Page 1 of 113

DOE Review Release 08/28/2020

Lawrence Livermore National Laboratory – 

Occupational Internal Dose

ORAUT-TKBS-0035-5 Rev. 04

Effective Date: 09/01/2020

Supersedes: Revision 03 

Subject Expert(s): Matthew G. Arno, Mutty M. Sharfi, and Stephen Spanos

Document Owner 

Approval: Signature on File

Stephen Spanos, Document Owner

Approval Date: 04/30/2020

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 05/04/2020

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 05/01/2020

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 05/01/2020

Approval: Signature on File

Timothy D. Taulbee, Associate Director for Science

Approval Date: 05/07/2020

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 2 of 113

 

PUBLICATION RECORD

EFFECTIVE

DATE

REVISION

NUMBER DESCRIPTION

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

Laboratory – Occupational Internal Dose. First approved issue.

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

Jay J. Maisler.

03/16/2010 01 Revision to incorporate Attributions and Annotations section and

Advisory Board comments associated with the issuance of a LLNL

SEC Class. Incorporates the internal coworker study (ORAUT-OTIB0065) into the TBD as Attachment B. Incorporates formal internal and 

NIOSH review comments. Constitutes a total rewrite of the

document. Training required: As determined by the Objective

Manager. Initiated by Mutty M. Sharfi, Stephen Spanos and Ralph W. 

Kenning.

12/13/2010 02 Revision to update Table B-9 uranium 95th percentile coworker 

intakes for those time periods that were based on the 50th percentile 

intakes with a GSD that were rounded up to 3. The previous revision 

used the actual GSDs in Table B-8 that were less than 3 to determine 

the 95th percentile coworker intakes for those time periods. Added 

information pertaining to solubility types for aged related fission 

products for P-32 and C-14 in Section 5.3.2.3 and added associated 

footnotes to Tables B-25 and B-26. Incorporates formal internal and 

NIOSH review comments. Training required: As determined by the 

Objective Manager. Initiated by Stephen Spanos and Ralph Kenning.

08/12/2016 03 Revision initiated to incorporate SEC-000221. Revised Section 

5.2.3.1 to incorporate additional guidance for which tritium exposures 

in the form of metal tritides at LLNL should be considered. Revised 

Section 5.3.1 to incorporate additional information on gross alpha 

measurements, and more detailed guidance on the analysis of gross 

alpha, and plutonium bioassay results. Revised Section 5.3.2.1 to 

incorporate revisions to OTIB-0054. Added Section 5.3.2.4 to 

incorporate guidance on the analysis of tritium measurements. 

Revised Section B.4.2.3 to incorporate revisions to OTIB-0054.

Revised Attachment C to incorporate additional guidance on the 

assignment of coworker intakes. Incorporates formal internal and 

NIOSH review comments. Training required: As determined by the 

Objective Manager. Initiated by Stephen Spanos and Ralph W. 

Kenning.

09/01/2020 04 Revision initiated to include the potential and intake rates for type SS 

plutonium. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by 

Matthew G. Arno.
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TABLE OF CONTENTS

SECTION TITLE PAGE

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

5.1 Introduction ............................................................................................................................ 10

5.1.1 Purpose...................................................................................................................... 11

5.1.2 Scope......................................................................................................................... 11

5.1.3 Special Exposure Cohort............................................................................................ 12

5.2 Source Term .......................................................................................................................... 12

5.2.1 Plutonium ................................................................................................................... 13

5.2.2 Uranium...................................................................................................................... 14

5.2.3 Tritium ........................................................................................................................ 15

5.2.3.1 Metal Tritides ............................................................................................... 16

5.2.3.2 Organically Bound Tritium............................................................................ 17

5.2.4 Other Limited-Exposure Radionuclides....................................................................... 17

5.3 Measurement Methods .......................................................................................................... 17

5.3.1 Analysis of Gross Alpha Measurements ..................................................................... 20

5.3.2 Analysis of Gross Beta Measurements ....................................................................... 22

5.3.2.1 Reactor Mixtures, 1974 to 1980 ................................................................... 23

5.3.2.2 Weapons Residue, 1974 to 1992 ................................................................. 23

5.3.2.3 Aged Fission Products and Research Radionuclides, 1992 to 2001............. 25

5.3.2.4 Analysis of Tritium Measurements ............................................................... 26

5.3.3 Measurement Types and Detection Levels................................................................. 26

5.3.4 Reporting Formats and Codes.................................................................................... 30

5.3.5 Assessment of Intake for Unmonitored Workers......................................................... 31

5.3.6 Unmonitored Intakes of Radionuclides Other than Primary Radionuclides ................. 31

5.4 Significant Incidents with Internal Dose Potential ................................................................... 32

5.5 Attributions and Annotations .................................................................................................. 33

References ........................................................................................................................................ 35

Glossary ............................................................................................................................................ 42

ATTACHMENT A ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001........................ 46

ATTACHMENT B INTERNAL DOSIMETRY CO-EXPOSURE DATA........................................... 61

ATTACHMENT C SCALING OF CO-EXPOSURE DOSE INTAKES FOR TYPE AND 

DURATION OF EXPOSURES ...................................................................... 112
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LIST OF TABLES

TABLE TITLE PAGE

5-1 Specific activity for specified ages of weapons-grade plutonium ............................................ 14

5-2 Activity ratios in mixture for specified ages of weapons-grade plutonium ............................... 14

5-3 Isotopic information for various uranium enrichment levels .................................................... 15

5-4 Y-12 Plant recommended RU contaminant ratios default levels based on upper levels of 

expected ranges ....................................................................................................................15

5-5 Feed Materials Production Center RU contaminant ratios for NU........................................... 15

5-6 Routine in vitro sampling frequencies..................................................................................... 19

5-7 In vivo sampling frequencies .................................................................................................. 19

5-8 Bioassay in vitro detection levels............................................................................................ 26

5-9 Bioassay in vivo detection levels ............................................................................................ 27

5-10 Bioassay database record entries .......................................................................................... 30

5-11 Dose assignment for radionuclides other than primary radionuclides ..................................... 32

5-12 Input parameters for significant incidents ............................................................................... 33

A-1 Environmental releases by building, 1993 to 2001 ................................................................. 46

B-1 Americium co-exposure sample statistic data......................................................................... 67

B-2 Beta co-exposure sample statistic data.................................................................................. 67

B-3 Plutonium co-exposure sample statistic data.......................................................................... 68

B-4 Uranium co-exposure sample statistic data............................................................................ 69

B-5 Time intervals for statistical analysis of uranium urine samples.............................................. 71

B-6 Time intervals for statistical analysis of gross beta urine samples .......................................... 75

B-7 Uranium urine bioassay data for 50th- and 84th-percentiles with the period and effective 

bioassay date, 1958 to 1996 .................................................................................................. 75

B-8 Uranium mass and activity intake rates for three different periods.......................................... 79

B-9 50th- and 95th-percentile uranium mass and activity intake rates for three different 

periods...................................................................................................................................79

B-10 50th- and 84th-percentiles for 239Pu urine bioassay data, 1957 to 1996 ................................. 80

B-11 50th- and 84th-percentile intake rates type M 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 81

B-12 50th- and 84th-percentile intake rates of type S 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 81

B-13 50th- and 84th-percentile intake rates of type SS 239Pu from co-exposure plutonium 

urine bioassay data, 1957 to 1996 ......................................................................................... 81

B-14 50th- and 95th-percentile intake rates of type M 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 82

B-15 50th- and 95th-percentile intake rates of type S 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 82

B-16 50th- and 95th-percentile intake rates of type SS 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 82

B-17 50th- and 84th-percentile excretion rates for 241Am urine bioassay data, 1957 to 1996 .......... 82

B-18 50th- and 95th-percentile intake rates of type M 241Am from co-exposure americium urine 

bioassay data, 1957 to 1996 .................................................................................................. 83

B-19 50th- and 84th-percentile excretion rates for gross beta urine bioassay data, 1974 to 

1995.......................................................................................................................................84

B-20 50th- and 84th-percentile intake rates of type F 90Sr, 1974 to 1980........................................ 84

B-21 50th- and 95th-percentile intake rates of type F 90Sr, 1974 to 1980........................................ 84

B-22 50th- and 84th-percentile intake rates of type F 103Ru, 1974 to 1992...................................... 85

B-23 50th- and 84th-percentile intake rates of type S 141Ce and type F 89Sr, 1974 to 1992............. 86

B-24 50th- and 95th-percentile intake rates of type F 103Ru, 1974 to 1992...................................... 86

B-25 50th- and 95th-percentile intake rates of type S 141Ce and type F 89Sr, 1974 to 1992............. 86
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B-26 50th- and 84th-percentile intake rates of type F 90Sr, 90Y, 32P, 137Cs, and 14C, after 1992 ....... 87

B-27 50th- and 95th-percentile intake rates of type F 90Sr, 90Y, 32P, 137Cs, and 14C, after 1992 ....... 87

C-1 Levels of contact for certain job categories .......................................................................... 113

LIST OF FIGURES

FIGURE TITLE PAGE

B-1 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1958 to December 1973 ........................................................... 87

B-2 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1974 to December 1987 ........................................................... 87

B-3 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1988 to December 1996 ........................................................... 88

B-4 Predicted and observed 50th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type F NU .................................................................................. 88

B-5 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1958 to December 1973 ........................................................... 88

B-6 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1974 to December 1987 ........................................................... 89

B-7 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1988 to December 1996 ........................................................... 89

B-8 Predicted and observed 84th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type F NU .................................................................................. 89

B-9 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1958 to December 1973 .......................................................... 90

B-10 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1974 to December 1987 .......................................................... 90

B-11 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1988 to December 1996 .......................................................... 90

B-12 Predicted and observed 50th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type M NU ................................................................................. 91

B-13 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1958 to December 1973 .......................................................... 91

B-14 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1974 to December 1987 .......................................................... 91

B-15 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1988 to December 1996 .......................................................... 92

B-16 Predicted and observed 84th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type M NU ................................................................................. 92

B-17 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1959 to December 1973........................................................... 92

B-18 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1974 to December 1987........................................................... 93

B-19 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1988 to December 1996........................................................... 93

B-20 Predicted and observed 50th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type S NU.................................................................................. 93

B-21 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1958 to December 1973........................................................... 94

B-22 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1974 to December 1987........................................................... 94
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B-23 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1988 to December 1996........................................................... 94

B-24 Predicted and observed 84th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type S NU.................................................................................. 95

B-25 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1957 to December 1959 ................................................ 95

B-26 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1960 to December 1964 ................................................ 95

B-27 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1965 to December 1966 ................................................ 96

B-28 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1967 to December 1981 ................................................ 96

B-29 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1982 to December 1990 ................................................ 96

B-30 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1991 to December 1996 ................................................ 97

B-31 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1957 to December 1959 ................................................ 97

B-32 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1960 to December 1963 ................................................ 97

B-33 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1964 to December 1966 ................................................ 98

B-34 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1967 to December 1981 ................................................ 98

B-35 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1982 to December 1990 ................................................ 98

B-36 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1991 to December 1996 ................................................ 99

B-37 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1957 to December 1959................................................. 99

B-38 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1960 to December 1964................................................. 99

B-39 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1965 to December 1966............................................... 100

B-40 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1967 to December 1981............................................... 100

B-41 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1982 to December 1990............................................... 100

B-42 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1991 to December 1996............................................... 101

B-43 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1957 to December 1959............................................... 101

B-44 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1960 to December 1963............................................... 101

B-45 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1964 to December 1966............................................... 102

B-46 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1967 to December 1981............................................... 102

B-47 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1982 to December 1990............................................... 102

B-48 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1991 to December 1996............................................... 103
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B-49 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1957 to December 1959 ............................................ 103

B-50 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1960 to December 1964 ............................................ 103

B-51 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1965 to December 1966 ............................................ 104

B-52 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1967 to December 1981 ............................................ 104

B-53 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1982 to December 1990 ............................................ 104

B-54 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1991 to December 1996 ............................................ 105

B-55 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1957 to December 1959 ............................................ 105

B-56 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1960 to December 1963 ............................................ 105

B-57 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1964 to December 1966 ............................................ 106

B-58 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1967 to December 1981 ............................................ 106

B-59 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1982 to December 1990 ............................................ 106

B-60 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1991 to December 1996 ............................................ 107

B-61 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1957 to December 1960 .................................................... 107

B-62 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1961 to December 1965 .................................................... 107

B-63 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1966 to December 1970 .................................................... 108

B-64 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1973 to December 1979 .................................................... 108

B-65 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1957 to December 1965 .................................................... 108

B-66 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1966 to December 1970 .................................................... 109

B-67 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1971 to December 1996 .................................................... 109

B-68 Predicted and observed 50th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 90Sr, January 1974 to December 1992........................................ 109

B-69 Predicted and observed 84th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 90Sr, January 1974 to December 1992........................................ 110

B-70 Predicted and observed 50th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 103Ru, January 1974 to December 1992...................................... 110

B-71 Predicted and observed 84th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 103Ru, January 1974 to December 1992 ...................................... 110

B-72 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F 90Sr, January 1993 to December 1995 ........................................................ 111

B-73 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F 90Sr, January 1993 to December 1995 ........................................................ 111
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ACRONYMS AND ABBREVIATIONS 

AMAD activity median aerodynamic diameter

ATR Advanced Test Reactor

AVLIS Atomic Vapor Laser Isotope Separation

AWE Atomic Weapons Employer

CFR Code of Federal Regulations

Ci curie

cm centimeter

cpm counts per minute

CWT chest wall thickness

d day

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

F fast (absorption type) 

g gram

GSD geometric standard deviation

HEPA high-efficiency particulate air

HEU highly enriched uranium

HPGe hyperpure germanium

hr hour

HT elemental tritium (tritiated gas) 

HTO tritiated water vapor

ICP-MS inductively coupled plasma–mass spectrometry

ICRP International Commission on Radiological Protection

IDOT Internal Dosimetry Tool

IMBA Integrated Modules for Bioassay Analysis

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

KPA kinetic phosphorescence analysis

L liter

LANL Los Alamos National Laboratory

LLNL Lawrence Livermore National Laboratory

LPTR Livermore Pool-Type Reactor

LSC liquid scintillation counting

M moderate (absorption type) 

MDA minimum detectable amount

MeV megaelectron-volt, 1 million electron-volts
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MFAP mixed fission and activation product

MFP mixed fission product

mg milligram

min minute

mrem millirem

MT metal tritide

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NTS Nevada Test Site

NU natural uranium

OBT organically bound tritium

ORAU Oak Ridge Associated Universities

pCi picocurie

R&D research and development

RTNS Rotating Target Neutron Source

RU recycled uranium

s second

S slow (absorption type)

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

SS super S (absorption type)

TBD technical basis document

TRIGA Training, Research, Isotopes, General Atomics

TRU transuranic

USC United States Code

WB whole body

WBC whole-body counter

yr year

μCi microcurie

μg microgram

μm micrometer
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5.1 INTRODUCTION

Technical basis documents (TBDs) 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 U.S. Department of Energy (DOE) or Atomic Weapons Employer 

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

relevant information is obtained about the affected DOE or AWE facility(ies), such as changing 

scientific understanding of operations, processes, or procedures involving radioactive materials. 

These documents may be used to assist NIOSH staff in the evaluation of Special Exposure Cohort 

(SEC) petitions and the completion of individual dose reconstructions under Part B of the Energy 

Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA).

