the Integrated Modules for Bioassay Analysis (IMBA) computer program, Version 4.0.9.

Fuel-coolant interaction experiments using corium thermite, which contains DU, ended in TA-V in 

1983. The corium thermite material used in these experiments was prepared using dust respirators 

as respiratory protection. While most of the material recovered after the experiments was >10-μm in 

diameter, preparation of the material involved working with DU in the powdered form (Marshall 1993).

Sampling Protocol

Historically and currently, there has been no routine sampling program for uranium. A special 

sampling program would be initiated if a radiological incident occurred or a positive routine bioassay 

sample result indicated the possibility of an unexpected dose of 100-mrem CEDE or more [40] 

(Bryson 1992). Bioassay sampling was initiated immediately after a suspected intake incident and 

continued for a “few” days [41]. According to a 1993 memorandum, individuals handling 10 μCi or 

more of DU would be required to participate in the bioassay program (Potter 1993b).

There is no preservative in the sample collection bottle, so an acid wash of the bottle is required to 

collect uranium that plated out on the surface of the bottle during sample transport and storage [42]. 

Also, because no workers participate in a routine uranium bioassay program, there is no protocol for 

routine sampling on a Friday or Monday (Potter 2006b). Friday sampling would include the promptremoval fraction and Monday sampling (after a weekend away from the site) would typically not. 

Uranium Analysis Techniques

Techniques performed at SNL include fluorometry, kinetic phosphorescence analysis (KPA), and 

inductively coupled plasma mass spectroscopy (ICP-MS). Until 2003, uranium was analyzed by 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 72 of 120

fluorometry using KPA or other fluorometric methods typical of the available instrumentation of the 

era. Fluorometry is based on the nonquantitative heavy-metal content in the urine. According to the 

Health Physics Department appraisal (Hyde 1989), a quantitative analysis was not available to 

evaluate positive samples. In 2003, KPA was replaced by ICP-MS, which can provide isotopic 

information as well as total uranium. 

Analytical Sensitivity

A discussion with a retiree indicated that bioassay for uranium was being performed by the Industrial 

Hygiene Department as early as 1950 [43]. Results of some of these bioassays have been found in 

SNL archives. Information on the analytical sensitivities for urine bioassay before 1992 is sporadic. 

Results identified for workers include blank and standard results. Table 5-16 lists the known MDAs. 

Because there was an MOU for analysis with LANL in the early years, LANL MDAs before 1975 can 

be assumed for the calculation of missed dose [44]. Samples could have been analyzed at contract 

laboratories other than LANL. After 1975, the MDAs in Table 5-14 are derived from analyses 

performed on site at SNL-NM or by a contract laboratory. 

Table 5-16. Routine uranium urinalysis MDAs.

Period Methoda MDA

Decision 

level

Reporting 

levelb

1949–1967 Fluorophotometric (DU or NU) None listedc 50 μg/L >100 μg/L

1968–2/1976 Fluorophotometric (DU or NU) 4 μg/Lc,d U

3/1976–1978 Fluorophotometric (DU or NU) 1 μg/Ld U

1949–1954 Anion exchange/gross alpha counting 

(possibly used)

25 dpm/L >100 dpm/L

1955–1971 Extraction/alpha proportional counting 

(U-234 alphas measured)

50 dpm/Lc,e 50 dpm/L >100 dpm/L 

3/1971–1974 Extraction/alpha proportional counting 15 dpm/Lc >100 dpm/L

1975–2002 Total uranium–KPA or other fluorometric 0.1 μg/L

2003–present Total uranium–ICP-MS 0.1 μg/L (0.05 μg/L 

usually achieved)

1993–present Alpha spectroscopy (U-233, -234, -235, 

-238) outside vendor

0.0033 Bq/L 

(0.089 pCi/L)

a. Method listed.

b. Exceeding reporting levels required investigation and evaluation (Lawrence 1984).

c. Lawrence (1984).

d. 50 μg/L considered positive indication of NU material in the body (Dummer and Barker 1958).

e. Specific for U-235 and U-233, 50 dpm/24-hr sample considered positive indication of enriched uranium in the body 

(Dummer and Barker 1958). Use 50 dpm/L as MDA because no other information is available. 

Bioassay results listed as having been performed by CEP between 1992 and 1994 should be 

considered invalid (Hartman 2007). However, if results of samples analyzed by CEP are in a worker’s 

record, the worker should be considered as monitored for the applicable period. 

Environmental Uranium 

New Mexico is known for high levels of NU in the soil and groundwater. Some SNL workers could 

have lived in areas of particularly high NU concentrations ranging from 0 to 4 Bq/L (108 pCi/L) in 1992

and up to 6 Bq/L (162 pCi/L) in 2001 (Little, Miller, and Guilmette 2003a). These areas of high 

concentration are primarily in the Espanola area. 

A 1992 study (Little, Miller, and Guilmette 2003b) listed the average drinking water concentrations for 

the area of Los Alamos, White Rock, and Santa Fe as 0.015 μg/L (0.01 pCi/L or 0.00037 Bq/L).

A study was conducted by SNL Internal Dosimetry personnel to determine the background range and 

mean of the background range of urinary excretion for nonradiation workers at SNL (SNL 1993b). 

That range was determined to be between 0.07 and 0.26 μg/L with a mean of 0.16 μg/L [45].
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 73 of 120

The bioassay results reported for individual workers do not have dietary uranium subtracted. 

However, decisions to classify a sample as positive for occupational exposure were based on the 

sample exceeding the expected urinary excretion rate of levels in nonradiation workers at SNL (SNL 

1993b). Urine samples with results above background are reanalyzed by alpha spectrometry (Potter 

1993c). Samples that remain above background are evaluated to determine if the uranium is natural 

or depleted and resampling is initiated if the results are above 0.3 μg/L (Potter 1993c). Uranium 

bioassay was performed at SNL-NM as early as 1949 [46]. As of 1993, SNL-NM had not identified 

any intakes of depleted uranium; only natural uranium had been identified. Natural uranium is not 

currently considered an occupational intake by the site. Natural uranium was on the SNL-NM site 

from at least 1967 through the mid-1990s. For the purposes of the dose reconstruction program, all 

detected uranium is included in the dose reconstruction.

5.2.5 Fission and Activation Product Analysis

SPR (sometimes referred to as Kiva, which was the building housing the reactor), SER, and other 

reactors have operated in TA-V. Examples of fission and activation products in plant air samples, 

effluents, and primary coolant have been taken from various reports (SL 1967; SNL 1988, 1989a,b). 

An incident at the SPR-II reactor in about 1968 necessitated the evacuation of control room personnel 

when fission products released from the pulse migrated to the control room. The design was later 

modified to prevent this leakage after the pulse [47]. In addition, activation products were produced 

from photoactivation of metals in locations such as PBFA-II. Photoactivation of stainless-steel parts 

produces long and short half-life radionuclides with specific activities that can be microcurie/gram 

quantities. Handling can be delayed to permit the decay of short half-life products, such as isotopes 

of aluminum and copper. However, the ferrous isotopes (59Fe, half-life 45 days) and other activation 

products, such as 54Mn (half-life greater than 300 days), can become an aerosol when these parts are 

machined. These radionuclides are listed in Table 5-17.

Urine bioassay is currently performed on the site by gamma spectroscopy for fission and activation 

products (SNL 2004, 2005c). A 1993 Statement of Work for an outside laboratory lists the required 

MDA for mixed fission products based on 137Cs as 0.83 Bq/L (22.4 pCi/L) (Vosburg 1993b). The 

current urine bioassay MDA for 137Cs is 11 pCi/L. Bioassay results listed as having been performed 

by CEP between 1992 and 1994 should be considered invalid (Hartman 2007). However, if results of 

samples analyzed by CEP are in a worker’s record, the worker should be considered as monitored for 

the applicable period. Whole-body counting was also performed for fission and activation products as 

addressed in Section 5.3.

Table 5-17. Fission and activation product nuclides.

Type Radionuclide

Limiting 

DAC μCi/mL Type Radionuclide

Limiting 

DAC μCi/mL

Fission gas Kr-85m Sb-124

Xe-133 Activation products W-187

Xe-135 2E-5 Ag-110m

Ar-41a Ta-182

Fission products Zr-95 Mn-54

I-131 2E-8 Mn-56b

Ru-103 Na-24b

Ru-106 Cu-64 b

Ba/La-140 5E-7 Cd-115b

Y-94 Al-28b

Mo-99/Tc-99m 6E-5 Co-56

Zr-97/Nb-97 5E-7/ 3E-5 Co-57

Rb-89 Co-58

I-132 3E-6 Co-60b

I-133 1E-7 Zn-65
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Type Radionuclide

Limiting 

DAC μCi/mL Type Radionuclide

Limiting 

DAC μCi/mL

I-134 Sb-122

I-135 Ni-57

Sr-91 1E-6 Sn-117m

Y-91m 7E-5 Sn-119m

Y-92 3E-6 Ta-182

Cs-136 Ta-183

Cs-137 As-76

Cs-138 Fe-59b

Ce-141 3E-7 Be-7

Ce-144 Other U-235

a. Main contributor to activity at SER.

b. From SER (SL 1967), PBFA-II (VanDevender 1984).

Strontium

Records of either routine or special 90Sr urinalyses are sparse. It appears that records of 90Sr analysis 

indicate that samples were sent to an outside laboratory. Since 1992, the requested MDA for 90Sr as 

total strontium analysis is 5 pCi/L (0.17 Bq/L). For isotopic strontium, 90Sr, and 89Sr, the desired MDA 

is 9 pCi/L (0.33 Bq/L). If results from a contract laboratory are available before 1992, these MDAs can 

be applied if no other MDAs are listed. LANL did not perform strontium analyses regularly; therefore, 

it is unlikely that samples would have been sent to LANL for analysis. Bioassay results listed as 

having been performed by CEP between 1992 and 1994 should be considered invalid (Hartman 

2007). However, if results of samples analyzed by CEP are in a worker’s record, the worker should 

be considered as monitored for the applicable period. 

A source term for 90Sr or 89Sr other than mixed fission products from reactors has not been identified. 