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

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

“AWE facility” or a “DOE facility.” The term “AWE facility” is defined in EEOICPA to mean “a facility, 

owned by an atomic weapons employer, that is or was used to process or produce, for use by the 

United States, material that emitted radiation and was used in the production of an atomic weapon, 

excluding uranium mining or milling” [42 United States Code (USC) 7384I(5)]. On the other hand, a 

DOE facility is defined as “any building, structure, or premise, including the grounds upon which such 

building, structure, or premise is located—(A) in which operations are, or have been, conducted by, or 

on behalf of, the [DOE] (except for buildings, structures, premises, grounds, or operations ... 

pertaining to the Naval Nuclear Propulsion Program); and (B) with regard to which the [DOE] has or 

had—(i) a proprietary interest; or (ii) entered into a contract with an entity to provide management and 

operation, management and integration, environmental remediation services, construction, or 

maintenance services” [42 USC 7384I(12)]. The DOE determines whether a site meets the statutory 

definition of an AWE facility and the U.S. Department of Labor (DOL) determines if a site is a DOE 

facility and, if it is, designates it as such.

Under EEOICPA, a Part B cancer claim for benefits must be based on an energy employee’s eligible 

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

designated time period and location (i.e., a “covered employee with cancer”). After DOL determines 

that a claim meets the eligibility requirements under Part B of EEOICPA, DOL transmits the claim to 

NIOSH for a dose reconstruction. EEOICPA provides, among other things, guidance on eligible 

employment and the types of radiation exposure to be included in an individual dose reconstruction. 

Under EEOICPA, eligible employment at a DOE facility includes individuals who are or were 

employed by DOE and its predecessor agencies, as well as their contractors and subcontractors at 

the facility [42 USC 7384I(11)]. Also under EEOICPA, the types of exposure to be included in dose 

reconstructions for DOE employees are those radiation exposures incurred in the performance of 

duty. As such, NIOSH includes all radiation exposures received as a condition of employment at DOE 

facilities in its dose reconstructions for covered employees, which may include radiation exposures 

related to the Naval Nuclear Propulsion Program at DOE facilities, if applicable. This is because 

NIOSH does not determine the fraction of total measured radiation exposure at a DOE facility that is 

contributed by the Naval Nuclear Propulsion Program at the DOE facility during a specified period of 

time for inclusion in dose reconstruction.

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

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

dose reconstructions for covered employees [NIOSH 2010a]:

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

conventional structures, and

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

reasons.
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5.1.1 Purpose

This TBD discusses historical practices at the Lawrence Livermore National Laboratory (LLNL) and 

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

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

This document provides a uniform and consistent approach to assessing occupational internal dose at 

LLNL for dose reconstructions for NIOSH in relation to EEOICPA. It provides guidance to dose 

reconstructors on input parameters that are specific to LLNL employees, as well as the approach for 

employees with either missing or no monitoring information.

This TBD includes guidance on selecting source terms, interpreting in vivo and in vitro measurement 

results, and instructions for assessing dose for monitored and unmonitored employees.

5.1.2 Scope

Occupational internal dose refers to exposures workers received inside the facilities at LLNL from 

ingestion, inhalation, or absorption of radioactive materials in the course of work. For workers who 

were employed at LLNL, this section provides site-specific information to the dose reconstructor for 

reconstructing radiation doses received through multiple pathways of exposure including inhalation, 

ingestion, and absorption. In some cases, exposures occurred via wounds and were not restricted to 

intakes inside buildings.

Three classes of LLNL employees have been added to the SEC as described in the section below.

The addition of these classes of employees has limited the scope of potential internal dose 

reconstructions at LLNL to the following.

• Internal dose from tritium, 1957 to present;

• Internal dose from mixed fission and activation products (MFAPs), 1974 to present;

• Internal dose from americium, 1957 to present;

• Internal dose from plutonium, 1957 to present;

• Internal dose from uranium, excluding 233U, 1958 to present; and

• Internal dose from other radionuclides at LLNL can be reconstructed for specific radionuclides

as described in later sections. 

The dose reconstructor should refer to the appropriate section later in this document for when the 

internal dose for a particular radionuclide can be reconstructed.

The internal dose from americium, plutonium, and uranium can be extended for employment before 

the above-listed periods based on a worker’s later bioassay results and work history.

The remainder of this section describes the SEC classes, certain historical events that are important 

to internal dose reconstruction, and the overall approach to internal dose reconstruction for LLNL 

workers. Section 5.2 discusses source term information. Section 5.3 discusses in vitro and in vivo

bioassay measurement methods. Section 5.4 describes significant incidents with internal dose 

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

justification, or clarification of the associated information, are presented in Section 5.5. 

Attachment A provides materials for reconstruction of occupational internal dose for monitored 

workers. Attachment B provides co-exposure intakes. Attachment C describes scaling of co-exposure

dose intakes consistent with type and duration of exposures.
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5.1.3 Special Exposure Cohort

The Secretary of the U.S. Department of Health and Human Services has added the following classes 

to the SEC:

January 1, 1950, through December 31, 1973

Employees of the DOE, its predecessor agencies, and DOE contractors or 

subcontractors who were monitored for radiation exposure while working at the 

Lawrence Livermore National Laboratory from January 1, 1950, through December 31, 

1973, for a number of workdays aggregating at least 250 workdays or in combination 

with workdays within the parameters established for one or more other classes of 

employees in the Special Exposure Cohort [Leavitt 2008, p. 3].

January 1, 1950, through December 31, 1973

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

contractors and subcontractors who worked at the Lawrence Livermore National 

Laboratory in Livermore, California from January 1, 1950 through December 31, 1973, 

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

[Sebelius 2010, p. 3]. 

January 1, 1974 through December 31, 1989

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

and subcontractors who worked in any area at the Lawrence Livermore National 

Laboratory in Livermore, California, during the period from January 1, 1974 through 

December 31, 1989, 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 SEC

[Burwell 2016, p. 3]. 

NIOSH determined that it cannot reconstruct with sufficient accuracy (1) MFAP doses to all workers 

for 1950 to 1973 [NIOSH 2007], (2) MFAP doses to workers who were not monitored for radiation 

exposure from 1950 to 1973 [NIOSH 2010b], and (3) internal doses to all workers from 233U from 1958

to 1989 [NIOSH 2016].

In effect, the second designation encompasses the first. The latter two SEC classes include all 

workers during the SEC period. Because of the identified dose reconstruction infeasibility, all dose 

reconstructions for all workers having employment during an SEC period are considered partial dose 

reconstructions. If monitoring data are available for workers in the SEC, dose is to be assigned as 

appropriate based on such data. However, such dose reconstructions are still considered partial dose 

reconstructions because NIOSH has determined that exposure to MFAPs and 233U, during the 

relevant SEC periods, cannot be bounded.

5.2 SOURCE TERM

LLNL workers handled a variety of radionuclides as part of their routine work. The key elements in the 

source term were plutonium and tritium, although others were used at various times and in various 

forms. Gloveboxes, extensive filtered ventilation systems, and other engineering controls were used 

to minimize personnel exposures from the outset of operations at LLNL [Balanda 1966]. 

Administrative controls were implemented to minimize the potential for an intake from handling 

radioactive materials. A majority of the buildings at LLNL handled or processed unsealed sources of 
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radioactive materials; persons who were assigned to operate processes or laboratory research had a 

potential for exposure [ORAUT 2005a; Mansfield 1989]. Administrative and other support personnel in 

the individual buildings had less potential for exposure because of engineering and administrative 

controls. For dose reconstruction, it can be assumed that internal source terms were introduced at 

LLNL when it began operations on September 2, 1952.

Before the early 1960s, the only bioassay methodology LLNL used to monitor employees for intakes 

of radionuclides was urine bioassay, and the primary focus was excreted tritium. A review of previous 

bioassay samples indicates that bioassay monitoring started in March 1957. The analysis of bioassay 

samples continues to this day for plutonium, americium, uranium, MFAPs, a variety of tracer 

radionuclides, iodine, and tritium [Mansfield and Wood-Zika 2000]. Air monitoring in workplaces has 

been a common surveillance method. A brief description of the building processes is provided in 

ORAUT-TKBS-0035-2, Lawrence Livermore National Laboratory – Site Description [ORAUT 2005b, 

Table 2-2]. The buildings that were assigned to handle plutonium are noted. Buildings in which 

uranium was processed and handled are also described. ORAUT [2005b, Table A-1] provides a cross 

reference of main site building number changes that took place in 1966. The dose reconstructor can 

refer to this list if needed when assessing building location information for a worker’s claim.

In vivo methodologies began on an investigational basis in 1964 with focus on high-energy gamma 

emitters (i.e., greater than 200 keV). In the 1970s attempts were made to detect low-energy photon 

emitters (i.e., 60-keV gamma radiation from the decay of 241Am and even plutonium L-shell X-rays).

LLNL can provide a broad spectrum of in vivo counting services with varying degrees of detectability 

[Mansfield and Wood-Zika 2000]. 

A review of in-house procedures LLNL used to assess the concentration of radioactivity in urine 

indicates that quality control steps were an integral part of the process [Dupzyk 1979]. For example, 

LLNL ran duplicates consistently, and comparisons of results to “known quantities” were a critical 

step. Therefore, in vitro results from in-house processing are generally reliable. However, 

interpretation of those results can be difficult, primarily because they might not have considered the 

contribution of environmental radioactivity (i.e., uranium, thorium). Because sample collection could 

have occurred at work (e.g., “in-field” tritium analyses, Monday morning urines), cross-contamination 

could be an issue.1

1 Twenty-four-hour collections typically took place at the employee’s home on weekends. However, spot 

samples were often collected in site restrooms [Mansfield 1989]. 

The primary LLNL missions have been weapons research and development (R&D), controlled nuclear 

weapons research, peaceful uses of nuclear explosives, biomedical research, and laser fusion 

research. In addition, work in nonnuclear technologies and materials testing has been ongoing since 

at least 1976 [LLL ca. 1978]. LLNL conducted aboveground (atmospheric) and underground tests at 

both the Pacific Proving Ground and the Nevada Test Site (NTS) from 1953 to 1992. Although the 

tests were conducted off site, test planning and preparation occurred at LLNL. In 1971, LLNL 

managed the CANNIKIN test at Amchitka Island, Alaska.

Unless site-specific information is available, the particle size is assumed to be 5-μm activity median 

aerodynamic diameter (AMAD), as recommended in ICRP [1995, paragraph 5]. 

5.2.1 Plutonium

If a monitoring result refers to weapons-grade plutonium or simply plutonium, dose reconstructors 

should use the isotopic mix listed in Tables 5-1 and 5-2 for material aging times [Mansfield and WoodZika 2000]. From September 1952 through 1957, fresh plutonium should be assumed. For 1958 

through 1962, assume a 5-year-old plutonium mixture. From 1963 forward, 10-year-old plutonium 
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should be assumed. For assessing absorption type SS, refer to ORAUT-OTIB-0049, Estimating

Doses for Plutonium Strongly Retained in the Lung [ORAUT 2020]. 

Table 5-1. Specific activity (Ci/g) for specified ages of weapons-grade plutonium.

a

Nuclide Fresh 5-yr 10-yr 20-yr

Pu-238 6.85E–03 6.59E–03 6.33E–03 5.85E–03

Pu-239 5.81E–02 5.81E–02 5.80E–02 5.80E–02

Pu-240 1.37E–02 1.37E–02 1.37E–02 1.37E–02

Pu-241 5.98E–01 4.70E–01 3.69E–01 2.28E–01

Pu-242 1.57E–06 1.57E–06 1.57E–06 1.57E–06

Am-241 0.00E+00 4.24E–03 7.54E–03 1.21E–02

Pu-239+240 7.18E–02 7.18E–02 7.17E–02 7.17E–02

Pu-alpha 7.87E–02 7.84E–02 7.80E–02 7.76E–02

a. Source: Mansfield and Wood-Zika [2000]. 

Table 5-2. Activity ratios in mixture for specified ages of weapons-grade plutonium.

a

Nuclide Fresh 5-yr 10-yr 20-yr

Pu-239+240:Am-241 Not applicable 1.69E+01 9.51E+00 5.93E+00

Pu-239+240:Pu-238 1.05E+01 1.09E+01 1.13E+01 1.23E+01

Pu alpha:Pu-239+240 1.10E+00 1.09E+00 1.09E+00 1.08E+00

Pu alpha:Pu-238 1.15E+01 1.19E+01 1.23E+01 1.33E+01

Pu-241:Pu alpha 7.60E+00 6.00E+00 4.73E+00 2.94E+00

a. Source: Mansfield and Wood-Zika [2000]. 

5.2.2 Uranium

If a monitoring result refers to “uranium,” dose reconstructors can use the isotopic mix in Table 5-3 for 

enrichment levels [Mansfield and Wood-Zika 2000]. LLNL received uranium metal from the Y-12 plant 

that was known to contain recycled uranium (RU) contaminants including technetium, neptunium, 

thorium, and plutonium [1]. ORAUT-TKBS-0014-5, Y-12 National Security Complex – Occupational 

Internal Dose [ORAUT 2012], discusses the typical concentration of RU contaminants. Table 5-4 

provides the list of RU ratios to be applied for uranium material from Y-12. The RU ratios in Table 5-4 

reflect upper bound values. Given that there is uncertainty in the specific batches of uranium LLNL 

received, the upper bound values in Table 5-4 should be used for dose reconstruction. LLNL also 

used natural uranium (NU) from the Feed Materials Production Center for the Atomic Vapor Laser 

Isotope Separation (AVLIS) project. The AVLIS project converted kilogram quantities of NU to slightly 

enriched uranium, up to 5%, as a demonstration plant. The AVLIS project operated in the 482 

complex and Buildings 175 and 177 from 1973 through 1999. Table 5-5 provides the list of RU ratios 

to be applied for uranium associated with the AVLIS project. In the absence of specific information, 

the Y-12 RU ratios in Table 5-4 should be used.

If the bioassay results are expressed in units of mass, the dose reconstructor should select the 

isotopic mix equivalent to NU in Table 5-3 in almost all cases at LLNL, in the absence of claim-specific 

information. The uranium in question for bioassay results expressed in units of mass would typically 

be either depleted uranium (DU) or NU. 