5.2.6 Accelerator Areas

The accelerator areas, such as Hermes III, present unique nuclides in addition to potential tritium 

exposures. Air activation products of 15O and 13N are produced. These nuclides have extremely short 

half-lives of 122.2 seconds and 9.97 minutes, respectively. Entry into areas with airborne 

contamination is restricted until the area had been ventilated for at least 20 minutes after shutdown. 

Therefore, occupational intakes of these nuclides are not expected to be of concern. Activation 

products can be produced in metallic materials as described for the PBFA-II (VanDevender 1984).

Tritium from targets was of greater concern; tritium is discussed in Section 5.2.3. 

5.2.7 Indoor Radon

Indoor radon is considered an incidental exposure at SNL. Air samples are counted immediately after 

sampling and 4 days later. In the reviewed reports and logs, most of the air filters were at background 

by the fourth day, which indicates that the initial activity was primarily radon daughters. However, the 

levels of radon are not considered above ambient levels encountered in general work areas [48]. 

In 1991, the DOE Indoor Radon Study was published (DOE 1991). Several areas of SNL 

Albuquerque were included in the study. Results ranged from 0.3 pCi/L to 7.6 pCi/L. The highest 

value was in a TA-I machine shop. The next highest value was recorded in a TA-V facility as a 

concentration of 3.1 pCi/L. All other locations sampled were 2.2 pCi/L or less. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 75 of 120

5.2.8 Other Limited-Exposure Radionuclides

SNL has always been a center for research. As such, small-scale use (in terms of either the number 

of persons involved or the activity of the source) of various radionuclides not addressed in previous 

sections has occurred throughout the history of SNL. Little or no documentation has been found on 

bioassay for these nuclides. These nuclides, as discussed below, should be assessed only if there is 

an indication that the worker had a potential for exposure to the respective nuclide.

Tantalum-182

Tantalum-182 is mentioned in several reports. Tantalum-182 is a gamma emitter and can be 

detected in gamma spectroscopy of urine samples or in a whole-body count. 

Selenium-75

Selenium-75 is mentioned in several reports. Selenium-75 was typically used as a sealed source in 

payload operations between 1974 and 1980. These sources contained quantities of more than 2 Ci. 

Intakes were not likely unless there was an encounter with leaking sources. Selenium-75 is a gamma 

emitter and can be detected in gamma spectroscopy of urine samples or in a whole-body count. 

Thorium 

Bioassay samples are sent off site for thorium isotope analysis. The current requested MDA for alpha 

spectroscopy analysis is 0.05 pCi/L. In 1993, a Statement of Work to an outside laboratory lists a 

desired MDA of 0.0033 Bq/L (0.09 pCi/L) (Vosburg 1993b). Results are expected to be reported in 

units of pCi/L. However, Table 5-18 lists the specific activity of thorium isotopes for use as necessary. 

Table 5-18. Specific activity of thorium isotopes.

Isotope Specific activity (pCi/μg)

Th-228 8.1946E+08

Th-230 2.0184E+04

Th-232 1.0966E-01

Thorium operations at SNL began in 1959. Thorium-232, as an aged oxide, was containerized and 

sealed in 1996 and 1997.

Neptunium-237

Neptunium-237 was a requested nuclide in the Statement of Work for an outside laboratory in 1993 

(Vosburg 1993b). The desired MDA using alpha spectroscopy was listed as 0.0017 Bq/L (0.05 pCi/L). 

However, 237Np is not a known source term in any of the site areas. 

Polonium-210

Incidents of a leaking 210Po source have been reported. A leaking 210Po source was handled by a 

worker in 1968. A bioassay sample was submitted (O’Neal 1968). Other potential source terms are 

not known. If 210Po bioassay results (before 1960) are found in a worker’s records, the detection 

levels from LANL may be applicable because of the MOU for bioassay. 

Detection limits for routine 210Po urinalysis are listed in Table 5-19. A procedure is listed for 210Po in 

urine in Gautier (1983). The 1958 detection levels can be considered to continue through the end of 

210Po bioassay. 

Table 5-19. Routine 210Po urinalysis MDAs.

Period MDA Recheck Tolerance

1943–1952 440 dpm/24 hra

1953 50 dpm/24 hra

1954 10 dpm/L 100 dpm/L 500 dpm/L

1955–1957 0.1 pCi/L
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1958 100 dpm/L 500 dpm/L

a. Source: UC (1979).

5.3 IN VIVO MINIMUM DETECTABLE ACTIVITIES, ANALYTICAL METHODS, AND 

REPORTING PROTOCOLS

In the early years, SNL had an MOU with LANL to perform whole-body and lung counting (Hallman 

1992a,b; SNL 1993c). However, it is not known how many workers actually participated in the 

program. It appears that any whole-body counts that occurred were performed because of suspected 

intakes and not as part of a routine monitoring program for workers with a potential for intakes of 

fission and activation products. In vivo counting equipment and techniques were developed in the late 

1950s and have been in routine use at LANL since at least 1970 and possibly as early as 1960. 

There is some indication that some of the counts recorded between the beginning of the program in 

1955 and the 1960s were performed for development of the program rather than actual suspected 

intakes. Counts during this period should be evaluated as closely as possible for validity in the dose 

reconstruction.

In vivo monitoring for uranium was performed by Helgeson Scientific Services in 1989 and possibly in 

1990 and 1991 (HSS 1989). The results of these counts were above the detection level for many of 

the counted TA-V staff (HSS 1991; SNL 1992, p. 4). A positive bias has been suggested in the 

Helgeson results (Brake 1989).

The whole-body counter program was established at SNL in 1993. The whole-body counting system 

is the commercially produced Canberra Accuscan II system with two hyperpure germanium (HPGe) 

detectors. This is a vertical shadow shield with two scanning germanium detectors. The system is 

capable of measuring the entire torso or specific areas. Currently, only Radiological Control 

Technicians and waste handlers receive annual whole-body counts to confirm the adequacy of the 

personal air sampling program (Mallon 1995, p. 26).

5.3.1 Whole-Body Counters–LANL (1955 to 1992)

When applying LANL bioassay results to SNL exposure, use the guidance of the LANL occupational 

internal dose TBD (ORAUT 2009a) for measurement and reporting parameters and protocols.

5.3.2 Whole-Body Counter–SNL

The SNL whole-body counter is a Canberra Accuscan II shadow-shield two-detector HPGe wholebody counter. The counting protocol was established for 137Cs, 60Co, and 40K sensitivities. MDAs 

reported in 2005 are listed in Table 5-20 and did not vary significantly from 2001 to 2005. Results 

reported in 1994 are approximately half the MDA reported beginning in 2001. The routine count time 

has remained the same since 1993. During that period, the counting system was relocated and the 

analytical software was upgraded, but no explanation for the change in MDA was provided [49].

Table 5-20. MDAs for whole-body counting, 

2005 (Reese 2006).

Radionuclide MDA (nCi)

K-40 140

Co-60 8.5

Cs-137 12

Table 5-21 lists MDAs reported for a typical 10-minute whole-body count. These are averages of 

MDAs reported for three workers evaluated in 1994 and the results of a single typical individual in 

2001, considering the MDA as twice the LC.
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Table 5-21. MDAs (nCi) for whole-body counting (SNL 1994b, 2001).

Radionuclide 1993–2000 2001–present Radionuclide 1993–2000 2001–present 

Be-7 49.0 107 U-235 Not reported 208a

Na-24 4.72 6.86 Gd-153 Not reported 87.4

Mn-54 5.45 11.2 Ir-192 Not reported 13.9

Co-56 5.37 8.84 Tl-201 Not reported 93.6

Co-57 13.2 29.8 Hg-203 Not reported 17.7

Ni-57 5.76 10.5 Tl-207 Not reported 3660a

Co-58 4.00 7.98 Tl-208 Not reported 37.0

Co-60 5.54 6.27 Pb-210 Not reported 5940a

Zn-65 8.10 19.0 Pb-211 Not reported 320a

Zr-95 8.27 15.3 Bi-212 Not reported 139a

Mo-99 11.3 57.8 Pb-212 Not reported 36.8

Ru-103 5.48 10.8 Bi-214 14.8 24.6

Ru-106 56.8 82.6 Pb-214 Not reported 32.2

Ag-110m 5.07 8.10 Rn-219 Not reported 140a

Cd-115 19.6 24.8 Ra-223 Not reported 127 a

In-115m 11.7 24.8 Ra-224 Not reported 414

Sb-122 7.71 11.7 Ra-226 Not reported 546a

Sb-124 5.31 8.82 Th-227 Not reported 137a

Sb-125 17.9 33.2 Ac-228 Not reported 39.4

I-131 6.75 13.2 Ra-228 Not reported 75.0

Ba-133 9.89 18.3 Th-228 Not reported 356a

Cs-134 4.96 9.24 Th-229 Not reported 210a

Cs-137 5.78 9.90 Pa-231 Not reported 770a

Ba/La-140 20.8 36.8 Th-231 Not reported 1660a

Ce-141 19.8 46.0 Th-232 Not reported 116

Pr-143 11.5 18.6 Pa-233 Not reported 28.8

Ce-144 94.3 212 Np-237 Not reported 252a

Nd-147 56.9 64.6 U-238 Not reported 554a

Eu-154 21.3 37.2 Pu-239 Not reported 36,800a

Eu-155 54.7 125a Cm-243 Not reported Not reported

Ta-182 15.0 26.8 Na-22 Not reported 5.58

Ta-183 25.3 156a Cr-51 Not reported 121

Am-241 97.5 186a Mn-52 Not reported 7.50

Fe-59 Not reported 14.8 Sr-85 Not reported 12.4

Ce-139 Not reported 24.8 Y-88 Not reported 4.38

Eu-152 Not reported 89.0 Ag-108m Not reported 8.10

a. The Accuscan II is a whole-body counter that is not optimized for low-energy photons or for the geometry typically 

associated with chest counts. However, these nuclides are reported as MDA (or possibly as positive results) in 

individual whole-body count reports. The reported MDAs should not be considered the primary method of assessing 

missed or bounding dose for these radionuclides if other types of bioassay results are available.

The nuclides listed in the MDA section of the whole-body count report do not necessarily indicate that 

the worker was potentially exposed to the entire list of nuclides. To determine if potential chronic 

exposure should be assumed, it is necessary to determine if the worker’s tasks resulted in the 

potential for intake. It is not reasonable to assume that a worker was exposed to all nuclides listed in 

Table 5-21. In addition, for overestimates, Table 5-22 lists nuclides that produce the highest dose to 

organs of interest based on the reported MDAs for that period. The organ of interest varies because 

the reported nuclides and the MDAs have changed over the history of the whole-body counting 

program. If it is necessary to assess missed dose, the nuclide can be used for the listed organs.
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Table 5-22. Nuclides that produce the highest dose to organ of interest.

Organ

Nuclide/type (MDA in nCi)

1971–1984 1985–1992 1993–1999 2001–present

Adrenals Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Urinary bladder Co-60/M (1.8) Ce-144/M (24.2) Ru-106/F (43) Ru-106/F (98.2)

Brain Co-60/M (1.8) Ce-144/M (24.2) Ru-106/F (43) Ru-106/F (98.2)

Gall bladder Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Kidneys Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Liver Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Muscle Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Ovaries Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Pancreas Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Testes Co-60/M (1.8) Ce-144/M (24.2) Ru-106/F (43) Ru-106/F (98.2)

Thyroid Co-60/M (1.8) Ce-144/M (24.2) I-131/F (6) I-131/F (14.8)

RBM Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Bone surface Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Stomach Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

SI Co-60/M (1.8) Ce-144/S (24.2) La-140/M (21) Eu-154/M (24)

ULI Co-60/M (1.8) Ce-144/S (24.2) La-140/M (21) Mo-99/S (71.5)

LLI Co-60/M (1.8) Ce-144/S (24.2) Ce-144/S (98) Ce-144/S (201)

Skin Co-60/M (1.8) Ce-144/M (24.2) Ru-106/F (43) Ru-106/F (98.2)

Spleen Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Uterus Co-60/M (1.8) Ce-144/M (24.2) Ru-106/F (43) Ru-106/F (98.2)

ET Co-60/M (1.8) Ce-144/S (24.2) La-140/M (21) Ce-144/S (201)

Lung Co-60/M (1.8) Ce-144/S (24.2) Ce-144/S (98) Ce-144/S (201)

Colon Co-60/M (1.8) Ce-144/S (24.2) Ce-144/S (98) Ce-144/S (201)

ET1 Co-60/M (1.8) Ce-144/S (24.2) La-140/M (21) Mo-99/S (71.5)

ET2 Co-60/M (1.8) Ce-144/S (24.2) La-140/M (21) Ce-144/S (201)

LN(ET) Co-60/M (1.8) Ce-144/S (24.2) La-140/M (21) Ce-144/S (201)

bbsec Co-60/M (1.8) Ce-144/S (24.2) Ce-144/S (98) Ce-144/S (201)

BBbas Co-60/M (1.8) Ce-144/S (24.2) Ce-144/S (98) Ce-144/S (201)

Gonads Co-60/M (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Breast Co-60/S (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Heart wall Co-60/S (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Thymus Co-60/S (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

BB Co-60/S (1.8) Ce-144/S (24.2) Ce-144/S (98) Ce-144/S (201)

AI Co-60/S (1.8) Ce-144/S (24.2) Ce-144/S (98) Ce-144/S (201)

LN(TH) Co-60/S (1.8) Ce-144/S (24.2) Ce-144/S (98) Ce-144/S (201)

Esophagus Co-60/S (1.8) Ce-144/M (24.2) Eu-154/M (24) Eu-154/M (53.2)

Table 5-23 lists nuclides that do not represent a significant potential for chronic intakes. These 

nuclides were included in the whole-body counting analysis library because they can be encountered 

at the accelerators. However, accelerator workers do not routinely participate in the whole-body 

counting program.

Table 5-23. Accelerator nuclides 

in whole-body count library.

Nuclide Half-life

Be-7 53.3 d

Co-56 78.7 d

Ni-57 1.5 d

In-115m 4.49 hr

Cd-115 2.23 d
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Sb-122 2.7 d

Nd-147 11.0 d

Ta-183 5.1 d

5.3.3 Cesium-137 Intakes from Fallout

Most workers in the early days of whole-body counting had detectable activities of 137Cs. Most of this 

was attributed to fallout. Some workers had even higher levels of 137Cs from consumption of wild 

game. An LC used to establish the difference between occupational and nonoccupational sources of 

137Cs and other fallout radionuclide intakes has not been discovered in the records. In lieu of other 

information, the guidance in Table 5-24 can be used.

Table 5-24. Mean United States 137Cs body burdens from fallout.

a

Year Body burden (nCi) Year Body burden (nCi)

1953 0.27 1966 9.7

1954 1.1 1967 5.6

1955 2.2 1968 3.5

1956 4.3 1969 2.7

1957 5.1 1970 2.7

1958 6.5 1971 2.7

1959 8.1 1972 2.7

1960 6.8 1973 2.7

1961 4.6 1974 1.6

1962 6.0 1975 1.1

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

a. Source: NCRP (1987).

A 137Cs result should be considered occupational if the same measurement detected other fission or 

activation products or if a fission or activation product or radiostrontium urinalysis showed detectable 

activity and the sample was obtained in a reasonable time before or after the whole-body count or 

within the period between the previous and next whole-body counts. The reasonable time is based on 

the biological retention pattern of the radionuclide in the body.

NCRP Report 94 provides mean body burdens of 137Cs for the United States for the years most likely 

to produce interference with occupational whole-body count results (NCRP 1987). Those values are 

listed in Table 5-24. If no other fission or activation products are detected and the 137Cs result is less 

than the values in Table 5-24, the 137Cs result can be assumed to be due to fallout. All other fission or 

activation products identified in the whole-body or lung count should be considered occupational 

unless specifically stated otherwise in the records provided for the individual.

5.3.4 Lung Burdens 

Currently and historically, there is no chest counting system at SNL. With the exception of lung 

counts for uranium performed by Helgeson in 1989 and 1990, workers requiring lung counts would be 

sent to LANL. A 1992 memorandum indicates that this was done for at least one 241Am incident 

(Hallman 1992c). It also indicated that a worker was scheduled for an LANL whole-body count 

following a suspected plutonium intake (Crites 1967). Therefore, the LANL chest counting system is 

discussed here. The report from the Helgeson Scientific Services counts is also discussed below.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 80 of 120

Chest Counting at LANL

When applying LANL bioassay analysis to SNL dose reconstructions, apply the guidance of the LANL 

occupational internal dose TBD (ORAUT 2009a) for measurement and reporting parameters and 

protocols.

Chest Counting by Helgeson Scientific Services

Documents indicate that in 1989, 1990, and 1991 a mobile chest-counting system was brought to the 

SNL site to perform chest counts of workers (HSS 1989, 1991). Ninety-eight whole-body counts, 60 

lung counts for DU and 235U, and 18 lung counts for 239Pu were performed in 1989. The positive 

results were two whole-body counts for 137Cs, 59 lung counts for DU, and 13 lung counts for 235U. No 

positive counts were recorded for 239Pu. There is no indication of the MDA for these counts or a 

description of the counting equipment used. Brake (1989) suggests that the Helgeson counting 

system might have been exhibiting a positive bias during that period.

5.3.5 Wound Monitoring

No procedures for gamma counting of wounds have been found although wound counting is 

mentioned in the program overview (Potter 1994) and the Sandia National Laboratories Radiation 

Protection Internal Dosimetry Technical Basis Document (Mallon 1995).

5.4 AIR CONCENTRATIONS

5.4.1 Respiratory Protection Program 

Although the SNL Z-Division originated at LANL, the respiratory protection program at SNL is not 

clearly documented to reflect the LANL program. As late as 1993 (Marshall 1993), the use of “dust 

respirators” is listed as the protection for mixing powders containing DU. While documents have 

suggested that respirators were used during certain activities, protection factors should not be applied 

in the dose reconstruction because their use by specific individuals is uncertain. 

Contamination Surveys

Records of area contamination surveys have been found. A resuspension factor of 1 × 10-6 is usually 

considered when determining the potential air concentration from loose surface contamination [50].

Contamination limit action levels in 1961 were listed as >500 cpm of alpha per 60 cm2 or 1 mR/hr 

beta-gamma fixed or any removable (Hasenkamp 1961). A counting efficiency of 50% is usually 

assumed for alpha counting. A technique for taking smears was discussed in the SERF Health 

Physics Handbook (Devlin 1964) and limits were 100-to-200 cpm/1 ft2

. This was equivalent to 3,000 

to 6,000 dpm/ft2 given collection and counting efficiencies. Loose contamination was cleaned up 

promptly because of the adverse impact it could have on experiments. 

The Safe Operating Procedure of January 23, 1984, for a TA-I machine shop lists the limits for 

uncontaminated materials as <10 cpm removable alpha or <200 cpm removable beta-gamma. 

Materials containing plutonium were not permitted in the shop (VanDevender 1984).

A procedure for contamination surveys was issued in 1993 (Shanks 1993). Result flags on smear 

survey reports circa 1994 indicate that actions were taken in relation to the critical level 

(approximately half the MDA) and the 2-sigma error. Action levels in the 1990s are listed as 20 dpm 

alpha and 1,000 dpm beta-gamma for smearable contamination (SNL 1995c, p. 4). 

Continuous Air Monitors

Documents indicate that alpha and beta/gamma CAMs were in use in areas where airborne 

radionuclides were expected (SL 1967). Documents also indicate that standard maximum permissible 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 81 of 120

concentrations (MPCs) were observed for appropriate contaminants (SL 1967). In addition, CAMs 

were placed near the point of release where the air concentration would be at a maximum as a 

backup detection device. These were in addition to the general area CAMs. For areas involving 

gloveboxes, other methods of detecting possible breaches of integrity were in place (Gonzales 1985). 

Gloveboxes are operated with negative pressure and inert atmospheres. Oxygen sensors detect the 

presence of the smallest leak [51]. 