Highly enriched uranium (HEU) came from the Y-12 Plant and was used in Building 231 and the 

Oralloy Shop in Building 321C. HEU was restricted at LLNL, both in areas and amounts, due to 

criticality concerns. HEU machining was done in Building 321C in enclosures, hoods, or under closecapture ventilation systems. Building 231 is an experimental manufacturing, assembly, test, and 

materials-handling facility. Building 231 also housed R&D activities. Building 241, where materials 

science research was conducted, had an old glovebox line in Room 1600 that contained enriched 

uranium (EU). As a result, internal exposure potential to HEU in Building 241 would have been 

remote. Building 327 is a radiography facility that conducted nondestructive testing of materials 
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Table 5-3. Isotopic information for various uranium enrichment levels [Mansfield and 

Wood-Zika 2000].a,b,c

Mixture Radionuclide

Mass

fraction

Activity

fraction

Specific activity

of mix (Ci/g)

DU U-238 9.975E–01 9.02E–01 3.72E–07

DU U-235 2.50E–03 1.45E–02 3.72E–07

DU U-234 5.00E–06 8.40E–02 3.72E–07

NU U-238 9.93E–01 4.77E–01 7.00E–07

NU U-235 7.20E–03 2.23E–02 7.00E–07

NU U-234 5.60E–05 5.00E–01 7.00E–07

AVLIS <5% EU U-238 9.50E–01 4.11E–01 7.78E–07

AVLIS <5% EU U-235 5.00E–02 1.39E–01 7.78E–07

AVLIS <5% EU U-234 5.60E–05 4.50E–01 7.78E–07

95% HEU U-238 3.96E–02 1.98E–04 6.72E–05

95% HEU U-235 9.50E–01 3.06E–02 6.72E–05

95% HEU U-234 1.04E–02 9.69E–01 6.72E–05

a. Source: Mansfield and Wood-Zika [2000]. 

b. When “U238,” “U28,” or “U38” was reported, these were units of total U mass [Mansfield 2006a]. 

c. Notations in bioassay records indicate that starting in 2000 a mass spectroscopy system was used at 

LLNL. Results from the mass spectroscopy system are reported as mass of U-238. These U-238 

bioassay results need to be adjusted for total uranium by dividing by the appropriate U-238 activity 

fraction and multiplying by the appropriate specific activity.

Table 5-4. Y-12 Plant recommended RU contaminant ratios default levels based on upper levels of 

expected ranges.

a

Isotope

HEU & 

Oralloy

(pCi/μg U)

DU, NU, and 

2% EU

(pCi/μg U)

HEU & 

Oralloy

(dpm/μg U)

DU, NU, and 

2% EU

(dpm/μg U)

HEU and 

Oralloy 

(pCi/pCi or 

dpm/dpm 

total U)

DU, NU, and 

2% EU

(pCi/pCi or 

dpm/dpm 

total U)

Tc-99 4 0.3 8.88 0.666 4.00E–02 0.15

Th-228 0.5 0.003 1.11 0.00666 5.00E–03 1.50E–03

Np-237 0.6 0.003 1.332 0.00666 6.00E–03 1.50E–03

Pu-238 0.2 Not applicable 0.444 Not applicable 2.00E–03 Not applicable

Pu-239 0.2

(Oralloy only)

0.003 0.444

(Oralloy only)

0.00666 2.00E–03

(Oralloy only)

1.50E–03

a. Source: ORAUT [2012]. 

Table 5-5. Feed Materials Production Center RU 

contaminant ratios for NU.

a

Isotope pCi/μg U dpm/μg U

pCi/pCi or

dpm/dpm total U

Pu-239 0.00629 0.01396 0.00921

Np-237 0.00250 0.00555 0.00366

Tc-99 0.15435 0.34266 0.22599

a. Source: ORAUT [2017]. 

including HEU. Because of the nature of the work in Building 327, internal exposure potential to HEU 

would be low. LLNL performed in vitro and in vivo analysis for HEU. Uranium bioassay results for 

HEU would be expressed in activity.

5.2.3 Tritium

Tritium was encountered in several forms: tritiated water vapor (HTO), elemental tritium or tritiated 

gas (HT), organically bound tritium (OBT), and metal tritides (MTs). Each form has unique 

characteristics. Tritium handling at LLNL was first performed in Building 231 in the early to mid-1950s.
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Gas fill operations were performed along with other experimental activities. Tritium handling at 

Building 331 began in 1958. The second phase of Building 331 was completed in 1962. Tritium-filled 

microspheres were used at LLNL in Buildings 381 and 391 for the Shiva Nova II experiment in 1977.

Stable MTs were handled in Building 331, Rooms 154, 150, 138, 145 [Odell and Toy 1975], 

Building 212, and Building 292. Outside these buildings, in cases where exposure to OBT or MTs is 

not implied by case-specific information, assume HTO. Tritium gas work was done in processing 

hoods. Compounds of tritium, namely MTs, were usually worked in gloveboxes. The glovebox air was 

filtered because experimentation with MTs required dry, inert environments to prevent contamination 

of the material under experimentation with water from the environment. Guidance for estimating doses 

for OBT and MTs is available in ORAUT-OTIB-0066, Calculation of Dose from Intakes of Special 

Tritium Compounds [ORAUT 2007a]. 

Site locations for potential exposure to tritium are listed in Attachment A. Bioassay results for tritium 

are usually available in the dosimetry records.

5.2.3.1 Metal Tritides

Tritium exposures in the form of MT aerosols were primarily titanium tritide targets for the accelerator 

beams in Buildings 212 and 292. Operations at Building 212 were the forerunner to Rotating Target 

Neutron Source II (RTNS II) (Building 292). Tritium operations were conducted at Building 212 from 

1966 until 1978 when the RTNS II became operational. Operations at Building 212 were significantly 

reduced after this time and ended in 1987. Operations at Building 292 began in 1978 and ended in 

1987. This facility took over tritium operations that were previously conducted at Building 212. The 

titanium MT targets were handled in Building 331, Rooms 150, 138, and 145. The personnel with the 

potential for exposure to the MT targets were the chemists, researchers, health physics technicians, 

and accelerator operators and mechanical technicians who worked near the accelerators.

The targets in Buildings 212 and 292 were changed primarily by the accelerator operators.

Occasionally, some of the facility support staff, usually mechanical technicians, changed the targets.

The titanium tritide targets for the RTNS I accelerator in Building 212 were changed inside a hood that 

was exhausted through a high-efficiency particulate air (HEPA) filter. The target was inserted in one 

side of the hood and changed through glove ports on the other side of the hood. Individuals involved 

in target change wore gloves, protective clothing (typically a laboratory coat), and a half-mask 

respirator. A Hazards Control Monitor (Health and Safety Technician) was required to be present for 

target change. The Hazards Control Monitor was responsible for making radiation measurements and 

taking air samples to check for airborne contamination. Bioassay samples were collected within 

4 hours after the target handling operation. Because of the larger physical size of the titanium tritide 

targets for RTNS II in Building 292, they were not changed in a hood enclosure as were the RTNS I 

targets. The workers changing the targets wore anticontamination clothing, plastic shoe covers, a 

paper hood, gloves, a plastic face shield, and a half-mask respirator. Health and Safety Technicians 

provided radiation monitoring support (making radiation measurements, taking swipes, and collecting 

air samples) during the operation. After removing protective clothing, workers were swiped for clothing 

and skin contamination by the Health and Safety Technician. Bioassay samples were collected within 

4 hours after the target handling operations [Myers 2008]. 

Absorption type M should be assumed for assigning internal dose from titanium tritides. Based on 

available information, there is no indication that MT of type S was a possibility at LLNL.

ORAUT-OTIB-0066 [ORAUT 2007a] provides guidance on the evaluation of MT intakes. MTs are 

referred to as “tritium particulates.” In addition to potential exposure in the above-listed facilities, the 

claimant telephone interviews for the claim, if any, can provide indications that a person was exposed 

to MTs. Building exposure locations are also available in tritium dosimetry records as an aid to the 

dose reconstructor in assigning tritium dose from MTs. 
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Based on the above, the following is a list of buildings and job categories for which tritium exposures 

in the form of MT at LLNL should be considered.

1. Building 212 (RTNS I accelerator): 

a. Accelerators operators, 

b. Researchers, 

c. Mechanical technicians who performed work on the accelerators, and

d. Health and Safety Technicians (health physics technicians, hazard control monitors).

2. Building 292 (RTNS II accelerator): 

a. Accelerators operators, 

b. Researchers, 

c. Mechanical technicians who performed work on the accelerators, and

d. Health and Safety Technicians (health physics technicians, hazard control monitors).

3. Building 331 (Tritium Facility), Rooms 150, 138, and 145: 

a. Chemists, and

b. Health and Safety Technicians (health physics technicians, hazard control monitors).

4. Cases where a claimant telephone interview provide indications that a person was exposed to 

MTs.

5.2.3.2 Organically Bound Tritium

Tritium exposures in the form of OBT are primarily found in pump oils associated with fume hoods.

Exposure to OBT would occur during maintenance if contact with the skin was made. If claim 

information indicates the need for assigning tritium dose based on OBT, guidance on the calculation 

of doses from OBT is given in ORAUT-OTIB-0066 [ORAUT 2007a].

5.2.4 Other Limited-Exposure Radionuclides

LLNL has always been a center for research. As such, small-scale use of various radionuclides (in 

terms of either the number of persons or the activity of the source) that is not addressed above has 

occurred throughout the history of LLNL. The internal dose potentials due to these nuclides were 

remote because of the requirements for handling unsealed sources in fume hoods, hot cells, 

gloveboxes, and dryboxes. LLNL can provide a broad spectrum of in vivo counting services with 

varying degrees of detectability [Mansfield and Wood-Zika 2000]. This is also evident by the 

radionuclides assessed by lung counting or whole-body (WB) counting (see Table 5-9 in 

Section 5.3.3). The dosimetry records should provide this information to the dose reconstructor as 

part of the case-specific information. In addition, Attachment C provides a method to scale the 

provided intakes to estimate a credible exposure to individuals who were exposed to rarely 

encountered radionuclides.

5.3 MEASUREMENT METHODS

In vivo measurement methods began at the site as far back as 1964 when LLNL used sodium iodide 

detectors connected to simple multichannel analyzers to evaluate 40K and fission products in humans.

In 1964, a scanning bed, referred to as the “RIDL,” was in use at LLNL. This device was a shadow 

shield counter equipped with a 3- by 2-in. sodium iodide detector for detection and quantification of 

relatively high energy gamma photons (i.e., greater than 200 keV). It had some capability for 
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determining the source of radionuclide deposition in subjects. This device was calibrated with a water 

tank phantom spiked with 137Cs and 40K [Hickman no date]. 

In April 1965, thin crystal studies began [LRL 1965, pp. 18–23]. In 1965 and 1966, LLNL identified 

65Zn in people working at the Building 153 cyclotron, which dominated the study groups at that time 

[Summary of health physics activities ca.1988]. 

Workers were selected for WB or specific organ counting based on direction by the program 

supervisor and health physics staff. Baseline and termination counts occurred as early as 1966 

[Balanda 1966]. 

From at least 1957, in vitro measurement methods were in use including routine monitoring occurring 

annually and semiannually during those early days [2]. LLNL selected workers for internal monitoring 

based on the potential for internal exposure. Employees whose work with radionuclides warranted 

routine testing for internal radiation were determined based on the supervisor of each operation and 

the Radiation Safety Section Health Physicist assigned to the area [Balanda 1966]. Building 331 had 

a monthly monitoring program in place. During weapons test support periods, routine sampling 

frequencies appear to have been increased to at least quarterly for involved employees, with weekly 

sampling for Building 331 (for tritium) and monthly sampling for Building 251 [Balanda 1966]. Through 

the 1980s, the program supervisor and the area Health Physicist selected workers for internal 

monitoring based on the potential for internal exposure [Mansfield 1989]. The selection of workers for 

internal monitoring based on the potential for internal exposure continues at LLNL [Mansfield and 

Wood-Zika 2000]. 

Bioassays were supplemented with workplace airborne monitoring [Mansfield 1989]. In addition, nasal 

smears were collected after incidents as a means of assessing intake potential. The dose 

reconstructor might come across some of these data in worker dosimetry records [Mansfield 1989, 

Appendix A]. 

Site 300 had a routine bioassay monitoring program as early as 1971 in which all Site 300 personnel 

received an employment start and termination WB count. In addition, workers in the firing areas and 

other radioactive materials handling (i.e., process) areas were bioassayed for gross alpha and beta 

and underwent fluorometric analysis for uranium. Additional WB count or bioassays were requested at 

the discretion of the Health Physicist [Greenhouse and Kloepping 1971]. Starting in 1980 at Site 300, 

all workers in, or who could be expected to have spent a considerable portion of their time in, areas 

where an airborne uranium hazard might exist were required to receive an annual lung count and a 

semiannual urine analysis. Other schedules might have been required if radionuclides other than 

uranium were used and considered a potential internal hazard [Straume 1980]. Tables 5-6 and 5-7 list 

in vivo and in vitro sampling frequencies for various employment periods. In the case of in vitro

methods, sample sizes could have been spot (single void), 24-hour collections, or simulated 24-hour 

collections with creatinine corrections performed if the results were indicative of significant intakes 

[Mansfield and Wood-Zika 2000]. 

Selected records report the presence of 239Pu and 240Pu as determined by accelerator mass 

spectrometry (also called CAMS – Center for Accelerator Mass Spectrometry) and are reported in 

units of activity that were converted from units of mass. The accelerator mass spectrometry data in a 

worker’s records could come primarily from two sources (or both): (1) For select individuals with 

known or suspected uptakes, an intercomparison was performed between alpha spectrometry data 

and accelerator mass spectrometry data to validate the mass spectrometry techniques, and (2) in the 

case of suspected uptakes or for follow-up, mass spectrometry was used because it typically has a 

detection limit 10 times to as much as 100 times lower than alpha spectrometry. Where two positive 

values are reported for the same sample, the higher of the two should be used. If two values are 

reported at the minimum detectable amount (MDA) for the same sample, the lower of the two values 
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Table 5-6. Routine in vitro sampling frequencies.

a,b,c

Radionuclide Period Building

Measurement

type Frequency

Gross alpha, beta (LSC) 2000–present Sitewide Urine As applicable

HTO 1989d Sitewide Urine Weekly, monthly

HTO 2000–present Sitewide Urine Quarterly, monthly, biweekly, or weekly

Workplace-specific 1966d 101 Urine Quarterly, annual

Workplace-specific 1966d 102 Urine Annual

Workplace-specific 1966d 110 Urine Annual

Workplace-specific 1966d 112 Urine Annual

Workplace-specific 1966d 114C Urine Annual

Workplace-specific 1966d 117 Urine Quarterly, annual

Workplace-specific 1966d 121 Urine Semiannual

Workplace-specific 1966d 125 Urine Semiannual

Workplace-specific 1966d 127 Urine Semiannual

Workplace-specific 1966d 170 Urine Annual

Workplace-specific 1966d 171 Urine Annual

Workplace-specific 1966d 172 Urine Weekly, monthly

Workplace-specific 1966d 173B Urine Annual

Workplace-specific 1966d 190 Urine Semiannual, monthly

Workplace-specific 1966d 193 Urine Annual

P-32 1989d Sitewide Urine Monthly

P-32, C-14 (LSC) 2000–present Sitewide Urine As applicable

Pu 2000–present Sitewide Urine Semiannual

TRU 2000–present Sitewide Urine Annual, semiannual

U 1980d 231 Urine Quarterly

U 1983d 251 Urine Quarterly

U 1985d Site 300 Urine Semiannual (collected on Mondays)

U 2000–present Sitewide Urine Quarterly, monthly

Weapons-grade Pu 1989d Sitewide Urine Semiannual

a. Sources: Mansfield and Wood-Zika [2000]; Balanda [1966, Appendix A]; LLNL [1987]; Ozaki [1980]; Wilson [1982]; 

Leahy [1983]; LRL [1961, 1964]; Gibson [1985]. 

b. LSC = liquid scintillation counting; TRU = transuranic.

c. These are typical in vitro monitoring frequencies; they varied depending on workplace conditions, suspected inhalation 

events, and positive or suspected positive bioassay results.

d. Periods listed as a single year were taken from annual reports that did not indicate a beginning date for the stated 

bioassay period.

Table 5-7. In vivo sampling frequencies.

a,b

Radionuclide Period Building

Measurement

type Frequency

As detected 2000–present Sitewide WB scans Annual

I-131 1989–1999c Sitewide Thyroid Monthly

I-131 2000–present Sitewide Thyroid count Coordinated with work schedule

MFAPs 1964–present Sitewide WB scans Not available

Pu, Am 1970–present Sitewide Lung count Annual

Pu, weapons-grade Pu 1989c Sitewide Lung Annual

Pu-239 1961–present Sitewide Wound Incident-specific

U 1985c Site 300 Lung count Annual

U, Th 1987 Bldg. 321 Lung count Annual

U, Th 2000–present Sitewide Lung count Annual, semiannual

a. Sources: Mansfield and Wood-Zika [2000], Balanda [1966, Appendix A], LLNL [1987], Ozaki [1980], Wilson [1982], 

Leahy [1983], LRL [1961, 1964], Gibson [1985]. 

b. These are typical in vivo frequencies; they varied depending on workplace conditions, suspected inhalation events, and 

positive or suspected positive bioassay results.

c. Periods listed as a single year were taken from annual reports that did not indicate a start date for the stated bioassay 

period.
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should be used. If reported as 239Pu, the alpha spectrometry value is 239Pu and 240Pu. In the case of 

accelerator mass spectrometry, the sum of the 239Pu and 240Pu should be used to compare to the 

alpha spectrometry value.

5.3.1 Analysis of Gross Alpha Measurements

Urine samples were analyzed for the presence of alpha radiation from about 1957 through 1996 [3] 

using three possible procedures, including “Analytical Method for Determining Gross Alpha Activity in 

Urine,” “Determination of Plutonium and Americium in Urine by Separation,” and “Urinalysis for 

Curium by Electrodeposition” [Dupzyk 1979]. The results were reported as gross alpha, but all three 

involved the chemical separation of several different elements depending on the method. The records 

are not clear about which method was used; they require some review by the dose reconstructor. The 

methods were capable of separating uranium from the sample before analysis to reduce the potential 

interference [Dupzyk 1979].

The three urine procedures involving alpha (not including uranium) are described below:

1. “Analytical Method for Determining Gross Alpha Activity in Urine” [Dupzyk 1979, p. 43]. This 

method involved bismuth phosphate coprecipitation and then coprecipitation with lanthanum 

fluoride. Thorium, americium, curium, actinium, and neptunium are carried through to 

determination with this procedure. This procedure also captured plutonium.

2. “Determination of Plutonium and Americium in Urine by Separation” [Dupzyk 1979, p. 7]. This 

method was specific for plutonium and americium. It involved precipitation with ammonium 

hydroxide, dissolution in nitric, and then adsorption on an anion exchange column. The 

americium did not stick on the column so it came right through with the wash. If gross alpha 

was to be measured as well as the plutonium, the effluent was counted on a proportional 

counter. The plutonium was eluted from the column, electroplated, and counted in a 

scintillation counter.

3. “Urinalysis for Curium by Electrodeposition” [Dupzyk 1979, p. 32]. This method had the same 

first steps as the first procedure above, but included a third precipitation using lanthanum

hydroxide. This separated curium from all the other alpha emitters. The precipitate was 

electroplated, then counted using pulse-height analysis, low-background proportional counting, 

or autoradiography.

The gross alpha procedure described above was used until around mid-2002.

Notations in bioassay records indicate that in mid-2002 the LLNL gross alpha procedure was 

significantly changed to increase the sensitivity and usefulness of the method. A sequential procedure 

in which plutonium was first separated from the sample by anion exchange chromatography, followed 

by separation of the remaining actinides from the sample matrix by extraction chromatography, is now 

employed. The resulting sample is then analyzed by alpha spectroscopy.

From mid-2002 to late 2003, bioassay records contained analyte codes in the form of “GA-#.” These 

samples, indicated by GA-1, GA-2, or GA-3, represent low, middle, and high alpha energies and were 

reported along with a total gross alpha that represented the sum of activity in all three regions. In late 

2003, the low- and middle-energy regions (GA-1 and GA-2) were combined into a single region 

labeled GA-1. The new GA-1 region spanned alpha energies from about 3.9 MeV to 5.1 MeV, and 

would normally include 232Th, 238U, 235U, 234U, and 237Np. The GA-3 region spanned alpha energies 

from about 5.3 MeV to 6.1 MeV, and would normally include 241Am, 228Th, 244Cm, 249Cf, and some 

higher energy naturally occurring radionuclides [4]. As stated previously, plutonium was separated 

from these GA-# samples.
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 21 of 113

Claims that have GA-1, GA-2, and GA-3 bioassay results are to be assessed on a case-by-case 

basis. The dose reconstructor should contact the site lead for any claim that has GA-1, GA-2, and 

GA 3 bioassay results for further guidance.

The method described below applies only to gross alpha results before the GA-# analyte method was 

implemented in mid-2002.

Gross alpha results from before the adoption of the GA-# analyte method in mid-2002 were a mix of 

procedures 1 and 2 above. The database does not indicate which procedure was used. Sometimes 

the gross alpha analysis might have been specific for americium; at other times it included natural 

thorium and plutonium. Separated americium was used in Building 251 as early as 1955 for tracer 

studies. Separated americium is assessed as type M 241Am. To account for either possibility [5]:

1. If a worker has gross alpha and plutonium results on the same day (i.e., one sample analyzed 

sequentially):

a. Assess the plutonium results with any other plutonium results that might have been 

collected (regardless of whether they were collected on the same day as a gross alpha 

sample).

i. Assign the plutonium mixture to the plutonium results, and assign the dose.

b. Assume there is no plutonium in the gross alpha result and assess it as type M 241Am, 

and assign the dose.

2. If there is a gross alpha result without a plutonium result on the same day, assume it contains 

both plutonium and americium:

a. If there are also plutonium results for the individual:

i. Assess the plutonium results first for types M, S, and SS:

(1) Assess missed dose.

(2) Assess fitted dose.

(3) Assign the plutonium mixture to the plutonium results.

(4) Compare missed versus fitted dose on an annual basis and assign the 

larger of the two in each year.

ii. Assess the gross alpha results:

(1) If there are gross alpha results less than the MDA, assess missed dose:

(a) Subtract the plutonium MDA from the gross alpha MDA at the 

time of the last gross alpha sample.

(b) Use the adjusted sample result to assess type M 241Am missed 

dose (i.e., use one-half the result to assess a chronic intake).

(2) If there are gross alpha results >MDA, assess fitted dose:

(a) If there are plutonium results >MDA:

(i) Run a projection from the plutonium intakes calculated in 

2.a.i.2 to the dates of the positive gross alpha results.
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Use whichever absorption type was determined to give 

the largest dose in that step.

(ii) Subtract the projected plutonium results from the gross 

alpha results and assess the remainder as type M 241Am.

(b) If there are no plutonium results >MDA, do not adjust the gross 

alpha results; use them as reported and assess as type M 241Am.

(3) Compare missed versus fitted dose on an annual basis and assign the 

larger of the two in each year.

3. If there are gross alpha results and no plutonium results for the individual:

a. If the work location is Building 251:

i. Assess the gross alpha results twice, once assuming plutonium and once 

assuming type M 241Am: 

(1) Assign the plutonium mixture to the plutonium results. 

(2) Compare the plutonium mixture doses (types M, S, and SS) and the 

type M 241Am doses. Select whichever yields the largest dose to the 

organ of interest.

b. If the work location is not Building 251 or is unknown, assess the gross alpha results 

assuming plutonium: 

i. Assign the plutonium mixture to the plutonium results.

Exceptions occur if there is something in the file that shows the worker has been exposed to 

neptunium or curium. Another possible exception would involve thorium. LLNL performed some 

machining of natural thorium. Thorium machining was done in either enclosures, hoods, or closecapture ventilation systems. Natural thorium, along with DU and NU, was also used at Site 300 in high 

explosive tests. Indications are that WB counting was used to monitor for thorium intakes. If the 

records show that a worker was getting regular uranium analyses and then had a gross alpha 

bioassay analysis, that suggests that the worker was working with uncontained thorium, and the gross 

alpha result should be interpreted as thorium. The worker’s job description should be evaluated to 

determine if this is appropriate. If there is a WB count, the thorium intake calculated from the gross 

alpha urinalysis could be compared with the detection limit of the WB count.

5.3.2 Analysis of Gross Beta Measurements

Urine samples were also analyzed for beta-emitting radionuclides. The gross beta procedure included 

some separation chemistry that removed 40K [ORAUT 2006a] before counting. The procedure 

included a broad spectrum of other beta-emitting radionuclides that would have been captured and 

counted, although the percent recovery might have differed among the elements. Radioelements 

likely to have been captured and counted would have been strontium, barium, radium, all the rare 

earths, all the actinides, and most of the transition metals unless they form strong amine complexes 

(such as cobalt, nickel, copper, zinc, and silver). This means that beta-emitting progeny of thorium 

and uranium were likely to be included in the gross beta results. Ruthenium-103 and -106 were likely 

captured and counted [ORAUT 2006a]. LLNL used other methods to separate 32P, 14C, and 90Sr and 

listed the results for these specific nuclides in the results. A result described as gross beta 
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represented many potential nuclides. Where gross beta samples were reported in cpm, the dose 

reconstructor should assume a conversion of 0.4 cpm/dpm [ORAUT 2013]. 

There are three different source terms for beta/gamma-emitting radionuclides:

• MFAPs from a reactor (through 1980),

• The residual from weapons tests (through 1992), or

• Aged fission products and research-related radionuclides (after 1992).

Note that for intakes before 1981 both reactor mixtures through 1980 and weapons residue mixtures 

should be considered. The mixture that gives the highest dose to the organ of interest should be 

assigned for the relevant work periods. From 1981 through 1992, only weapons residue should be 

assumed for the gross beta results. Reactor mixtures should only be assumed through 1980 because 

the Livermore Pool-Type Reactor (LPTR), which began operations in December of 1957, was 

permanently shut down on March 31, 1980.

Dosimetry records may contain whole-body counter (WBC) information for a worker. WBC information 

can be used to limit the intakes determined by the three gross beta methods listed below. This can be 

done by multiplying the gross beta intakes determined using the three gross methods listed below by 

the ratio of the specific WBC radionuclide intake to the specific gross beta radionuclide intake.

5.3.2.1 Reactor Mixtures, 1974 to 1980

ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal Dose-Related Gross Beta 

and Gross Gamma Analyses [OTIB-0054; ORAUT 2015], discusses how to use gross beta urine 

bioassay results to assign intakes of MFAPs. Of the three types of gross beta analysis discussed in 

OTIB-0054, the gross beta analysis at LLNL is considered to be best described as “Gross beta 

analysis that includes all MFAPs except radioiodines and beta emitters naturally in urine such as 40K” 

(see Section 7.1.3 of OTIB-0054). The MFAP decay times assessed should be 10-day-old, 40-dayold, 180-day-old, and 1-year-old reactor fuel (Table 7-1a in OTIB-0054). Therefore, the following 

method is used to assign internal dose from gross beta results from reactor mixtures:

1. Determine the 90Sr intake by multiplying the gross beta results by the 90Sr average fraction of 

beta activity in urine sample for 10-day-old, 40-day-old, 180-day-old, and 1-year-old reactor 

fuel for the Advanced Test Reactor (ATR), N Reactor, and Training, Research, Isotopes, 

General Atomics (TRIGA) reactor types in Table 7-1a in OTIB-0054.

2. Determine the associated radionuclides, excluding 131I, using the ratios for estimating intakes 

for intakes based on intake activity relative to 90Sr for 10-day-old, 40-day-old, 180-day-old, and 

1-year-old reactor fuel for the ATR, N Reactor, and TRIGA reactor types in Tables 7-3a 

through 7-3c, and Tables 7-3f through 7-3i in OTIB-0054.

3. Assign the highest dose determined for the above decay times. 

5.3.2.2 Weapons Residue, 1974 to 1992

A significant amount of the work involving beta-emitting radionuclides that would have been detected 

by the gross beta procedure involved analysis of soil cores from NTS [ORAUT 2007b]. The MFAPs

from atomic weapons testing generally were not the same as those for reactor fuel.

Some results of analyses of drill-back core samples from NTS test shots are documented in Hicks 

[1981] in terms of activity per square meter. Results from five shots were analyzed to determine the 

more abundant contributors to beta exposure in the core samples. The five shots analyzed were 
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SEDAN (1962), SCHOONER (1968), SULKY (1964), CABRIOLET (1968), and PALANQUIN (1965) 

[ORAUT 2007b]. All were underground shots. The cores usually arrived within a few days, as 

opposed to months, after the shot [6]. Exposure to workers depended not only on the age of the 

samples being worked but also, presumably, from some build-up of contamination in hoods and 

ventilation systems. Therefore, exposure was probably a mix of contamination from current cores and 

older contamination from previous cores. For that reason, the analysis compared ratios for 100 days 

after each shot as being representative of the age of contamination to which the LLNL workers might 

have been exposed [ORAUT 2007b]. 

The analysis revealed that even among these five shots, all taken in the 1960s, the radionuclides of 

greatest abundance were not consistent. Some shots showed a high abundance of activation 

products and others showed mostly fission products. For the SEDAN, SCHOONER, and CABRIOLET

shots, isotopes of tungsten were most predominant, especially 185W. Iron-55 was also noticeably 

present in the SCHOONER and CABRIOLET shots. In the SULKY and PALANQUIN shots, these 

radionuclides were minor with the usual list of fission products dominating (e.g., 91Y, 89Sr, 95Nb, 95Zr, 

144Ce, 103Ru, and 106Ru). When the activity per square meter values were totaled for the five shots and 

presented as a fraction of the total activity, the fractions of the more significant radionuclides were as 

follows:

• 185W, 0.58; 

• 181W, 0.36; 

• 103Ru, 0.0070; 

• 188W, 0.0064; 

• 91Yt, 0.0051; 

• 55Fe, 0.0046; 

• 95Nb, 0.0044; 

• 89Sr, 0.0041; 

• 141Ce, 0.0036; 

• 95Zr, 0.0033; 

• 106Ru, 0.0022; 

• 7

Be, 0.0019; 

• 54Mn, 0.0017; 

• 144Ce, 0.0013; and

• 58Co, 0.0013. 

Tungsten-181 is not a beta emitter and produces little organ dose per unit activity in the organ. The 

same is true for 7

Be [ORAUT 2007c]. It is unlikely that tungsten isotopes or 7Be were source terms for 

workers outside the NTS core analysis program. For example, they were not likely sources for reactor 

workers or accelerator workers.