CAM alarms were historically used as an indication that room conditions had changed and that 

respiratory protection was required. The selection of the location of the CAM was critical to the 

reliability of the response (Whicker et al. 1997). The typical location of general area CAMs at the 

ventilation exhaust point of a room was not optimal. Results suggest that when a worker caused a 

release and was at or near the release point, the worker could be exposed for a significant period 

before the released radioactivity reached the CAM. 

The above limitations suggest that, without other interventions, the possibility exists that workers 

could be exposed to intakes that did not trigger alarms.

5.4.2 Incident Intake Information

Intake parameters can be derived from airborne contamination levels for buildings with the highest 

exposure potential or highest intakes for various periods. Examples of incidents and intakes are listed 

in Table 5-25. Average airborne contamination levels either are derived as simple averages or are 

reported as averages listed in reports. Simple averaging is assumed for SNL reports, but no 

information on the methods used to obtain these reported averages is available.

Table 5-25. Reported exposure incidents and results.

a

Date Incident

06/04/1967 Contamination of a laboratory area in TA-I; water leaking on tritium-contaminated equipment, 

2,440 dpm/ft2 (SL 1975a).

11/19/1967 Positive bioassay for Pu-239 from TA-I for an individual; 1.5 times maximum permissible body 

burden (Crites 1967).

3/1968 to 

4/1968

Positive bioassay for tritium for workers; maximum 18 μCi/L. No incident discussed (Crites 

1968). 

3/13/1968 to 

3/25/1968

Division 4233 individual with 27-μCi/L tritium urine sample; last sample 11.7 μCi/L (Crites 1968).

3/08/1968 Pu-238 “spill,” particle found on CAM air filter in TA-I (Hudson 1968).

11/22/1968 Leaking Po-210 source transferred between laboratories. No intakes confirmed (O’Neal 1968).

1/14/1970 Pu-238 spill; a leak discovered in a glovebox during cleaning. Contamination of the area in TA-I 

(SL 1975a). Employees submitted bioassay samples.

9/21/1983 Be-7, Zn-65, and Ag-110m found on Radeco air filters in accelerator area of TA-V (SNL 1985).

4/12/91 Leaking neutron calibration device containing gaseous tritium in area. Approximately 150 

concerned employees requested bioassay.

3/26/1992 Am-241, spill location unknown, urine bioassay 1.38E-1 pCi/sample, listed as false positive.

3/30/1993 Am-241 contamination >10,000 dpm on tray in TA-I; area not posted as Controlled Area. Tray 

in use for 10 years (Byers 1993). 

a. While an attempt has been made to report only incidents with quantitative results, some incidents, for which only 

qualitative comments were available in the records, have been included in the list. Information is not available for every 

year of operation.

6.0 OCCUPATIONAL EXTERNAL DOSE

6.1 DOSIMETERS USED

The external dosimetry records for workers employed by SNL before November 1, 1949, when the ZDivision officially separated from LANL, may be only in the LANL database [52]. If employment dates 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 82 of 120

before November 1, 1949, are listed for a claimant, dosimetry records should be requested from LANL 

if none are in the worker’s file. Beta/gamma and neutron dosimeters used at SNL over the years are 

listed in Tables 6-1 and 6-2. SNL switched from film to thermoluminescent dosimetry after its supplier 

of dosimetry film discontinued production in November 1969 (Tucker 1970). Stockpiling of film 

enabled use of film dosimeters through approximately April 1971. 

Table 6-1. Beta/photon dosimeters.

Period Dosimeters used

Pre-1949 LANL dosimetry 

1949–1958

Film Badge 1: Metal holders were used that had a brass clip that covered one end of 

the film packets. The brass clip was intended to attenuate beta rays but not stop 

gamma rays (Kingsley 1953a). This holder was also issued as a wrist badge. 

1959–April 1971

Film Badge 2: Plastic holder had four windows: open window, 0.035-in. Al filter for 

beta/gamma, tungsten/cadmium and tin filters for thermal neutrons. Used DuPont 558 

film packets for beta/gamma exposures based on calibrations with Co-60 and 70-keV 

X-rays. Packets contained DuPont 519 film (stated range 30 mR to 10 R) and 1290 

film (stated range 10R to 3000 R) (Drake 1959; Tucker and Drake 1960).

May 1971–~1988

Harshaw Model 2271: The first TL badge had an open window and a 0.035-in. Al filter. 

The TLD card consisted of two TLD-100 elements, each 0.035-in. thick. (Tucker 1970; 

Thompson, Tucker, and Armijo 1972; Kingsley 1971; Author unknown 1988). 

~1989–~1994

Harshaw Model 8801 (7776-1141) cards in Model 8812 holder (Bradley 1993a,b,c; 

Friedman 1993; Rhea and Bradley 1991; Ward 1994)

1: 0.015-in. thick TLD-700 under 600 mg/cm2 ABS plastic (deep dose)

2: 0.015-in. thick TLD-700 under 242 mg/cm2 ABS plastic, 0.004-in. Cu (low-energy 

X-rays)

3: 0.0036-in. thick TLD-700 under open window, 0.0025-in. Mylar (shallow dose) 

4: 0.015-in. thick TLD-600 under 600 mg/cm2 ABS plastic (neutron dose)

~1995–present

Harshaw/Bicron EXTRAD put into use 2nd quarter 1997 for extremity dose.

Harshaw Model 8802 (7776-1161) or 8801 cards (see above) in Model 8812 holder

1: 0.015-in. thick TLD-700 under 600 mg/cm2 ABS plastic (deep dose)

2: 0.015-in. thick TLD-700 under 242 mg/cm2 ABS plastic, 0.004-in. Cu (low-energy 

X-rays)

3: 0.006-in. thick TLD-700 under open window, 0.0025-in. Mylar (thicker shallowdose chip than in the Model 8801)

4: 0.015-in. thick TLD-600 under 600 mg/cm2 ABS plastic (neutron dose)

(Walker 1996, 1997a; Bradley 1993d, 1994, 1995a,b)

Table 6-2. Neutron dosimeters.

Period Dosimeters used

1945–1958 No neutron dosimeters were used.

1959–April 1971 Film Badge 2: In a plastic holder with four windows; those with cadmium and tin filters 

were used to measure thermal neutrons. Used DuPont film for thermal neutron dose 

and NTA film for fast neutron dose based on calibrations with a Van de Graaff 

accelerator at energies of 1, 5, and 14 MeV. (Drake 1959; Tucker and Drake 1960).

May 1971–~1973 The first TL neutron badge consisted of three LiF elements: two LiF-600 that were 

sensitive to thermal neutrons and one LiF-700 that was insensitive to neutrons but was 

used to subtract any gamma contribution. The LiF elements were placed behind tin 

and cadmium filters to support determination of incident and scattered thermal 

neutrons (Tucker 1970; Kingsley 1971). 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 83 of 120

Period Dosimeters used

~1973–~1990 First dedicated neutron dosimeter incorporated two cards, each with a TLD-600 and a 

TLD-700 element. One TLD-600 and TLD-700 were enclosed in a borated 

polyethylene pouch and one TLD-600 and TLD-700 were uncovered. The holder had 

an open window and three 0.035-in. Al filters. Deep and shallow measurements were 

made with the TLD-700 elements and bare and boron-filtered measurements of 

neutrons with the TLD-600 elements to discriminate energies near thermal (Author 

unknown 1988).

July 1984 An albedo TLD was added for neutron dosimetry (Author unknown 1988).

~1990–1994 Harshaw ”Sandia Beta-Gamma-Neutron Configuration, Model 8812/8801”

1. 0.015-in. TLD-700 under 600 mg/cm2 ABS plastic (0.277-in.) 

2. 0.015-in. TLD-700 under 242 mg/cm2 ABS plastic and 0.004-in. copper 

3. 0.0015-in. TLD-700 under 0.0025-in. Mylar 

4. 0.015-in. TLD-600 under 600 mg/cm2 ABS plastic (neutron dose)

(Bradley 1993a,b,c; Friedman 1993; Rhea and Bradley 1991; Ward 1994)

1995–present Harshaw -“Sandia Beta-Gamma-Neutron Configuration, Model 8812/8802”

1. 0.015-in. TLD-700 under 600 mg/cm2 ABS plastic (0.277-in.) 

2. 0.015-in. TLD-700 under 242 mg/cm2 ABS plastic and 0.004-in. copper 

3. 0.006-in. TLD-700 under 0.0025-in. Mylar 

4. 0.015-in. TLD-600 under 600 mg/cm2 ABS plastic (neutron dose)

(Bradley 1993d, 1994, 1995a,b)

6.2 RECORDED DOSE PRACTICES 

Table 6-3 summarizes the dose-related quantities that have been reported in SNL annual dose 

summaries from 1949 to the 1990s (Widner 2008a,b,c). LANL TBD parameters (ORAUT 2009b)

apply to pre-1949 SNL dosimetry unless the monitoring records indicate that a different site managed 

the applicable dosimetry. 

The following relationships were used starting around 1971 to calculate compliance dose quantities 

(Ward 1994):

Skin dose = shallow dose + deep dose + neutron dose

Whole-body dose = deep dose + neutron dose

Extremities dose = shallow dose + deep dose + neutron dose

Electronic files of annual dose records received from SNL for years since 1986 include deep dose, 

neutron dose, eye dose, shallow dose, extremity dose, CEDE, and Total Effective Dose Equivalent

(Widner 2008a,b,c).

Table 6-3. Recorded dose practices.