Ruthenium-103 is the third most abundant radionuclide in the five NTS shots, is fairly volatile, and 

would apply well to the reactor workers. Its half-life compares reasonably well to that of 185W (39 

versus 75 days, respectively) [Kocher 1981], and it produces higher dose in all organs (except 

extrathoracic region 1) per unit activity in urine than 185W [ORAUT 2007c]. It also produces higher 

dose per unit intake in all organs than 181W and 7

Be. Overall, considering abundance, biokinetics, 

dose per unit activity in urine, and universal application, 103Ru was considered a reasonably 

representative beta-emitting radionuclide for all five shots. However, some of the other major 

radionuclides concentrate in different organs. Because there was a mix of radionuclides in any one 

sample and there was a change in the mix of predominant radionuclides from core to core, it was 

considered important not to assume that the gross beta urinalysis values were represented by a 

single radionuclide.
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To determine a representative mix, consideration was given to how the major radionuclides 

concentrate in organs. For instance, barium, beryllium, strontium, and niobium are bone seekers; 

cerium concentrates mostly in the liver and to some extent bone; yttrium deposits primarily in bone 

and to some extent the liver; zirconium deposits half to bone and half to all other tissues; ruthenium 

and iron are mostly distributed evenly to all tissues. Based on all the above considerations, it is 

recommended that the intake be estimated as consisting of 50% type F 103Ru, 25% type S 141Ce, and 

25% type F 89Sr. These values should be used for 1974 through 1992. In terms of activity in urine at 

equilibrium from chronic exposure, 59% is from 103Ru and 41% is from 89Sr. The 141Ce contribution 

would be negligible. Beta yield per disintegration from both 103Ru and 89Sr is 100%, so no adjustment 

is necessary for beta yield. Additional intakes of type S 141Ce and type F 89Sr should be included at 

50% each of the 103Ru intake. Tungsten-181 is not a beta emitter, so it would not have been detected 

in the gross beta urinalysis; however, the additional dose to each organ from using 103Ru accounts for 

the unmeasured intake (or dose) from 181W. Therefore, the following method is used to assign internal 

dose from gross beta results from weapons residue:

1. Multiply the gross beta results by 0.59 to obtain the 103Ru contribution.

2. Use the adjusted results to calculate a type F 103Ru intake.

3. Assign additional intakes relative to the 103Ru intake:

a. Type S 141Ce intake = 0.5 × 103Ru intake, and

b. Type F 89Sr intake = 0.5 × 103Ru intake.

5.3.2.3 Aged Fission Products and Research Radionuclides, 1992 to 2001

If specific beta bioassay results are available (i.e., 90Sr, 32P, 137Cs, or 14C), the dose reconstructor 

should use that information for determining the internal dose.

If no specific beta bioassay results are available, intakes should be modeled as type F 90Sr using 55% 

of the gross beta result. Additional intakes of 137Cs, 32P, 14C, and 90Y should be included, each equal 

to the 90Sr intakes. This was based on the following methodology. For 1992 and later, the 

radionuclides in Attachment A, which apply to 1992 to 2001, were qualitatively used. Of the 

radionuclides in Attachment A that would have been measured by the gross beta procedure, 90Sr and 

137Cs were considered representative. They tended to be among the more significant radionuclides in 

many of the beta-exposure facilities, and one is a bone seeker while the other is a WB seeker. In 

terms of activity in urine at equilibrium from chronic exposure, 38% is from 90Sr and 62% is from 137Cs.

However, beta yield per disintegration from 90Sr is 200% because the 90Y beta would have been 

counted as well. The beta yield from 137Cs is 100%, so 90Sr would have constituted 55% of the total 

beta counts and 137Cs would have constituted 45% of the total beta counts. Therefore, the gross beta 

results should be assessed based on 55% type F 90Sr. The 137Cs intake is assumed to be equal to the 

90Sr intake to be favorable to the claimant. In addition, the listing in Attachment A for the waste 

operations buildings (513, 514, 514 tank farms, 613, and 625) showed that 32P and 14C also 

constituted significant fractions of the total activity. To account for these sources, the activity intake of 

32P and 14C is assumed to be equal to the 90Sr intake. Yttrium-90 was undoubtedly present with any 

intake of 90Sr, and it is not known if it was accounted for as part of the reported activity of 90Sr.

Therefore, an intake of 90Y equal to the intake of 90Sr should be included. The following method 

should be used to assign internal dose from gross beta results from aged fission products and 

research-related radionuclides:

1. Determine the type F 90Sr intake assuming 55% of the gross beta results.

2. Type F or M 32P intake = 90Sr intake.

3. Type F, M, or S 14C intake = 90Sr intake.

4. Type F 137Cs intake = 90Sr intake.

5. Type F 90Y intake = 90Sr intake.
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5.3.2.4 Analysis of Tritium Measurements

External dose records for LLNL might contain tritium doses without actual tritium bioassay records for 

a worker from 1958 through 1961. The following method should be used when assessing internal 

tritium dose for a worker.

1. If there are reported positive or zero tritium doses in the external dosimetry records from 1958 

through 1961, but no bioassay results for the worker, the following should be done:

a. If all tritium doses are zero, do not report any tritium dose for these time periods, and 

assign environmental internal tritium dose.

b. If any of the tritium doses are positive, report the positive doses for these periods in the 

output from the Interactive RadioEpidemiological Program (IREP). Assign the positive 

tritium doses as a lognormal distribution with a geometric standard deviation (GSD) of 

3.0.

2. If there are tritium bioassay results for the worker:

a. Assign tritium dose for years for which there is tritium bioassay. For remaining years, 

assign only environmental internal tritium dose.

b. If there is no tritium bioassay, excluding (1.b) above, assign environmental internal 

tritium dose.

5.3.3 Measurement Types and Detection Levels

A variety of program documents report detection levels for the various in vivo and in vitro bioassay 

methods. These methods are listed in Tables 5-8 and 5-9, and the documents are cited in the table 

footnotes.

Table 5-8. Bioassay in vitro detection levels.

a,b

Measurement

type Analysis method

Selected

radionuclidec Period MDA

Fecal Radiometric alpha 

spectrometry

Pu-239, 238 1977–1980 1 nCi/sample

Fecal Coprecipitation and 

proportional counting

Pu-239, 238 1990–1999 0.04 dpm

Urine Coprecipitation, separation

(gross alpha method)

Am 1957–1989 0.718 dpm/sample

Urine Coprecipitation, separation

(gross alpha method)

Am 1990–1999 0.365 dpm/sample

Urine 24-hr sample, alpha spec Am-241 2000–present 0.01 dpm

Urine Liquid scintillation C-14 2000–present 0.002 μCi/L

Urine Electrodeposition and 

proportional counting

Cm-244 1960–1999 0.2 dpm

Urine 24-hr sample, alpha spec Cm-244 2000–present 0.01 dpm

Urine Liquid scintillation HTO 1955–1978 1 μCi/L

Urine Liquid scintillation HTO 1979–present 0.01 μCi/L

Urine Liquid scintillation I-125, I-131, S-35 2000–present 0.002 μCi/L

Urine 24-hr sample, alpha spec Np-237 2000–present 0.01 dpm

Urine Liquid scintillation P-32, P-33 2000–present 0.002 μCi/L

Urine Extraction and gross 

alphad

Po-210 1955–1978 0.1 pCi/L

Urine 24-hr sample, alpha spec Po-210 2000–present 0.01 dpm
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Measurement

type Analysis method

Selected

radionuclidec Period MDA

Urine Coprecipitation Gross alpha 1957–1989 0.78 dpm/sample

Urine Coprecipitation Gross alpha 1990–1999 0.365 dpm/sample

Urine Coprecipitation, 

separation, and 

proportional counting

Pu-total alpha 1957–1989 0.051 dpm/sample

Urine Coprecipitation, 

separation, and 

proportional counting

Pu-total alpha 1990–1999 0.012 dpm/samplee

Urine 24-hr sample, alpha spec Pu 2000–present 0.01 dpm

Urine Gross beta MFAP 1957–1989 600 dpm/Lf

Urine Gross beta MFAP 1990–1999 32 dpm/L

Urine TRU-spec column with 

alpha spectrometry

Th-232, Th-228 1989–present 0.2 dpm/sample (with 

Th-232 being 

equivalent to Th-228)

Urine Fluorometric DU, NU, U AVLIS 1958–1973 5 μg/L

Urine Fluorometric DU, NU, U AVLIS 1974–1978 1 μg/L

Urine Fluorometric DU, NU, U AVLIS 1979–1989 0.3 μg/L

Urine Spot sample KPA DU, NU, U AVLIS 1990–1999 0.05 μg/L

Urine Spot sample ICP-MS U 2000–present 0.002 μg/L

Urine Extraction/alpha 

proportional counting 

(U-234 alphas measured)d

HEU 1955–1970 100 dpm/Lg

Urine Extraction/alpha 

proportional countingd

HEU 1971–1976 15 dpm/L

Urine Extraction/alpha 

proportional countingd

HEU 1977–1986 4 pCi/L

Urine Alpha spectrometryh HEU 1987–1999 0.3 pCi/L

a. Sources: Mansfield and Wood-Zika [2000], Dupzyk and Biggs [1960], LRL [1964, 1970], Griffith [1980], ORAUT [2005c], 

LRL [1965, pp. 18–23], ORAUT [2007a, 2013], LLL [ca. 1978], Dupzyk [1979], Simpson [1990], Simpson [1989], 

Explanatory notes [2006]. 

b. ICP-MS = inductively coupled plasma–mass spectrometry; KPA = kinetic phosphorescence analysis.

c. Other nuclides might have been detected by the method; these might be reported in worker files.

d. Based on information from Los Alamos National Laboratory (LANL).

e. MDA range of 0.012 to 0.051 dpm/sample was indicated. Dose reconstructors should use available sample-specific 

information.

f. Gross beta method MDA was 3.84 cpm/sample with a detector efficiency range of 20% to 31% and sample recovery of 

60%. This results in an MDA range of approximately 400 to 600 dpm/sample depending on detector efficiency [Dupzyk

1979]. Dose reconstructors should use available sample-specific information.

g. Based on information from LANL. Decision level of 50 dpm/L reported. Assume the MDA is 2 times the decision level.

h. Based on ICRP [1988] information.

Table 5-9. Bioassay in vivo detection levels.

a,b

Measurement

type Analysis method Radionuclide Period MDA

Lung Phoswichc Pu-239, Am-241 1966–1969 400 nCi, 2 nCi

Lung Phoswichd Pu-239, Am-241 1970–1979 54 nCi, 0.5 nCi

Lung Phoswich Pu-239 Circa 1978 19 nCi

Lung Phoswich Am-241 Circa 1979 0.3 nCi

Lung Phoswich Am-241 1984–1985 0.1–0.3 nCi (2.3- to 

5.25-cm CWT)

Lung Dual phoswich Pu-239 1980–present >16.3 nCi

(2.1 cm CWT)

Lung Dual phoswich “Pure” Pu-239 1984–1985 9–200 nCi (1.45-to 

5.95-cm CWT)

Lung HPGe arrays “Pure” Pu-239 2000–present 200 nCi
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Measurement

type Analysis method Radionuclide Period MDA

Lung HPGe arrays “Pure” Pu-238 1984–1985 4–100 nCi (1.5- to 4.1-cm 

CWT)

Lung ACTII “Pure” Pu-238 1999 433 nCi (3.33 cm CWT)

Lung HPGe arrays “Pure” Pu-238 2000–present 150 nCi

Lung ACTII “Pure” Pu-238 2006 87 nCi (3.6 cm CWT)

Lung ACTII Weapons-grade Pu 1984–1985 1–4 nCi (1.45- to 5.85-cm 

CWT)

Lung HPGe arrays (Am-241) Weapons-grade Pu 2000–present 0.15 nCi

Lung ACTII Am-241 1999 0.14 nCi (3.33 cm CWT)

Lung HPGe arrays Am-241 2000–present 0.15 nCi

Lung ACTII Am-241 2006 0.11 nCi (3.6 cm CWT)

Lung ACTII DU, NU Circa 1977 5 nCi

Lung ACTII DU 1984–1985 0.5–2 nCi

Lung ACTII NU 1984–1985 0.7–2 nCi

Lung ACTII U-235 1984–1985 ~100 μg (depending 

upon enrichment)

Lung ACTII U-238 (Th-234) 1999 1.27 nCi (3.33 cm CWT)

Lung HPGe arrays DU, NU 2000–present 1 nCi

Lung ACTII U-238 (Th-234) 2006 0.68 nCi (3.6 cm CWT)

Lung ACTII U-235 1999 0.07 nCi (3.33 cm CWT)

Lung HPGe arrays AVLIS U, Th, U2352000–present 1 nCi

Lung ACTII U-235 2006 0.05 nCi (3.6 cm CWT)

Lung HPGe arrays 40% U, 95% U 2000–present 0.1 nCi

Lung HPGe arrays Cm-244 2000–present 100 nCi

Lung HPGe arrays Np-237 2000–present 0.35 nCi

Lung ACTII Np-237 2006 0.3 nCi (3.6 cm CWT)

Lung HPGe arrays Pa-233 2000–present 1 nCi

Lung ACTII Pa-233 2006 0.13 nCi (3.6 cm CWT)

Thyroid count ACTII I-131 1984–1985 1 nCi

Thyroid count ACTII I-131 2000–present 0.01 nCi

Thyroid count ACTII I-131 2006 0.04 nCi

Thyroid count Portable NaI I-131 2006 0.27 nCi

Thyroid count Portable NaI I-125 2000–present 0.02 nCi

Thyroid count ACTII I-125 2006 0.31 nCi

Thyroid count Portable NaI I-125 2006 0.02 nCi

WB scan Shadow shield counter K-40 1964–1995 150 g K in 20 min 

(~120 nCi K-40)

WB scan Shadow shield counter 150- to 300-keV 

gamma emitters

1964–present 500 gamma/s between 

150 and 300 keV

WB scan Shadow shield counter Th-232 1970–present 1 nCi (based on Tl-208 

with no correction for 

branching ratio)