Data on annual dose reports

1949

–

1957 1958

1959 

–

1965 1966

1967 

–

1969 1970

1971 

–

1976

1983 

–

1987

1988 

–

1991

1992

–

1994

“Total Body” ■ ■ ■ ■

“Total Wrist” ■

Total Dose (“Total”) ■

Total This Period (“TTP”) ■

Total This Quarter (“TTQ”) ■

Year to Date (“YTD”) ■ ■ ■ ■

Cumulative Total (“CT”) ■ ■ ■ ■

Gamma (“GAM” or “YGAM”) ■ ■ ■ ■ ■ ■

Fast Neutron (“FAST” or “YFAS”) ■ ■ ■ ■ ■

Thermal Neutron (“TH” or “THER” or ■ ■ ■ ■ ■
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 84 of 120

Data on annual dose reports

1949

–

1957 1958

1959 

–

1965 1966

1967 

–

1969 1970

1971 

–

1976

1983 

–

1987

1988 

–

1991

1992

–

1994

“YTH”)

“Beta” ■ ■

Wrist (“WR”) ■ ■

Plutonium (“PLU”) ■ ■ ■ ■

Tritium (“TRIT”) ■ ■ ■ ■

Uranium (“U” or “UR”) ■ ■ ■ ■

Other (“OTH”) ■ ■ ■ ■

Cumulative Beta (“CBETA”) ■

Cumulative Wrist (“CWR”) ■

Cumulative Nonpenetrating (“CNPR” or 

“CUM NPR”) ■ ■

Cumulative Extremity (“CEX”) ■ ■ ■

Extremity (“EXTR”) ■ ■ ■ ■ ■

Nonpenetrating (“NPR”) ■ ■ ■ ■

Annual Penetrating (“YPEN”) ■

Cumulative Penetrating (“CUM PEN”) ■

Annual Skin (“YSKIN”) ■ ■ ■

Cumulative Skin (“CUM SKIN”) ■

Cumulative Gamma (“CUM GAM”) ■

Cumulative Fast (“CUM FAS”) ■

Cumulative Thermal (“CUM THER”) ■

Annual Internal (“YINT”) ■ ■ ■ ■

Cumulative Internal (“CUM INT”) ■

“Eye” ■ ■

Neutron (“NEUT”) ■ ■ ■

“Deep” ■ ■

“Shallow” ■

6.3 UNMONITORED PHOTON AND ELECTRON DOSE

Detailed dose data from SNL facilities have been obtained for 1949 to 2005 in forms that support 

calculation of lognormal probability statistical parameters desired for characterization of annual dose 

distributions for use in assignment of unmonitored photon doses (Widner 2008a,b,c). Tables 6-4 and 

6-5 summarize the respective lognormal probability statistical parameters for SNL-NM dosimeter 

results for penetrating and nonpenetrating dose that are equal to or exceed 50 mrem for years of 

record between 1949 and 2005 (Widner 2008a,b,c). These data are useful to examine trends in the 

recorded doses. The reported doses that are the bases of these tables have not been corrected for 

potential missed doses, which is crucial to estimating potential dose for unmonitored workers.

Tables 6-4 and 6-5 demonstrate the bases for coworker data and can be applied if necessary for 

external dose reconstruction. However, compiled coworker dosimetry results with adjustments for 

missed dose provided in External Coworker Dosimetry Data for Sandia National Laboratories in 

Albuquerque, New Mexico (ORAUT 2008a) shall be used for applicable years when coworker doses 

are warranted. Coworker doses for years earlier than those addressed by ORAUT (2008a) can be 

assigned per Tables 6-4 and 6-5 and shall include applicable missed dose. If a worker appears likely 

to have had occupational external exposure to greater onsite ambient doses at SNL before 1949, the 

1949 data can be applied, as the exposure would have been to activities started in TA-II that carried 

into 1949. In addition, should it be necessary to evaluate unmonitored external shallow dose for years 

earlier than those addressed by ORAUT (2008a), it can be noted that 50th-percentile coworker

nonpenetrating-to-penetrating dose ratios were always less than 1.1 through the 1960s and 95th-
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 85 of 120

percentile ratios were always less than 1 for the same period. Therefore, the respective ratios of 1.1 

and 1 can be used to assign unmonitored shallow dose when necessary.

Table 6-4. SNL-NM worker penetrating dose statisticsa

, 1949 to 2005 (Widner 2008a,b,c).

Year

Recorded gamma dose data Lognormal fit

No. of workers reported gamma 

dose > 50 mrem

Gamma dose (mrem) Gamma dose (mrem)

Mean Maximum Median 95%-ile GSD

1949 10 110 420 Use 159 (b) (b)

1950 41 240 1,850 159 588 2.21

1951 84 200 3,000 134 452 2.09

1952 68 770 5,100 279 3,054 4.29

1953 92 180 3,030 111 399 2.18

1954 77 460 16,000 122 739 2.99

1955 65 260 2,520 131 690 2.74

1956 180 1,030 6,450 451 4,517 4.06

1957 428 610 4,570 238 2,134 3.79

1958 407 520 14,000 238 1,662 3.26

1959 358 250 5,000 148 636 2.43

1960 412 240 4,940 152 588 2.27

1961 598 190 2,370 122 441 2.18

1962 1,088 220 4,420 139 490 2.15

1963 548 180 2,430 117 419 2.17

1964 340 210 2,650 131 514 2.30

1965 248 240 2,080 130 642 2.64

1966 175 450 2,860 212 1,450 3.22

1967 316 280 3,710 133 685 2.72

1968 247 430 4,450 201 1,274 3.07

1969 235 390 3,850 183 1,074 2.93

1970 323 310 3,500 170 884 2.72

1971 485 240 4,090 130 592 2.51

1972 778 240 4,340 134 586 2.45

1973 839 200 4,510 120 476 2.31

1974 535 200 4,220 123 480 2.28

1975 722 180 8,140 114 401 2.15

1976 918 160 4,090 98 325 2.07

1977 757 176 4,260 104 388 2.23

1978 540 212 4,580 106 462 2.45

1979 306 236 3,990 120 553 2.53

1980 252 186 3,070 111 447 2.33

1981 137 274 3,210 151 727 2.59

1982 231 380 4,340 178 1,090 3.01

1983 231 250 4,360 123 570 2.55

1984 116 268 3,560 144 688 2.59

1985 150 263 2980 138 688 3.97

1986 105 292 2630 153 863 4.42

1987 143 320 3,580 143 869 2.99

1988 188 232 2,260 139 602 2.44

1989 111 198 1,194 130 507 2.29

1990 88 205 1,497 120 523 2.45

1991 90 170 1,256 99 395 2.31

1992 43 206 924 147 545 2.21

1993 37 180 521 134 479 2.17

1994 36 192 828 129 508 2.30

1995 38 151 495 119 348 1.92

1996 33 250 844 167 747 2.49
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 86 of 120

Year

Recorded gamma dose data Lognormal fit

No. of workers reported gamma 

dose > 50 mrem

Gamma dose (mrem) Gamma dose (mrem)

Mean Maximum Median 95%-ile GSD

1997 29 230 972 141 633 2.49

1998 36 194 945 115 512 2.48

1999 25 189 603 135 518 2.26

2000 30 189 720 141 498 2.16

2001 3 83 95 Use 141 (a) (a)

2002 1 52 52 Use 141 (a) (a)

2003 14 135 374 Use 141 (a) (a)

2004 64 164 572 137 377 1.85

2005 36 128 396 112 253 1.64

a. ORAUT 2008a coworker dose values shall be used for available years.

b. Too few data are available from 1949, 2001, 2002, and 2003 to define a distribution adequately. It is recommended that 

the median value given (highest surrounding year) be used as an estimate of unmonitored doses for these years.

Table 6-5. SNL-NM worker nonpenetrating dose statisticsa, 1987 to 2005 (Widner 2008a,b,c).

Year

SNL-NM recorded gamma dose data Lognormal fit

No. of workers 

reported shallow 

dose > 50 mrem

Nonpenetrating dose (mrem) Nonpenetrating dose (mrem)

Mean Maximum Median 95%-ile GSD

1977b 190 117 1,000 93 249 1.82

1978 b 116 168 1,060 111 427 2.26

1979 b 88 195 1,580 130 477 2.20

1980 b 74 121 740 95 248 1.79

1981 b 23 127 750 99 272 1.85

1982 b 43c 150 500 117 361 1.99

1983 b 38 186 870 127 491 2.28

1984 b 19 151 590 108 391 2.19

1985 36 272 2,080 137 767 4.39

1986 28 176 530 130 463 2.98

1987 138 370 4,050 152 1,017 3.18

1988 125 310 2,260 165 908 2.82

1989 112 221 1,392 137 575 2.39

1990 105 236 1,691 138 628 2.51

1991 130 203 1,639 112 507 2.51

1992 46 356 3,565 175 914 2.73

1993 54 216 702 145 622 2.42

1994 39 236 1,034 157 648 2.37

1995 42 180 750 135 437 2.04

1996 45 306 1,225 179 991 2.83

1997 48 256 1,906 138 678 2.63

1998 53 201 1,217 119 513 2.44

1999 33 218 858 146 629 2.43

2000 34 226 902 159 637 2.32

2001 3 105 126 (d) (d) (d)

2002 1 206 206 (d) (d) (d)

2003 17 121 374 105 252 1.70

2004 85 164 611 131 404 1.99

2005 55 129 396 109 276 1.76

a. ORAUT 2008a coworker dose values shall be used for available years.

b. For these years, values are based on worker dose records in which both penetrating and nonpenetrating dose were 

50 mrem or greater because those are the data that are available in a form that supports analysis at this time.

c. One extreme value of 49.93-rem nonpenetrating dose in 1982 was excluded from this analysis as accident-related, most 

certainly investigated in detail for the individual involved, and not typical of anticipated worker exposures.

d. Too few data are available from 2001 and 2002 to define a distribution adequately. It is recommended that the median 

value from 2000 be used as an estimate of unmonitored doses for these years.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 87 of 120

6.4 UNCERTAINTY FACTORS

For the usual analysis of measured film badge doses, it is possible to read a photon dose of 

100 mrem to within ±15 mrem if the exposure involved photons with energies between several keV 

and several MeV (Morgan 1961). The estimated standard error in recorded film badge doses from 

photons of any energy is ±30%.

The situation for neutrons was not as favorable as for photons. With NTA films used at SNL before 

1971, the estimated standard error was likely to be larger and varied significantly with the energy of 

the neutrons. Measured neutron doses to workers were very likely to be underestimated. The 

recommended approach is to estimate the neutron dose with neutron-to-photon dose ratios and the 

more reliably measured photon doses. 

Given the lack of specific technical information for dosimetry systems for much of the SNL history, it is 

necessary to estimate respective measurement uncertainty based on reported values for 

contemporary systems in use at other facilities. Given general analogies between the Hanford Site 

and SNL in terms of dosimetry technologies, the SNL uncertainty is based on the TBD for the Hanford 

Site (ORAUT 2010c).