WB scan Shadow shield counter Natural Th (Tl-208) 2000–present 1 nCi

WB scan Shadow shield counter I-131 (TB) 1970–1995 1 nCi

WB scan Shadow shield counter Cs-137 1970–1995 1 nCi

WB scan Shadow shield counter Co-60 1970–1995 1 nCi

WB scan Shadow shield counter Ce-144 1970–present 8 nCi

WB scan Shadow shield counter Ce-141 1970–present 2 nCi

WB scan Shadow shield counter Ru-103 1970–present 3 nCi

WB scan Shadow shield counter Ru-106 1970–1995 5 nCi

WB scan Shadow shield counter Zr-95 1970–present 1 nCi

WB scan Shadow shield counter Mn-54 1970–1995 1 nCi

WB scan Shadow shield counter Zn-65 1970–present 1 nCi
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Measurement

type Analysis method Radionuclide Period MDA

WB scan Shadow shield counter Na-22 1970–1995 1 nCi

WB scan Shadow shield counter K-40 1996–1999 4.46 nCi

WB scan Shadow shield counter Mn-54 1996–1999 0.63 nCi

WB scan Shadow shield counter Fe-59 1996–1999 0.98 nCi

WB scan Shadow shield counter Co-60 1996–1999 0.66 nCi

WB scan Shadow shield counter Ru-106 1996–1999 6.53 nCi

WB scan Shadow shield counter I-131 1996–1999 1 nCi

WB scan Shadow shield counter Cs-134 1996–1999 0.7 nCi

WB scan Shadow shield counter Cs-137 1996–1999 0.89 nCi

WB scan Shadow shield counter Co-57 2000–2005 0.97 nCi

WB scan Shadow shield counter Co-60 2000–2005 0.51 nCi

WB scan Shadow shield counter Cs-137 2000–2005 0.7 nCi

WB scan Shadow shield counter I-131 2000–2005 0.66 nCi

WB scan Shadow shield counter K-40 2000–2005 2.6 nCi

WB scan Shadow shield counter Na-22 2000–2005 0.44 nCi

WB scan Shadow shield counter K-40 2006 2.6 nCi

WB scan Shadow shield counter Cs-137 2006 0.72 nCi

WB scan Shadow shield counter Co-60 2006 0.43 nCi

WB scan Shadow shield counter I-131 2006 0.67 nCi

WB scan Shadow shield counter Na-22 2006 0.42 nCi

WB scan Shadow shield counter Pa-233 2006 1.5 nCi

WB scan Shadow shield counter Ac-228 2007–present 2.74 nCi

WB scan Shadow shield counter Ba-133 2007–present 0.85 nCi

WB scan Shadow shield counter Bi-207 2007–present 0.47 nCi

WB scan Shadow shield counter Bi-212 2007–present 7.71 nCi

WB scan Shadow shield counter Bi-214 2007–present 1.82 nCi

WB scan Shadow shield counter Cf-249 2007–present 0.83 nCi

WB scan Shadow shield counter Cm-243 2007–present 3.59 nCi

WB scan Shadow shield counter Co-57 2007–present 0.97 nCi

WB scan Shadow shield counter Co-60 2007–present 0.37 nCi

WB scan Shadow shield counter Cs-134 2007–present 0.56 nCi

WB scan Shadow shield counter Cs-137 2007–present 0.7 nCi

WB scan Shadow shield counter Eu-152 2007–present 1.75 nCi

WB scan Shadow shield counter Eu-154 2007–present 1.25 nCi

WB scan Shadow shield counter Eu-155 2007–present 2.58 nCi

WB scan Shadow shield counter I-131 2007–present 0.68 nCi

WB scan Shadow shield counter K-40 2007–present 2.52 nCi

WB scan Shadow shield counter Na-22 2007–present 0.4 nCi

WB scan Shadow shield counter Np-237 2007–present 6.38 nCi

WB scan Shadow shield counter Np-239 2007–present 4.73 nCi

WB scan Shadow shield counter Pa-231 2007–present 21.83 nCi

WB scan Shadow shield counter Pa-233 2007–present 1.6 nCi

WB scan Shadow shield counter Pa-234 2007–present 3.93 nCi

WB scan Shadow shield counter Pb-212 2007–present 2.23 nCi

WB scan Shadow shield counter Pb-214 2007–present 2.36 nCi

WB scan Shadow shield counter Po-210 2007–present 3.40E+04 nCi

WB scan Shadow shield counter Ra-224 2007–present 20.6 nCi

WB scan Shadow shield counter Ra-226 2007–present 23.9 nCi

WB scan Shadow shield counter Th-231 2007–present 11.6 nCi

WB scan Shadow shield counter Th-234 2007–present 15.8 nCi
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Measurement

type Analysis method Radionuclide Period MDA

WB scan Shadow shield counter U-235 2007–present 1.61 nCi

a. Sources: Mansfield and Wood-Zika [2000], LRL [1964, 1970], LRL [1965], Griffith [1980], ORAUT [2005c, 2007a, 2013], 

LLL [ca. 1978], Anderson et al. [1979], King [1977], Whole-body and lung counter MDAs [ca. 1985], Sundsmo and 

Hickman [2009a, 2009b, 2009c], Hickman [2008]. 

b. CWT = chest wall thickness; HPGe = hyperpure germanium.

c. Based on LANL information.

d. Based on ICRP [1988] information.

In some cases, there is evidence that LLNL implemented monitoring using a particular methodology, 

but site-specific detection levels for that methodology were not available. In these cases, detection 

levels for comparable Los Alamos National Laboratory (LANL) methods were used to complete 

Tables 5-8 and 5-9 because there is evidence of interlaboratory communications on analytical 

methods over the years, and because both LLNL and LANL were operated by the same contractor, 

which increases the likelihood for shared procedures [ORAUT 2005c]. 

In a few instances, neither LLNL nor LANL detection levels were available, requiring a surrogate 

source of information. ICRP Publication 54 specified nominal detection levels for a variety of 

measurement methods that were comparable to those used at LLNL [ICRP 1988]. 

A review of in vivo bioassay records provided by LLNL indicates that the data typically include the 

assessed radionuclide, MDA, activity result, height, weight, and chest wall thickness (CWT). In most 

cases, the dose reconstructor will have all the information necessary to assess the internal dose from 

in vivo bioassay data. In the absence of specific in vivo MDA information in the worker’s dosimetry 

records, the dose reconstructor should refer to Table 5-9. The expected intake pattern in most cases 

is an acute intake. At LLNL, airborne and surface contamination was typically controlled to prevent 

intakes, so most intakes would have been the result of unexpected releases. Small intermittent 

releases that were not immediately detectable could have occurred, so an individual could have had 

multiple acute intakes.

5.3.4 Reporting Formats and Codes

There are a variety of entries in LLNL urine bioassay records. These bioassay records exist in several 

bioassay databases named MAPPER, SYMPHONY, and BLIMS used by LLNL throughout its history.

Table 5-10 provides a listing of the bioassay information reported in the outputs of these databases 

provided in claim dosimetry record files.

Table 5-10. Bioassay database record entries. 

Type of entry Entry description

Sample number Number assigned to sample by Bioassay Laboratory

Employee number 6-digit integer

Employee name Last name, comma, other names and/or initials

Nuclide analyzed MFPa, ALPHA, PU239, PU238, U238, AM241, CM244, or other nuclide

Sample date Month, day, and year

Sample result Floating point number or exponential number

Sample error One standard deviation as a percentage of the result

Sample media U = urine, B = blood, f = feces, n = nose swipe

Sample type R = routine; S = special

Result units Dpm, dpm/L, dpm/sample, μg/L, μCi/L, μCi/sample, or other unit

Sample volume or mass ml, or g

Decision level Decision level

MDA MDA

a. MFP = mixed fission product. 
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5.3.5 Assessment of Intake for Unmonitored Workers

LLNL had an established in vitro and in vivo bioassay sampling program, and there was clear 

direction to supervisors on individual employee participation in the program. In the case where a 

worker might not have been monitored but should be assigned an intake based on job description and 

information in the claimant telephone interview, a means of assessing intakes for unmonitored 

workers is necessary. There is little correlation between external dosimetry and the bioassay program.

From 1953, external dosimetry was administered as a sitewide requirement, whereas bioassay was 

determined based on the potential of exposure by facility supervisors and Health Physics. Dose 

reconstructors should refer to the co-exposure data in Attachment B to assign intakes for unmonitored 

workers. Attachment C provides guidance on assigning the 50th- or 95th-percentile co-exposure

intakes based on job category and level of contact with the radioactive material.

5.3.6 Unmonitored Intakes of Radionuclides Other than Primary Radionuclides

A short description of the work in relation to each of the listed radionuclides follows. Most were 

handled as part of basic research, and the number of exposed individuals would have been small.

Actinium-227 was used in Building 251, which housed basic research in the chemical and nuclear 

behavior of the transuranic (TRU) elements starting in 1955. Building 251 was restricted to an 

as-needed basis [LLL 1980]. Given the nature of the work, access and radiological controls, and an in 

vitro and in vivo bioassay monitoring program in place for workers in areas where radioactive material 

was routinely handled, unmonitored intakes were unlikely. Starting around mid-1978, a solution of 

227Ac stored in a glovebox in Building 251 was “milked” periodically for its 223Ra, which was used as an 

alpha-emitting contaminant in exercises [Gibson 1980]. Actinium-227 was also stored in a cave in 

Room 1235 of Building 251 [Meadows et al. 1981]. Given the containment and restrictive use controls, 

the potential for exposure to unmonitored individuals was remote. In 1992 all U.S. nuclear testing 

ceased. In 1993 LLNL decided to discontinue programmatic operations in Building 251 based on new 

regulatory requirements. Building 251 was placed in standby mode in 1995. During the standby 

period, Building 251 was staffed with three people. Building 251 transitioned from Category II 

Nonreactor Nuclear Facility in 2002 to a radiological facility in 2005 where equipment and inventory 

were removed [Gray et al. 2006]. 

Thorium-232 (natural thorium) was associated primarily with Site 300 and two other facilities, 

Buildings 231 and 321C. Building 231 is a large experimental manufacturing, assembly, test, and 

materials-handling facility, and Building 321C machined DU and thorium. Machining only occurred in 

321C. Site 300 was a high-explosive test facility that opened in mid-1955. DU, NU, and natural 

thorium were used with high-explosive materials tests at Site 300.

Curium-244 was associated primarily with Building 251 starting in 1955 when the building became 

operational. As indicated previously, access to Building 251 was restricted to an as-needed basis.

Given the nature of the work, access and radiological controls, and in vitro and in vivo bioassay 

monitoring programs in place for workers in areas where radioactive material was routinely handled, 

unmonitored intakes were unlikely. Curium-244 was also used in Building 332 (Plutonium Facility) with 

similar controls in place. Building 332 became operational in 1961.

Neptunium-237 work was conducted primarily in Building 251 and, as stated above, unmonitored 

intakes were unlikely. Research on 237Np was conducted in Room 1319 of Building 281, which 

included gloveboxes. Building 281 conducted tracer work, dissolution studies, and flow studies. As a 

result, the potential for internal exposure to 237Np in Building 281 would have been remote.

Based on the foregoing, the potential for an unmonitored intake of the above-listed radionuclides is 

limited. However, it is necessary to demonstrate that appropriate doses can be assigned when 
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 32 of 113

 

presumptive intakes, beyond those that might be characterized by bioassay records, are implied by 

case-specific information. In the absence of case-specific information, the dose from unmonitored 

intakes of LLNL radionuclides other than primary radionuclides should not be assigned. Based on the 

foregoing information on Building 251, no dose from any of the above-listed radionuclides should be 

assigned after 1995.

Table 5-11 lists the radionuclides to consider along with the work locations. Workers who are 

assigned doses from unmonitored intakes of LLNL radionuclides other than primary radionuclides 

based on the co-exposure intakes described in the table below will not be assessed for the period 

before 1957 for plutonium and americium, and before 1958 for uranium, because they are included in 

the designated SEC period.

Table 5-11. Dose assignment for radionuclides other than primary radionuclides. 

Radionuclide Building

Likely exposed 

individuals

Period of potential 

exposure

Dose estimation techniques

(in order of preference)

Ac-227 251 Chemists, 

researchers, health 

physics technicians, 

technicians, and 

maintenance

1978–1995 1. Base on bioassay results.

2. Assign dose based on coexposure intakes for Pu-239 

assuming Ac-227 types F, M, and 

S.

Np-237 251, 281 Chemists, 

researchers, health 

physics technicians, 

technicians, and 

maintenance

1955–1995 

(Building 251)

and

1952–present 

(Building 281) 

1. Base on bioassay results.

2. Assign dose based on coexposure intakes for Pu-239 

assuming type M Np-237.

Cm-244 251, 332 Chemists, 

researchers, health 

physics technicians, 

technicians, and 

maintenance

1955–1995 

(Building 251)

and

1961–present 

(Building 332)

1. Base on bioassay results.

2. Assign dose based on coexposure intakes for Am-241 

assuming type M Cm-244.

Th-232 

(natural 

thorium)

231, 

321C, 

Site 300

Chemists, 

researchers, 

machinists, health 

physics technicians, 

and technicians

1952–present 

(Buildings 231, 

321C)

and

1955–present

(Site 300)

1. Base on bioassay results.

2. Assign dose assuming a natural 

thorium mass intake equal to the 

co-exposure uranium mass intake.

Assume a natural thorium specific 

activity of 4.88E+02 dpm/mg to 

determine the total thorium activity 

intake. Assume 50% of the total 

thorium activity intake is Th-232 

and 50% of the total thorium 

activity intake is Th-228. Assume 

types M and S.

5.4 SIGNIFICANT INCIDENTS WITH INTERNAL DOSE POTENTIAL

During operations at LLNL, a number of incidents increased the potential for intakes of radioactive 

materials. If a worker recalls involvement in one or more of those incidents, dose reconstructors can 

use the information in Table 5-12 as input to an incident-specific assessment. This list is not all

encompassing, and many other incidents probably occurred; these are the incidents identified from

review of the data capture records. Individual worker records could provide documentation of 

involvement in other incidents.
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Table 5-12. Input parameters for significant incidents.

a

Incident date Incident description Facility Other information

03/26/1963 Criticality Building 110 Potential for internal and external exposure (I, 

Kr, Xe); monitored participants listed by name in 

1963 reference (see below).

11/06/1975 Radium contamination Building 343, 

Room. 1005

Crystal was removed from radium dial resulting 

in spread of contamination.

11/23/1971 Pu-contaminated squib valve Building 343 Direct and indirect bioassays performed on 14 

individuals

a. Sources: LRL [ca. 1963a, ca. 1963b], Leahy [1975], AEC [1972]. 

5.5 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction.

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database (SRDB).

[1] Szalinski, Paul A. Oak Ridge Associated Universities (ORAU) Team. Health Physicist. March 

2007.

The LLNL Technical Basis Manual for Internal Dosimetry [Mansfield and Wood-Zika 2000] 

discusses RU in the DOE complex, so the internal dosimetrists at LLNL are aware of the 

issues. 

[2] Thomas, Bill R. ORAU Team. Health Physicist. March 2007.

Results for gross alpha in urine were reviewed in the claimant files. The frequency of sample 

collection was dependent on the particular building where the worker was assigned. The 

frequency ranged from samples several times per year to a single sample per year.

[3] Szalinski, Paul A. ORAU Team. Health Physicist. March 2007.

In ORAUT [2006a], when asked if the procedures were in place from 1957 to 1996, an LLNL 

employee replied that they were, at least to 1987. Because these procedures were based on 

DOE Environmental Measurement Laboratory Procedures Manual HASL-300 [Cheico 1997], 

which did not undergo its next significant revision until 1997, it is likely that the procedures 

were unchanged through 1996.

[4] Thomas, Bill R. ORAU Team. Health Physicist. March 2007.

The information in this paragraph comes from Explanatory Notes on “Gross Alpha” Results in 

LLNL claimant files [Explanatory notes 2006] for Summary of Bioassay Results that were 

reported after approximately 2002.

[5] Szalinski, Paul A. ORAU Team. Health Physicist. March 2007.

Because the first two gross alpha analysis methods might contain the same radionuclides and 

there is no information as to which method was used on a given sample, the steps are given to 

assess the various possibilities that might be encountered.

[6] Bihl, Donald E. ORAU Team. Principal Health Physicist. April 2007.

Personal communication with M. H. Chew, October 16, 2006.

[7] LaBone, Thomas R. ORAU Team. Deputy Principal Internal Dosimetry Scientist. 

February 2007.

The uranium results were recorded as μg/L and the statistical analyses were performed in 

those units. However, the Integrated Modules for Bioassay Analysis (IMBA) computer program 
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requires that all excreta data be entered as total excretion per day; hence, the statistical 

parameters were converted to excretion per day before intake calculations were made using 

IMBA.

[8] Brackett, Elizabeth M. ORAU Team. Principal Internal Dosimetry Scientist. February 2007.

The uniform absolute error of 1 weights all results equally; other fitting schemes weight high 

values or low values disproportionately. Because the median and 84th-percentile values were 

determined from statistical analysis of many samples in each interval, there was no a priori

reason to weight results from one interval over another.

[9] LaBone, Thomas R. ORAU Team. Deputy Principal Internal Dosimetry Scientist. 

February 2007.