6.5 MISSED BETA/PHOTON DOSE

Missed dose is the dose that may not have been accounted for in an individual’s records because of 

loss of or damage to the applicable dosimeter or because the records indicated zero dose due to the 

detection limitations of the film or TLD. 

For the usual analysis of measured film badge doses, minimum detection limits (MDLs) in the 

literature range from about 30 to 50 mrem for beta/photon irradiation (Ward 1994; Wilson et al. 1990). 

The film badge dosimetry at SNL was likely similar to dosimeters used at Hanford during the period 

1959 to 1968 (ORAUT 2010c). MDLs for the most recently used TLDs (Models 8801 and 8802) are 

estimated based on Walker (1997b), Stanley (1980), and Potter (1993b). Because of the lack of sitespecific data on MDLs for the early TLDs used from 1971 to 1990, these dosimeters are estimated to 

have had MDLs between those for the film badges and the more advanced TLDs. 

Several exchange frequencies were in use at any one time, so the dose reconstructor needs to 

determine the exchange frequency that applies to a specific worker from individual records. The 

values in the last two columns of Table 6-6 can be considered maximum annual missed doses for 

dose reconstructions. Beginning with 1979, badge exchange frequencies are assumed to be monthly 

if specific information to the contrary for the worker is not available (NIOSH 2007).

Table 6-6. SNL beta/photon dosimeter period of use, type, MDL, exchange frequency, and potential 

annual missed dose.

Period of 

use Dosimeter

Deep 

MDLa

(mrem)

Nonpenetrating 

MDLa

(mrad)

Exchange 

frequency

GM annual missed doseb

Deep dose

(mrem)

Nonpenetrating 

dose (mrad)

Pre-1949 If employment dates before November 1, 1949, are evaluated, dosimetry records should be 

requested from LANL if none are in the provided SNL files. LANL TBDc parameters apply to pre1949 dosimetry unless the monitoring records indicate a different site managed the dosimetry. 

1949–

~1958

Film in metal 

holder with open 

window and Pb 

filter

40 Not measured Biweekly 

(n=26)d

520 Not applicable

Monthly (n=12) 240

Quarterly (n=4) 80

1959–April 

1971

Film in plastic 

holder with open 

window, Al, Cd, 

30e 30e Biweekly 

(n=26) d

390 390

Monthly (n=12) 180 180
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 88 of 120

Period of 

use Dosimeter

Deep 

MDLa

(mrem)

Nonpenetrating 

MDLa

(mrad)

Exchange 

frequency

GM annual missed doseb

Deep dose

(mrem)

Nonpenetrating 

dose (mrad)

and Sn filters. Quarterly (n=4) 60 60

May 1971–

~1973

2271 TLD 20 38 Monthly (n=12) 120 228

Quarterly (n=4) 40 76

1973–

~1990

Two-card neutron 

TLD

20 38 Monthly (n=12) 120 228

Quarterly (n=4) 40 76

~1990–

~1994

8801 TLDf 10 35 Monthly (n=12) 60 210

10 35 Quarterly (n=4) 20 70

~1995–

present

8802 TLDf 5 5 Monthly (n=12) 30 30

8 10 Quarterly (n=4) 16 20

a. Estimated MDLs in the workplace for each dosimeter technology. 

b. Geometric mean (GM) annual missed dose calculated using MDL/2 from NIOSH (2007).

c. LANL occupational external dose TBD (ORAUT 2009b).

d. Dosimeters in reactor areas (Kingsley 1971) and for organizations handling radioactive materials 

consistently were commonly exchanged on a biweekly basis (i.e., every 2 weeks; Kingsley 1953b).

e. MDLs for photons and beta during this period are based on Hanford dosimeter values (ORAUT 2010c).

f. References for TLD MDLs: Walker (1997b), Stanley (1980), Potter (1993b). 

6.6 PARTITIONING OF BETA/PHOTON DOSES TO ENERGY CATEGORIES

Very little spectroscopy data to indicate gamma spectrum in SNL work areas have been found. To 

estimate the gamma spectrum to which workers were exposed, facilities were grouped into 

categories. 

Reactor facilities have dispersed fields of higher energy photons from fission as well as fission and 

activation products. Radioactive materials handling and processing facilities included a wide variety 

of activities. For electron accelerator facilities, bremsstrahlung photons dominate the secondary 

radiation field. Thick shields of concrete or other materials result in photons in the MeV energy range. 

For proton or positive ion accelerators, neutrons generally constitute the greatest hazard. Thick 

shields designed for the neutron hazard eliminate all but the most energetic photons. With shielding 

and safety interlock systems, exposure to the direct beam of SNL accelerators was rare and the 

personal exposure records should document them. However, maintenance personnel were exposed 

to activated accelerator components during repairs, target changes, etc. One reaction of particular 

importance is the thermal neutron capture of sodium in the concrete of the accelerator shielding 

(NCRP 2003). This reaction [23Na(n,γ)24Na] produces a radioactive isotope that decays with a 

15-hour half-life by emitting gamma rays of 1.4 and 2.8 MeV. Therefore, the gamma energy spectrum 

for accelerators was judged to consist primarily of photons with energies above 250 keV based on 

Argonne National Laboratory–East data (ORAUT 2006b). Radiation from X-ray machines and some 

radioisotopes presented low-energy photon hazards, but these appear to have been generally used in 

conjunction with higher energy sources rather than in distinct facilities. 

Table 6-7 lists the energy ranges for beta and photon exposures at SNL, which has possessed 

separated plutonium and sources of low-energy X-rays, so nonpenetrating doses could have resulted 

from exposures to low-energy photons as well as beta particles.

Plutonium facilities have mostly 17-keV photons with some 59-keV photons from 241Am. Plutonium 

handled at SNL had been separated from fission products. Low-energy photons that would have 

been received by workers handling plutonium components were not measured by early film badges 

that had no unfiltered areas. Before 1959, when nonpenetrating radiation was first measured (and 

reported in terms of beta dose), 100% of the measured photon (gamma) dose should be attributed to 

the 30-to-250-keV category. An additional dose of 1.86 times the measured dose should be attributed 

to the low-energy photon category (<30 keV). From 1959 through 1966, nonpenetrating dose 

(reported as beta dose) should be attributed to the <30-keV photon category for plutonium exposure. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 89 of 120

From 1967 to 1986, nonpenetrating dose (reported as such, “NPR”) should be attributed to the <30-

keV photon category. From 1987 to the present, shallow dose should be attributed to the <30-keV 

photon category. 

Table 6-7. Recommended beta and photon radiation energies and percentages for SNL-NM facilities.

Process/

buildings Description

Operations

Radiation 

type

Energy 

selection 

Begin End (keV) Percentage

Plutonium 

component 

operations

Weapon component inspection, 

testing, and assembly:

1949 Present Beta

photon

>15 100

<30 65a

30–250 35a

Reactors

(TA-V)

During Operation: Highly dispersed fields of higher 

energy photon radiation fields from fission process, 

activation and fission product nuclides. Potentially 

narrow beams of higher energy neutron radiation 

from test ports, etc., into reactor core. Potential for 

significant airborne nuclides; could be significant 

higher energy beta radiation.

Beta

photon

>15 100

30–250 25

>250 75

Not in Operation: Highly dispersed fields of higher 

energy photon radiation fields from activation and 

fission product nuclides. No significant neutron 

radiation. Possibly higher energy beta radiation 

during maintenance work resulting from fission 

products.

Various reactors

(see Table 2-2 for listing and 

date ranges)

1958 Present

Uranium 

component 

operations

Weapon component inspection, 

testing, and assembly: depleted 

and enriched uranium.

1949 Present Beta

photon

>15 100

30–250 100

Thorium 

component 

operations

Weapon component inspection, 

testing, and assembly:

1959? Present Beta

photon

>15 100

30–250 50

>250 50

Accelerator 

operations 

(in TA-IV)

Simulation of weapon environments, material and 

component testing, fusion and particle beam 

research:

Beta

photon

>15 100

30–250 10

>250 90

Various accelerators

(see Table 2-2 for listing and 

date ranges)

1958 Present

Calibrations 

and 

irradiations

Onsite irradiation of instruments, dosimeters, 

components:

Beta

photon

>15 100

30–250 5

>250 95

HCF Isotope production for medical 

use.

1998 2003? Beta

photon

>15 100

30–250 40

>250 60

Waste 

handling

Radiation characteristics highly 

dependent on source of waste.

1949 Present Beta

photon

>15 100

30–250 50

>250 50

a. Low-energy photons were not measured by early SNL-NM film badges that had no unfiltered areas. For years before 

1959, 100% of the measured photon (gamma) dose should be attributed to the 30-to-250-keV category. An additional 

dose of 1.876 times the measured dose should be attributed to the low-energy photon category (<30-keV) based on 

Traub, Scherpelz, and Taulbee (2005).
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 90 of 120

6.7 UNMONITORED NEUTRON DOSE

There should not typically be significant neutron exposure of unmonitored workers. However, if 

estimation of neutron dose is called for, apply the neutron-to-photon distribution data from Table 6-8

to measured and estimated missed or unmonitored photon doses. This process of calculating a 

neutron dose is based on the assumption that the distribution of neutron doses to unmonitored 

workers is equivalent to the distribution of neutron doses measured for monitored workers. 

Table 6-8. Recommended distributions for neutron-to-photon ratio.

Neutron source type

Neutron-to-photon dose ratio

GM GSD 95%-tile

General operations 1987 to present (based on 

16 annual records between 1987 and 2005a)

0.36 2.55 1.70

General operations in earlier years (based on 

38 annual records from 1977 through 1984a)

0.39 4.97 5.43

a. Source: Widner (2008a,b,c).

The distributions of neutron-to-photon ratios should not be used to estimate neutron doses if neutron 

measurements are available after 1971, when thermoluminescent dosimeters replaced NTA film at 

SNL for neutron dosimetry. 