The breathing rate and particle size distribution are project default values to be used unless 

site-specific information indicates otherwise. No information has been found concerning 

intakes at LLNL that shows that the default values should not be used. See, for instance, 

OCAS-IG-002, Internal Dose Reconstruction Implementation Guide [NIOSH 2002], and ICRP 

Publication 66, Human Respiratory Tract Model for Radiological Protection [ICRP 1994]. 

[10] Bihl, Donald E. ORAU Team. Principal Health Physicist. February 2007.

The minimum GSD of 3 is established in ORAUT-OTIB-0060, Internal Dose Reconstruction

[ORAUT 2018]. It reflects the overall uncertainty associated with biokinetic modeling as well as 

usual radiochemical analysis, and indicates that even though the spread in co-exposure

excreta results for a given population (e.g., a year of excreta samples) can have a GSD of <3, 

the uncertainty of intakes determined using the biokinetic models is no less than 3.

[11] LaBone, Thomas R. ORAU Team. Deputy Principal Internal Dosimetry Scientist. 

February 2007.

Use of the midpoint of the period represented by the excreta data point is established in 

ORAUT-OTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment

[ORAUT 2005a], and is standard practice for assigning chronic intakes or acute intakes with 

unknown dates. See, for instance, ANSI/HPS N13.39-2001, Design of Internal Dosimetry 

Programs [ANSI/HPS 2001]. 
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 35 of 113

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GLOSSARY

absorption type

Categories for materials according to their rates of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively insoluble in the 

respiratory tract (slow solubilization).

activity

Amount of radioactivity. The International System unit of activity is the becquerel 

(1 disintegration per second); the traditional unit is the curie [37 billion (3.7 × 1010) becquerels].

activity median aerodynamic diameter (AMAD)

Diameter of a unit density sphere with the same terminal settling velocity in air as that of the 

aerosol particle whose activity is the median for the entire aerosol.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

atomic vapor laser isotope separation (AVLIS)

Process for enriching 235U by ionizing uranium compounds in a vapor based on a laser tuned 

to the electronic structure of uranium compounds.

bioassay

Measurement of amount or concentration of radionuclide material in the body (in vivo 

measurement) or in biological material excreted or removed from the body (in vitro 

measurement) and analyzed for purposes of estimating the quantity of radioactive material in 

the body. Also called radiobioassay. 

chronic exposure

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years).

claim

Petition filed by a covered employee or survivor(s) thereof and submitted to the 

U.S. Department of Labor for compensation under the Energy Employees Occupational Illness 

Compensation Program Act of 2000. There can be multiple claims filed per worker case.

claimant

Individual who has filed for compensation under the Energy Employees Occupational Illness 

Compensation Program Act of 2000. This individual can be the energy employee (worker), a 

survivor, or the legal representative of the energy employee.

contamination

Radioactive material in undesired locations including air, soil, buildings, animals, and persons.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.
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depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of rad, rep, or gray.

Energy Employees Occupational Illness Compensation Program Act of 2000, as amended 

(EEOICPA; 42 USC 7384 et seq.)

Law that provides for evaluation of cause and potential compensation for energy employees 

who have certain types of cancer or other diseases and conditions.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens. In this document, 

exposure does not refer to the radiological physics concept of charge liberated per unit mass 

of air.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

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.

half-life

Time in which half of a given quantity of a particular radionuclide disintegrates (decays) into 

another nuclear form. During one half-life, the number of atoms of a particular radionuclide 

decreases by one half. Each radionuclide has a unique half-life ranging from millionths of a 

second to billions of years.

highly enriched uranium (HEU)

Uranium enriched to at least 20% 235U for use as fissile material in nuclear weapons 

components and some reactor fuels. Also called high-enriched uranium.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds.

Integrated Modules for Bioassay Analysis (IMBA)

Computer program that uses bioassay results and other information to calculate intakes of 

radionuclides and subsequent doses. The main tool for internal dose assessment.

internal dose

Dose to the organ of interest from sources inside the body (e.g., plutonium or uranium) that 

are inhaled, ingested, absorbed, or injected through a wound).
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in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay

Measurements of radioactive material in the human body using instrumentation that detects 

radiation emitted from the radioactive material in the body.

isotope

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

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

Isotopes have very nearly the same chemical properties.

kiloelectron-volt (keV)

Unit of particle energy equal to 1,000 (1 × 103

) electron-volts.

megaelectron-volt (MeV)

Unit of particle energy equal to 1 million (1 × 106

) electron-volts.

minimum detectable amount (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

monitoring (personnel)

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.

National Institute for Occupational Safety and Health (NIOSH)

U.S. agency responsible for dose reconstruction under the Energy Employees Occupational 

Illness Compensation Program Act of 2000. Part of the Centers for Disease Control and 

Prevention, which is part of the U.S. Department of Health and Human Services, NIOSH is the 

federal agency responsible for conducting research and making recommendations for the 

prevention of work-related injury and illness.

natural uranium (NU)

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries 

per gram).

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.
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routine monitoring

Monitoring carried out at regular intervals during normal operations.

Special Exposure Cohort (SEC) [42 USC 7384l(14)]

[A] “member of the Special Exposure Cohort” means a Department of Energy employee, 

Department of Energy contractor employee, or atomic weapons employee who meets any of 

the following requirements:

(A) The employee was so employed for a number of work days aggregating at least 250 

work days before February 1, 1992, at a gaseous diffusion plant located in Paducah, 

Kentucky, Portsmouth, Ohio, or Oak Ridge, Tennessee, and, during such 

employment— 

(i) was monitored through the use of dosimetry badges for exposure at the plant of 

the external parts of employee’s body to radiation; or

(ii) worked in a job that had exposures comparable to a job that is or was monitored 

through the use of dosimetry badges.

(B) The employee was so employed before January 1, 1974, by the Department of 

Energy or a Department of Energy contractor or subcontractor on Amchitka Island, 

Alaska, and was exposed to ionizing radiation in the performance of duty related to 

the Long Shot, Milrow, or Cannikin underground nuclear tests.

(C) (i) Subject to clause (ii), the employee is an individual designated as a member of 

the Special Exposure Cohort by the President for purposes of the compensation 

program under section 7384q of this title.

(ii) A designation under clause (i) shall, unless Congress otherwise provides, take 

effect on the date that is 180 days after the date on which the President submits to 

Congress a report identifying the individuals covered by the designation and 

describing the criteria used in designating those individuals.

spot sample

In relation to bioassay, a single void of urine.

whole-body counter (WBC)

Equipment used to perform in vivo bioassay. Radiation emitted from radioactive material 

deposited throughout the body is measured.
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ATTACHMENT A 

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001

Table A-1. Environmental releases by building, 1993 to 2001.

a

Current

building no. Building activity

Radionuclides and

activity fractions

131 Mechanical and Electrical Divisions Th-232 2.29E–10

U-234 8.47E–02

U-235 1.18E–02

U-238 9.04E–01

132 Analytical and Nuclear Chemistry Laboratories,

Forensic Sciences Center

H-3 7.3E–09

Co-60 1.5E–07

Th-228 1.2E–10

Th-230 3.5E–10

Th-232 7.9E–05

U-234 3.1E–04

U-235 7.2E–03

U-238 9.9E–01

Pu-238 4.4E–03

Pu-239 6.7E–06

Pu-240 1.7E–06

Pu-241 1.6E–05

Pu-242 2.0E–10

Am-241 3.1E–07

151 Isotope Sciences,

Environmental Services Laboratory

Ni-63 1.1E–01

Zn-65 5.1E–07

Sr-90 3.8E–05

Y-90 3.7E–06

Tc-99 1.3E–10

Tc-99m 5.3E–08

Ru-106 4.6E–09

Sn-113 1.4E–07

Sb-125 1.6E–06

I-129 4.6E–10

Ba-133 5.1E–08

Cs-134 5.1E–08

Cs-136 7.1E–07

Cs-137 9.5E–05

Ce-144 8.0E–07

Pm-147 7.6E–05

Eu-152 6.5E–08 

Eu-154 6.5E–08

Eu-155 1.6E–12

Bi-207 4.2E–05

Po-209 1.9E–12

Ra-226 2.5E–08

Pa-231 2.7E–07

Th-228 1.4E–09

Th-229 2.2E–12

Th-230 1.5E–09

Th-232 6.4E–08

U-232 2.7E–11

U-233 1.3E–07

U-234 2.0E–03

U-235 3.0E–04

U-236 7.3E–06

U-238 1.9E–02

Np-237 8.9E–04
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

151 (continued) Isotope Sciences,

Services Laboratory

Pu-236 5.3E–06

Pu-238 2.0E–02

Pu-239 1.7E–01

Pu-240 3.8E–02

Pu-241 6.1E–01

Pu-242 9.5E–05

Pu-244 5.1E–05

Am-241 1.0E–02

Am-243 7.6E–04

Cm-242 5.1E–06

Cm-244 1.5E–05

Cm-246 1.2E–05

Cm-248 8.0E–06

Cf-249 1.5E–04

175 U-AVLIS U-234 3.11E–01

U-235 9.50E–03

U-238 6.79E–01

177 U-AVLIS U-234 3.05E–01

U-235 1.04E–02

U-238 6.85E–01

179 Unknown U-234 4.38E–01

U-235 1.24E–01

U-238 4.38E–01

212 Physics and Space (RTNS) H-3 1

222 Unknown H-3 6.96E–01

C-14 5.44E–10

Ni-63 3.04E–01

Th-232 6.86E–08

U-234 2.84E–08

U-235 1.25E–09

U-238 7.13E–06

Pu-239 6.53E–08

223 Unknown Pu-238 1.09E–02

Pu-239 8.78E–01

Am-241 1.25E–04

Am-243 1.11E–01

226 Unknown H-3 1.00E+00

U-238 9.67E–06

227 Unknown U-234 5.05E–01

U-235 2.26E–02

U-238 4.73E–01

231 Safeguards and engineering Th-232 1.3E–02

U-234 8.5E–02

U-235 1.2E–02

U-238 8.9E–01
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

235 Characterization studies and ion beam experiments Th-232 4.0E–08

U-234 2.7E–01

U-235 1.6E–02

U-238 6.7E–01

Pu-238 3.0E–04

Pu-239 1.1E–02

Pu-240 2.5E–03

Pu-241 2.9E–02

235 Characterization studies and ion beam experiments Pu-242 1.7E–07

Am-241 1.4E–03

241 R&D C-14 2.1E–07

P-32 8.7E–08

Th-232 1.9E–12

U-234 1.0E+00

U-235 1.8E–06

U-238 4.0E–05

251 Heavy element facility U-233 6.58E–09

Pu-238 2.17E–04

Pu-239 5.54E–09

Pu-243 1.45E–06

Am-241 3.29E–01

Cm-243 1.69E–01

Cm-244 5.02E–01

Cm-248 4.18E–05

Cf-252 1.37E–05

251 Laboratories and counting rooms H-3 1.1E–04

C-14 1.0E–07

P-32 1.0E–06

Sr-90 5.5E–06

Sr-90/Y-90 2.8E–08

Y-90 3.7E–08

Cs-137 1.4E–07

Bi-214 3.3E–01

Pb-214 3.3E–01

Po-218 3.3E–01

Ra-226 1.5E–02

Th-230 2.0E–08

U-234 3.7E–04

U-235 1.6E–05

U-238 3.5E–04

Np-237 4.3E–08

Pu-238 2.6E–06

Pu-239 6.6E–05

Pu-240 1.5E–05

253 Unknown Pu-241 6.5E–04

Pu-242 9.9E–10

Am-241 3.1E–06
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 49 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

254 Bioassays and analytical services H-3 4.4E–02

C-14 1.4E–02

P-32 4.0E–03

P-33 4.0E–01

S-35 4.1E–02

Sr-90 9.9E–02

Y-90 9.9E–02

I-125 6.3E–02

Po-209 2.6E–05

Th-230 6.0E–03

U-232 2.3E–04

U-233 1.0E–05

U-234 4.1E–04

U-235 1.0E–05

U-236 3.2E–03

U-238 8.9E–07

Pu-238 0.0E+00

Pu-239 9.0E–05

Pu-240 0.0E+00

Pu-241 0.0E+00

Pu-242 5.5E–04

Np-237 4.4E–02

Np-239 1.7E–01

Am-241 3.2E–05

Am-243 9.2E–04

Cm-242 2.5E–04

Cm-244 3.3E–04

Cf-249 2.5E–04

Cf-252 5.4E–03

255 Calibration laboratory H-3 1.00E+00

C-14 3.50E–10

P-32 2.80E–08

S-35 1.20E–09

Sr-90 4.89E–13

Y-90 4.89E–13

I-125 5.91E–08

I-131 1.80E–07

Th-230 9.53E–11

Th-232 2.50E–15

U-233 2.50E–10

U-234 1.10E–11

U-235 2.90E–13

U-236 8.69E–09

U-238 4.19E–14

Pu-239 2.77E–12

Pu-242 1.40E–11

Np-237 1.70E–09

Np-239 1.70E–09

Am-241 9.54E–13

Am-243 2.54E–11

Cm-242 9.09E–12

Cm-244 9.49E–13

Cf-252 8.39E–11
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 50 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

281 Tracer and dissolution studies H-3 3.7E–02

Be-10 1.5E–02

C-14 2.8E–01

Na-22 1.2E–04

Cl-36 1.5E–02

Ca-41 1.5E–01

Mn-54 4.9E–06

Ni-59 1.0E–04

Ni-63 3.1E–01

Co-60 5.0E–06

Sr-90 1.5E–02

Tc-99 1.5E–04

Sb-125 4.9E–08

Eu-152/Tm-171 1.7E–08

Eu-154/Eu-155 2.1E–05

Th-232 1.1E–06

U-233 1.9E–04

U-234 1.9E–04

U-235 8.1E–06

U-238 1.8E–04

U-233/U-238 1.9E–06

Pu-239 6.3E–04

Pu-242 2.0E–03

Pu-244 2.4E–06

Np-237 1.7E–01

Am-241 5.4E–04

Am-241/Np-237 1.4E–04

282 Residual contamination H-3 1.0E+00

Rb-86/87 7.5E–09

292 Residual contamination (RTNS) H-3 1

298 Laser fusion program H-3 9.9E–01

U-234 4.1E–04

U-235 1.9E–05

U-238 6.0E–03

321 Milling and shaping DU, Thoriumb Unknown

321A Milling and shaping U-234 1.8E–01

U-235 2.4E–02

U-238 7.9E–01

321C Milling, machining, and shaping U-234 6.5E–01

U-235 8.1E–03

U-238 3.4E–01

322 Mechanical and Electrical U-234 8.1E–01

U-235 1.2E–02

U-238 1.8E–01

327 Mechanical and Electrical U-234 7.7E–02

U-235 8.8E–02

U-238 8.4E–01

331 Research and laboratories H-3 (HTO) 4.69E–01

H-3 (HT) 5.19E–01

H-3 1.24E–02

332 Gloveboxes, HEPA filters Pu-239 TRU Unknown
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 51 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