The application of the data from Table 6-8 to measured or estimated photon doses can be 

accomplished through Monte Carlo simulation. As an alternative, the geometric mean value of the 

ratio can be applied in a lognormal distribution with a geometric standard deviation as listed in 

Table 6-8 or, to obtain an overestimate of unmonitored neutron dose, the appropriate 95th-percentile 

value from Table 6-8 can be applied to the measured or estimated photon dose for each period of 

interest. The “general operations” values from Table 6-8 are based on the reported annual doses for 

1987 through 2005 and for 1977 through 1984 for individuals who had both gamma and neutron dose 

results of 100 mrem or greater (Widner 2008a,b,c). This assessment was based on 16 cases for 

1987 through 2005 and 38 cases for 1977 through 1984.

6.8 MISSED NEUTRON DOSE

Neutron radiation has been present at SNL in association with various weapons components, 

reactors, and accelerators. For other locations, missed neutron dose is very unlikely because of the 

very low potential for neutron exposure. To calculate the missed neutron dose, the dose 

reconstructor must first determine if the person worked in an area with potential neutron radiation and 

the category of neutrons. This can best be determined by examining the work location records and if 

a worker met any of the following conditions:

• Positive neutron dosimetry results were recorded.

• A specific neutron dosimeter was issued to the worker. This would be indicated for SNL-NM

dosimeters by a recorded neutron zero or by the use of multiple dosimeters, aside from 

extremity dosimeters, for a given monitoring period. For certain periods, such as 1990 and 

later, recorded neutron zeros will often be simply a recording practice due to a neutron 

element in the issued TLD. In such a case, the conditions of exposure and likelihood of 

significant neutron radiation must be considered in determining if missed neutron dose is 

warranted (ORAUT 2008b).

• Summary dose records contain recorded neutron dose results, zero or otherwise. As with the 

example above, summary records for 1990 and later may contain recorded neutron zeros 

simply as a recording practice due to a neutron element in the issued TLD(s). In such a case, 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 91 of 120

the conditions of exposure and likelihood of significant neutron radiation must be considered in 

determining if missed neutron dose is warranted (ORAUT 2008b).

If any of the above conditions are met, at least some potential for exposure to neutron radiation was 

likely. If none of the above conditions are met, the dose reconstructor should assume that the person 

was not exposed to significant neutron radiation.

For periods before 1971, distribution of neutron-to-photon ratios from Table 6-8 should be applied to 

the missed photon doses to estimate missed neutron doses. Table 6-9 lists the neutron missed doses 

for those monitored for neutron exposure during 1971 and later.

Table 6-9. Neutron dosimeter period of use, type, MDL, exchange frequency, and potential annual 

missed dose.

Period of use Dosimeter

MDL 

(mrem)

Exchange 

frequency

Mean annual 

missed dose (mrem)a

1959–April 1971 Film badge 2 with NTA film <50 Monthly (n = 12) <300b

May 1971–1989 First TL dosimeter and Model 

2271 TLD badgec

30 Monthly (n = 12) 180

Quarterly (n = 4) 60

1990–present Model 8801/8802 TLD badgesc 10 Monthly (n = 12) 60

Quarterly (n = 4) 20

a. Mean annual missed neutron dose calculated using MDL/2 from NIOSH (2007). 

b. Neutron-to-photon ratio should be used to estimate missed doses during these periods.

c. The source for 8801/8802 badge MDLs is Walker (1997b). The lower limit of detection for neutron doses on the 8801 

dosimeter is a function of neutron dose and ranges from 0.010 to 0.120 rem for low-energy to 14-MeV neutrons. 

Because of the lack of site-specific data on MDLs for the early TLDs used from 1971 to 1989, these dosimeters are 

estimated to have had MDLs between those for NTA film and the more advanced TLDs.

6.9 PARTITIONING OF NEUTRON DOSES TO ENERGY CATEGORIES

Table 6-10 lists default neutron dose fractions by energy range for SNL operational areas where 

neutron exposures were possible along with the associated ICRP Publication 60 correction factors 

(ICRP 1991). The neutron dose equivalent is calculated by multiplying the recorded neutron dose by 

the area-specific correction factors.

Because of the variety of reactors used at SNL, and the paucity of neutron spectrum data for them, an 

assumption of 100% fission spectrum neutrons (0.1-to-1 MeV) is used. The energy distribution for 

plutonium component handling is based on data from the LANL occupational external dose TBD 

(ORAUT 2009b). Energy characteristics for neutron generators are based on Ward (1995). Data for 

PuBe sources are based on Buckner and Sims (1992). Neutrons from Cockroft-Walton generators 

vary based on target and energy applied, but are most commonly 2.5 MeV and 14.1 MeV (Elliott 

1971). The energy ranges for other accelerators are based on a combination of the data for the Zero 

Gradient Synchrotron and Intense Pulsed Neutron Source from the external dosimetry TBD for 

Argonne National Laboratory–East (ORAUT 2006b).

The following sources of photon/beta radiation identified in Table 6-7 are not listed in Table 6-10

because they are not significant sources of neutron exposure at SNL-NM: uranium component 

operations, thorium component operations, Advanced Nosetip Test (ANT)/Nosetip payload 

operations, HCF operations, and routine waste handling. Neutron generator testing is not a significant 

source of photon/beta exposure, so it is listed only in Table 6-10. Cockroft-Walton generator 

operations, a component of Accelerator Operations included in Table 6-7, are a potentially significant 

source of neutron exposures. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 92 of 120

Table 6-10. Recommended dose fractions and ICRP Publication 60 correction factors for SNL 

neutron sources.

Process Facility description

Operations Neutron 

energy

Default dose 

fraction (%)

ICRP 60 

correction 

Begin End factor

Plutonium

component 

operations

Weapon component inspection, testing, and assembly:

Plutonium handled at SNL had 

been separated from fission 

products.

1949 Present <10-100 keV

0.1-2 MeV

2-20 MeV

11

56

33

0.23

1.1

0.44

Reactor 

operations

Neutrons of varying energies from reactor operations:

Various reactors (see 

Table 2-2 for a listing and date 

ranges)

1958 Present <10-100 keV

0.1-2 MeV

2-20 MeV

0

100

0

─

1.9

─

Neutron 

generators

Testing of neutron sources for 

weapons applications (typically 

14 MeV).

1959 1997+ <10-100 keV

0.1-2 MeV

2-20 MeV

0

0

100

─

─

1.3

Calibration 

and 

irradiation

Onsite irradiation of instruments, dosimeters, components:

Neutron source range 1950? Present <10-100 keV

0.1-2 MeV

2-20 MeV

0

20

80

─

0.38

1.1

PuBe neutron sources 1950? Present

CockroftWalton 

Generator

Simulation of weapon environments, material and component testing:

Sandia Cockroft-Walton

accelerator facility

Late 

1950s

Present <10-100 keV

0.1-2 MeV

2-20 MeV

5

5

90

0.11

0.095

1.2

Other 

accelerators

Various accelerators (see 

Table 2-2 for a listing and date 

ranges)

1958 Present <10-100 keV

0.1-2 MeV

2-20 MeV

>20 MeV

5

14

46

35

0.11

0.27

0.61

0.35

6.10 RECOMMENDED DOSE CONVERSION FACTORS

Dose Conversion factors (DCFs) should be selected from NIOSH (2007) for the dose quantities 

specified in Table 6-11 for the periods of interest.

Table 6-11. Recommended DCFs for SNL dose assessments.

Period Recommended photon DCFs Recommended neutron DCFs

1949–1971 Exposure (R) to organ dose equivalent (HT) Deep dose equivalent (Hp,slab(10)) 

1972–present to organ dose equivalent (HT) Deep dose equivalent (Hp(10)) to organ dose 

equivalent (HT)

7.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in the preceding text, bracketed callouts indicate information, conclusions, and 

recommendations to assist in the process of worker dose reconstruction. These callouts are listed in 

this section with information that identifies the source and justification for each item. Conventional 

references are provided in the next section that link data, quotations, and other information to 

documents available for review on the Oak Ridge Associated Universities Team servers.

[1] Buddenbaum, Jack E., Certified Health Physicist (CHP). ENSR. Health Physicist. 2007.

Statements about name changes to the site are based on discussions in February 2006 that 

were confirmed with Tracy Ikenberry on February 9, 2007.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 93 of 120

[2] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

A summary of the history of SNL from information in previous sections of the site profile is 

included here for completeness.

[3] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is taken from documented discussions from 2005 through February 2007 with 

an SNL retiree about his activities at nuclear storage sites and SNL general activities at 

storage sites and from information in previous sections of the site profile.

[4] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This information is included to alert the dose reconstructor that additional dosimetry records 

might need to be requested if there is indication that the worker could have visited other SNL 

sites during the employment period and there is no indication of dosimetry records in the file or 

requests to those facilities. This statement is included here for completeness.

[5] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

A summary of the activities of SNL from information in previous sections of the site profile is 

included here for completeness.

[6] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This summary statement is taken from discussions during 2005 through February 2007 with 

SNL retirees and current staff members about the types of activities at SNL and the perceived 

intake potentials from those activities. 

[7] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

The discussion in this section is included to assist the dose reconstructor to determine the 

potential for exposure for a worker based on activities that might have been described as 

“assembly” of weapons. The information is from the cited document.

[8] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

The nuclides of primary concern are based on those nuclides listed in documents and reports 

and discussions with SNL retirees and current staff.

[9] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

Tables 5-1 and 5-2 are provided based on the cited reference and information in other 

sections of the document. These locations are summarized here for completeness and to 

provide the dose reconstructor a source of information on the intake potential of various areas 

on the site.

[10] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

The table of nuclear weapons tests is included to assist the dose reconstructor in recognizing 

the names of these tests in the event these names are mentioned in the telephone interview or 

other records. While these tests are included in other documents, they are included here for 

completeness. 

[11] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of worker dosimetry records and conversations with SNL 

retirees and current staff members between 2005 and February 2007.

[12] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of documents that are cited throughout Section 5.0 and 

conversations with SNL retirees and current staff members between 2005 and February 2007.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 94 of 120

[13] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

Conversations with SNL retirees between 2005 and February 2007 indicated that bioassay 

was being conducted and analyzed on site as early as 1949. The program was not formalized 

because there was a perception that there was little need for a routine bioassay program. 

Fewer than 100 workers participated annually in bioassay in the pre-1992 era. This is contrary 

to the perception of current SNL personnel. Conversations with current personnel indicated 

that bioassay was extremely rare and, if it was conducted, the samples were sent off site. 

[14] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This paragraph is based on review of documents that are cited throughout Section 5.0 and 

conversations with SNL retirees and current SNL personnel between 2005 and February 2007. 

The retirees participated in the bioassay programs. One retiree was involved in the analysis of 

the samples, and another was involved in the archiving of the data. Research is currently 

being done to retrieve the bioassay results that were archived during that period. Because of 

statements made during interviews with individuals currently involved with the bioassay 

program, outsourced bioassay and an MOU with LANL were considered to be the prime 

source of bioassay results for years before 1991. However, information is contradictory to the 

outsourcing of in vitro samples. Because the methods of that time were similar to those used 

at LANL for uranium and tritium, MDAs listed for LANL are most likely to be applicable to the 

analyses at SNL if no other information is available. 

[15] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

The reference cited lists a number of sites that reported positive results in control subjects 

counted with the Helgeson system during that period. Because the majority of counted 

workers in 1989 and 1990 at SNL had positive results, Sam Glover, NIOSH Health Physicist, 

suggested inclusion of this reference. However, it is favorable to the claimant to include the 

positive results in the dose reconstruction. 

[16] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with an SNL retiree in February 2007. The retiree

participated in the archiving of the dosimetry records beginning in 1967.

[17] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of records supplied by LLNL.

[18] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of records supplied by LLNL.

[19] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with active SNL personnel between 2005 and 

February 2007.

[20] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is included in the discussion to show that SNL reviewed the requirements for 

bioassay and made decisions accordingly.

[21] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

The bioassay results that have been reviewed have been accompanied by background and 

standard calibration results. According to conversations with retirees, the procedures used to 

analyze these samples were standard protocol similar to the protocol used by LANL. 

Therefore, the MDAs reported by LANL are generally applicable. 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 95 of 120

[22] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

Although the actual procedures used by the Industrial Hygiene Department have not been 

found, conversations in February 2007 with retirees indicated that the analytical protocol for 

analysis of bioassay was basically the same as that used at LANL during that time. Therefore, 

the assumption of MDAs in the same range can be made if no other information is available.

[23] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

Review of the radionuclides associated with the locations on site and at offsite locations and 

conversations with SNL retirees and current staff members indicated that plutonium was not a 

consistent source term across the site. 

[24] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of documents that are cited throughout Section 5.0 and 

conversations with SNL retirees and current staff members between 2005 and February 2007.

[25] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations in February 2007 with an SNL retiree .

[26] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations in February 2007 with an SNL retiree .

[27] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of documents that are cited throughout Section 5.0 and 

conversations with SNL retirees and current staff between 2005 and February 2007.

[28] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of documents that are cited throughout Section 5.0 and 

conversations with an SNL retiree and current SNL personnel between 2005 and February 

2007.

[29] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of documents that are cited throughout Section 5.0 and 

conversations with SNL retirees and current staff between 2005 and February 2007.

[30] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

Until alpha spectroscopy was available to analyze the electroplated samples, only gross alpha 

was counted on either a proportional counter or NTA film. However, the label on the results 

might have been 239Pu. Analysis involving alpha spectroscopy would include results for both 

239Pu and 238Pu. It is not possible to distinguish between 239Pu and 240Pu using alpha 

spectroscopy techniques; therefore, the activity of 240Pu is included in the activity listed for 

239Pu. 

[31] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

The term pure 238Pu was typically used in reference to a mixture of 238Pu that is actually 

predominantly 238Pu, as shown by the ratios of the mixtures listed in the cited references. The 

pure 238Pu mixture is found in heat source technology work. 

[32] Bihl, Donald. Hanford. Health Physicist. 2006.

Calculations were performed to determine ratios using the Pu.ex software (Rittmann 1984). 

[33] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of documents that are cited throughout Section 5.0 and 

conversations with SNL retirees and current staff between 2005 and February 2007.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 96 of 120

[34] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on a conversation with an Internal Dosimetry Department employee in 

2006. 

[35] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of documents that are cited throughout Section 5.0 and 

conversations with SNL retirees and current staff between 2005 and February 2007. 

[36] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

Records of tritium surveys with positive results are found throughout the archives for various 

locations on the site, including the locations given as examples in this section.

[37] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

Records of tritium surveys with positive results are found throughout the archives for various 

locations on the site, including the locations given as examples in this section.

[38] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This protocol was established for LLNL tritium doses as an overestimate.

[39] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on review of documents that are cited throughout Section 5.0 and 

conversations with SNL retirees and current staff between 2005 and February 2007. While the 

general internal dosimetry program was based on the concept that the majority, if not all, of the 

airborne radioactivity comprised nonrespirable particles, the examples of operations in this 

paragraph indicated that there were situations that could produce respirable particles of 

uranium and other radionuclides. 

[40] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is from the Internal Dosimetry Program procedures. Special bioassay 

sampling would only be initiated by a routine sample that exceeded the limits when a routine 

sampling program was actually in place.

[41] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with SNL staff in 2006. 

[42] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with SNL staff in 2006.

[43] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with SNL staff in 2006 and SNL retirees in 2007. 

[44] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with an SNL retiree in 2006 and 2007. The retiree

stated that the protocol and analytical technique in the early days were based in general on 

standard techniques in use in the industry at the time. Protocol similar to that used at LANL 

was used at SNL, therefore the analytical sensitivities can be considered to be similar.

[45] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This decision to classify an intake as occupational based on whether the concentration was 

within the expected range from dietary uranium intakes is known to be the practice since 1993. 

Whether any similar decisions to classify positive results were made before that time is not 

known.
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 97 of 120

[46] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

Conversations with SNL retirees indicated that uranium bioassay was performed and recorded 

at SNL as early as 1949.

[47] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is included as an example of a contamination incident and the SNL response to 

such incidents, in which engineering controls were used prevent a reoccurrence. This type of 

response was frequently indicated in documents that were reviewed.

[48] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on the review of numerous log sheets that contain air sample results. 

These log sheets span many years of the history of SNL. 

[49] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with SNL staff in 2006.

[50] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with SNL staff in 2006.

[51] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with an SNL retiree in 2006.

[52] Argall, Rowena S., CHP. ENSR. Health Physicist. 2007.

This statement is based on conversations with an SNL retiree in February 2007. The retiree

participated in the archiving of the dosimetry records beginning in 1967.

 
Document No. ORAUT-TKBS-0037 Revision No. 01 Effective Date: 01/25/2013 Page 98 of 120

REFERENCES

ANSI (American National Standards Institute), 1972, American National Standard: Practice for 

Occupational Radiation Exposure Records Systems, ANSI N13.6-1966 (R1972), New York, 

New York. [SRDB Ref ID: 21974]

ANSI (American National Standards Institute), 1980, American National Standard: Practices for 

Respiratory Protection, ANSI Z88.2-1980, New York, New York.

Antonsen, R. G. Jr., 1997, Radiation Protection Survey of the Sandia National Laboratory, U.S. 

Department of Health and Human Services, Food and Drug Administration, March 

(unpublished). [SRDB Ref ID: 27228, pp. 2, 4-6]

Author unknown, 1988, Operational Procedures of the Sandia Personnel Radiation Dosimetry 

Program 1979, Sandia Corporation, Sandia National Laboratories, Albuquerque, New Mexico, 

July 18. [SRDB Ref ID: 23883]

Author unknown, date unknown, “Layout of Glovebox Laboratory,” Figure 2.1-1 from unknown 

document. [SRDB Ref ID: 22281, p. 6]

Ball, S., 1991, “Possible Exposure to Tritium,” memorandum to TTR Security Personnel, Sandia 

Corporation, Sandia National Laboratories, Albuquerque, New Mexico, April 30. [SRDB Ref 

ID: 19273]

Bradley, E. W., 1993a, Trip Control Data Review, Sandia Corporation, Sandia National Laboratories, 

Albuquerque, New Mexico, January 12. [SRDB Ref ID: 23578, p. 204]

Bradley, E. W., 1993b, Trip Control Dosimeters Test Criteria Determination, Sandia Corporation, 

Sandia National Laboratories, Albuquerque, New Mexico, January 13. [SRDB Ref ID: 23578]

Bradley, E. W., 1993c, Harshaw 8800 TLD Card Readers System Background Data Review, Sandia 

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GLOSSARY

aging

In relation to reactor fuel and mixtures of plutonium isotopes, time since the step in the 

refinement process that separates americium from the mixture.

ALIAS

A particle accelerator facility.

cohort 

Group of individuals selected for inclusion in a study.

depleted uranium (DU) 

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. 

fabrication

Manufacturing.

glovebox 

Enclosure with special rubber gloves through which an operator can handle radioactive or 

toxic material without risk of injury or contamination normally operated at a slightly reduced 

pressure so that air leakage, if any, is inward.

Hermes II

High-energy, pulsed, field-emission, electron-beam or bremsstrahlung gamma ray generator.

hot cell 

Shielded laboratory for handling of radioactive materials with the aid of remotely operated 

manipulators. The walls and windows are made of materials that protect workers from 

radiation. 

Hydramite II

Dual-line, high-energy, short-pulse source for electron-beam or bremsstrahlung X-ray

Kiva

One of the remotely controlled critical assembly buildings associated with the Critical 

Experiment Facility at LANL. From Hopi, ki- means house, but the meaning of -va is unknown. 

particle accelerator 

Device that accelerates ions using magnetic and/or electrostatic fields for focusing and 

redirecting ion beams. The main purposes of accelerators are the investigation of high-energy 

particle behavior and production of synthetic isotopes.

reactor 

Device in which a fission chain reaction occurs under controlled conditions to produce heat or 

useful radiation for experimental purposes or to generate electrical power or nuclear fuel. 

simulated 24-hour urine sample

Collection of all urine samples beginning with the void before retiring for the evening and 

ending with the first void after rising the next morning, for two consecutive nights, to simulate a 

24-hour urine sample.
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SPEED

High-energy, very-short-pulse bremsstrahlung X-ray accelerator.

TROLL

A particle accelerator facility.