341 Lasers Directorate U-234 5.70E–02

U-235 7.69E–03

U-238 9.35E–01

361 R&D H-3 5.6E–06

C-14 3.3E–03

P-32 8.3E–01

P-33 1.1E–01

S-35 5.4E–02

362 R&D H-3 7.22E–01

C-14 2.78E–01

363 R&D H-3 9.99E–01

C-14 1.11E–07

P-32 1.42E–03

364 R&D H-3 1.11E–03

C-14 2.76E–03

P-32 9.96E–01

365 R&D H-3 4.04E–03

C-14 9.96E–01

366 R&D H-3 5.00E–01

P-32 4.38E–01

P-33 6.25E–02

377 R&D H-3 7.66E–04

C-14 3.83E–03

P-32 9.95E–01

Ni-63 1.15E–04

378 Unknown Co-57 1.44E–03

Co-60 5.59E–02

Sr-85 6.52E–02

Cd-109 3.58E–02

Cs-134 6.98E–01

Cs-137 6.98E–02

U-233 2.14E–04

U-234 8.66E–07

U-235 4.84E–08

U-238 2.00E–06

378 Unknown Pu-236 2.13E–03

Pu-239 6.99E–05

Pu-240 5.59E–05

Pu-242 1.05E–02

Pu-244 3.58E–04

Np-237 1.44E–06

Am-241 6.00E–02

Am-243 4.01E–04

381 Unknown H-3 1

391 Unknown H-3 1

412W Unknown Ni-59 3.50E–07

Ni-63 1.00E+00

446 Unknown C-14 1
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 52 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

513 Waste processing H-3 9.85E–01

C-14 2.49E–04

P-32 3.17E–06

K-40 5.79E–04

Mn-54 8.34E–06

Co-57 7.28E–07

Co-60 1.03E–04

Sr-90 7.27E–04

Nb-95 8.34E–06

Zr-95 1.30E–05

Ru-106 6.07E–09

I-125 7.49E–07

I-131 2.70E–08

Ba-133 6.09E–07

Cs-134 4.13E–06

Cs-137 2.61E–05

Cs-138 1.52E–10

Ce-141 6.07E–09

Ce-144 9.56E–06

Eu-152 1.67E–04

Eu-154 2.12E–07

Eu-155 3.79E–08

Tl-208 6.22E–08

Bi-212 8.50E–08

Bi-214 5.92E–09

Pb-210 7.89E–06

Pb-212 1.43E–07

Pb-214 2.12E–08

Ra-223 4.55E–07

Ra-226 8.65E–08

Ra-228 4.55E–16

Ac-228 1.52E–09

Pa-231 8.50E–07

Th-226 9.71E–07

Th-227 1.67E–06

Th-228 4.04E–05

Th-232 3.51E–06

Th-234 3.03E–06

U-233 1.52E–05

U-234 1.08E–03

U-235 1.52E–04

U-238 1.04E–02

Pu-238 1.74E–09

Pu-239 2.28E–04

Pu-240 2.03E–08

Pu-241 1.27E–06

Pu-242 6.87E–07

Am-241 7.62E–04

Cm-244 3.03E–05

Cf-249 2.28E–07
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 53 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

514 Waste processing H-3 8.61E–01

Be-7 7.53E–06

C-14 2.40E–02

Na-22 9.05E–05

P-32 6.53E–02

S-35 8.42E–03

K-40 6.11E–05

Sc-46 5.46E–06

Cr-51 5.59E–06

Fe-55 5.42E–06

Mn-54 1.66E–05

Co-56 1.82E–07

Co-57 7.56E–05

Co-58 5.46E–06

Co-60 1.09E–04

Ni-63 1.45E–07

Zn-65 1.92E–08

Y-88 5.33E–05

Sr-89 4.96E–11

Sr-90 9.13E–04

Nb-94 1.19E–06

Nb-95 8.46E–06

Zr-95 1.20E–07

Tc-99 6.05E–05

Ru-103 2.18E–09

Ru-106 1.33E–06

Cd-109 2.89E–08

Sb-125 2.08E–05

I-125 1.39E–04

I-131 3.63E–06

Ba-133 1.02E–04

Cs-134 6.37E–05

Cs-137 1.16E–03

Ce-139 7.14E–13

Ce-141 1.82E–09

Ce-144 4.87E–03

Gd-148 6.04E–05

Pm-147 3.07E–06

Sm-151 2.45E–07

Eu-152 2.29E–04

Eu-154 2.32E–04

Eu-155 2.54E–05

Hf-172 7.33E–06

Lu-173 1.92E–06

Lu-174 5.42E–06

W-185 5.62E–09

Po-209 9.30E–06

Po-210 9.30E–06

Bi-207 4.71E–06

Bi-210 7.14E–07

Pb-210 5.62E–05

Ra-226 9.73E–06

Th-228 1.47E–05
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 54 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

514 (continued) Waste processing Th-229 6.04E–05

Th-230 5.81E–05

Th-232 7.87E–05

U-232 6.08E–05

U-233 7.01E–03

U-234 3.60E–03

U-235 2.18E–04

U-236 8.92E–07

U-237 1.82E–09

U-238 8.29E–03

Pu-236 3.67E–08

Pu-238 4.72E–04

Pu-239 6.74E–03

Pu-240 1.68E–03

Pu-241 1.49E–04

Pu-242 1.60E–04

Pu-244 1.09E–06

Np-237 6.77E–05

Np-239 2.41E–06

Am-241 4.23E–03

Am-243 8.33E–05

Cm-244 6.76E–05

Cf-249 9.30E–06

514 Tank Farm Unknown H-3 5.45E–01

Be-7 3.33E–07

C-14 2.59E–03

Na-22 5.74E–05

P-32 1.01E–01

P-33 1.86E–04

S-35 6.39E–03

K-40 5.16E–05

Sc-46 1.30E–06

Cr-51 4.79E–07

Fe-55 6.65E–07

Fe-59 1.53E–06

Mn-54 2.88E–06

Co-56 2.01E–05

Co-57 6.68E–03

Co-58 2.14E–05

Co-60 4.51E–05

Zn-65 9.48E–07

Ni-59 1.33E–06

Ni-63 4.09E–06

Y-88 2.12E–05

Y-91 5.66E–06

Sr-90 2.05E–01

Zr-95 1.40E–03

Nb-95 3.08E–05

Mo-99 7.99E–06

Tc-99 2.76E–05

Ru-103 2.54E–06

Ru-106 4.35E–06

Cd-109 8.97E–08
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 55 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

514 Tank Farm 

(continued)

Unknown Sn-113 4.66E–09

Ag-110m 8.65E–07

I-125 5.59E–05

I-131 1.06E–05

Sb-120m 7.32E–07

Sb-124 6.25E–06

Sb-125 9.93E–06

Te-132 2.00E–07

Ba-133 4.99E–05

Ba-140 9.99E–07

Cs-134 1.42E–05

Cs-136 9.31E–07

Cs-137 1.08E–01

La-140 2.00E–08

Ce-139 3.99E–09

Ce-141 2.93E–05

Ce-144 1.87E–04

Nd-147 4.46E–06

Pm-147 2.40E–06

Gd-148 1.28E–05

Sm-151 7.99E–08

Eu-152 1.82E–04

Eu-154 1.19E–04

Eu-155 1.48E–05

Eu-156 1.13E–07

Tb-160 1.20E–08

Hf-172 1.93E–06

Hf-181 2.66E–06

Lu-173 1.60E–06

Lu-174 3.33E–07

W-185 4.79E–09

Au-195 4.26E–06

Hg-203 3.06E–20

Bi-207 1.04E–06

Bi-210 5.06E–07

Po-209 4.50E–06

Po-210 5.06E–07

Pb-210 1.36E–05

Ra-226 1.55E–04

Pa-233 1.60E–09

Th-228 6.02E–03

Th-229 1.28E–05

Th-230 1.27E–05

Th-232 1.14E–04

U-232 1.31E–05

U-233 4.44E–05

U-234 1.63E–03

U-235 7.36E–04

U-236 3.33E–07

U-237 3.46E–04

U-238 1.02E–02

Pu-236 3.46E–08

Pu-238 1.01E–04
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 56 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

514 Tank Farm 

(continued)

Unknown Pu-239 1.50E–03

Pu-240 1.48E–04

Pu-241 3.86E–05

Pu-242 4.37E–05

Pu-244 7.16E–08

Np-237 2.31E–05

Np-239 1.01E–04

Am-241 1.02E–03

Am-243 1.97E–05

Cm-244 4.12E–05

Cf-249 4.63E–06

612 Waste storage and repackaging H-3 9.79E–01

Be-7 1.22E–06

C-14 7.04E–04

Na-22 1.87E–04

P-32 1.29E–02

P-33 1.05E–08

S-35 6.98E–06

Cl-36 3.66E–11

K-40 2.03E–05

Sc-46 1.25E–06

Cr-51 1.34E–06

Fe-55 1.23E–06

Mn-54 4.20E–06

Co-56 7.33E–08

Co-57 5.01E–06

Co-58 1.34E–06

Co-60 1.07E–03

Ni-63 4.89E–08

Zn-65 1.22E–08

Se-75 2.88E–11

Y-88 1.45E–07

Y-91 1.05E–11

Sr-85 8.72E–13

Sr-89 3.93E–12

Sr-90 6.02E–04

Nb-94 5.50E–05

Nb-95 1.34E–06

Zr-95 1.28E–07

Mo-99 1.55E–09

Tc-99 1.60E–05

Rh-102 2.01E–10

Rh-103m 4.10E–13

Ru-106 2.58E–08

Cd-109 4.02E–05

Cd-115 5.25E–09

Ag-110m 2.18E–14

I-125 2.45E–05

I-131 3.15E–09

Sb-124 4.45E–14

Sb-125 4.99E–06

Sm-151 5.10E–11

Ba-133 1.76E–05
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 57 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

612 (continued) Waste storage and repackaging Ba-140 7.64E–09

Cs-134 1.58E–05

Cs-137 5.11E–04

Ce-139 2.44E–13

Ce-141 2.85E–08

Ce-144 1.52E–04

Nd-147 1.59E–09

Pm-147 5.95E–10

Pm-151 4.73E–10

Sm-151 8.72E–13

Gd-146 1.05E–05

Gd-148 1.57E–05

Eu-149 2.79E–12

Eu-152 6.28E–05

Eu-154 6.10E–05

Eu-155 5.54E–06

Tb-160 1.74E–13

Hf-172 2.44E–06

Lu-173 1.22E–06

Lu-174 1.22E–06

Ta-182 1.40E–11

W-185 3.66E–09

Ir-192 8.72E–14

Au-195 5.41E–12

Pt-195m 9.59E–10

Hg-203 8.72E–17

Bi-207 1.17E–06

Bi-210 2.44E–07

Po-209 3.30E–06

Po-210 2.44E–07

Pb-210 1.68E–05

Ra-223 4.73E–12

Ra-226 1.72E–06

Ra-228 2.94E–09

Th-228 1.45E–06

Th-229 1.57E–05

Th-230 1.56E–05

Th-232 1.48E–05

Th-234 1.13E–12

U-232 1.58E–05

U-233 2.08E–05

U-234 4.93E–04

U-235 1.85E–04

U-237 3.39E–08

U-238 2.01E–03

U-239 9.60E–12

Pu-238 1.15E–04

Pu-239 3.89E–04

Pu-240 3.54E–05

Pu-241 1.28E–04

Pu-242 2.60E–05

Pu-244 3.56E–06

Np-237 1.69E–05
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 58 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

612 (continued) Waste storage and repackaging Np-239 6.09E–07

Am-241 3.74E–04

Am-242 6.81E–09

Am-242m 1.57E–09

Am-243 4.11E–04

Am-244 4.45E–06

Cm-244 1.05E–04

Cf-249 3.39E–06

Cf-250 2.24E–12

612 Yard Unknown H-3 1.00E+00

C-14 2.11E–06

P-32 1.61E–07

S-35 3.40E–08

K-40 4.48E–11

Cr-51 4.84E–11

Mn-54 6.11E–13

Co-57 6.38E–13

Co-60 1.19E–10

Ni-59 1.01E–12

Ni-63 2.92E–11

Se-75 2.45E–08

Sr-90 9.84E–12

Nb-95 3.19E–13

Tc-99 2.44E–12

Sb-125 4.09E–12

Cs-134 1.02E–11

Cs-137 3.60E–10

Ce-144 4.25E–10

Pm-147 1.70E–12

Sm-151 5.26E–13

Eu-152 5.45E–10

Eu-154 2.89E–10

Eu-155 6.11E–12

Bi-207 1.81E–13

Bi-214 4.52E–14

Ra-226 3.65E–11

Ra-228 1.22E–11

Th-228 5.50E–10

Th-230 1.17E–14

Th-232 5.69E–10

Th-234 1.17E–11

U-233 5.40E–10

U-234 1.87E–08

U-235 2.47E–09

U-238 1.87E–07

Pu-238 2.43E–09

Pu-239 5.46E–10

Pu-240 2.12E–11

Pu-241 4.43E–09

Pu-242 1.02E–08

Np-239 4.80E–11

Am-241 1.45E–09

Am-242 1.38E–12
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 59 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

612 Yard 

(continued)

Unknown Am-243 4.22E–10

Cm-243 4.89E–11

Cm-244 3.35E–13

Cm-245 3.03E–11

625 Waste operations H-3 2.14E–01

C-14 2.34E–04

P-32 3.61E–10

K-40 8.78E–05

Mn-54 1.36E–06

Co-57 5.74E–08

Co-60 9.89E–06

Sr-90 2.34E–06

Y-88 1.70E–10

Zr-95 2.55E–06

Ru-106 7.01E–07

Cd-109 4.40E–08

Sb-125 9.78E–07

Ba-133 1.01E–07

Cs-134 1.11E–06

Cs-137 1.05E–05

Ce-141 4.46E–09

Ce-144 1.57E–06

Eu-152 4.47E–06

Eu-154 2.55E–06

Eu-155 2.55E–07

Bi-214 2.98E–08

Pb-212 3.83E–08

Pb-214 4.89E–09

Pa-231 4.89E–08

Ra-226 1.51E–05

Ra-228 6.59E–04

Th-228 1.55E–06

Th-230 5.53E–10

Th-232 3.62E–06

Th-234 4.89E–04

U-233 4.46E–08

U-234 2.76E–01

U-235 4.89E–01

U-238 2.32E–03

U-239 2.55E–07

Pu-238 7.68E–05

Pu-239 3.40E–03

Pu-240 3.83E–06

Pu-241 1.30E–02

Pu-242 4.46E–08

Np-237 2.34E–07

Am-241 2.87E–04

Am-243 5.10E–06

Cm-243 9.78E–09

Cm-244 2.34E–09

Cm-245 1.38E–09
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 60 of 113

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING, 1993 to 2001 (continued)

Current

building no. Building activity

Radionuclides and

activity fractions

2561 Unknown U-234 5.00E–01

U-238 5.00E–01

U-234 8.01E–02

U-235 1.12E–02

U-238 8.66E–01

Site 300-801 Unknown H-3 1

Site 300 Pit 7 Unknown H-3 1

Site 300-810A Unknown U-234 8.38E–02

U-235 1.16E–02

U-238 9.05E–01

Site 300-810B Unknown U-234 8.47E–02

U-235 1.13E–02

U-238 9.04E–01

Site 300-850 Unknown H-3 9.6E–01

U-234 3.2E–03

U-235 4.4E–04

U-238 3.4E–02

Site 300-851 Unknown H-3 9.57E–01

U-234 9.57E–04

U-235 1.34E–04

U-238 1.02E–02

Site 300 Well 8 

Spring

Unknown H-3 1

a. Sources, unless otherwise noted: This table was derived from data reported for calendar years 1992 through 2001 as 

part of the National Emission Standards for Hazardous Air Pollutants (40 CFR Part 61, Subpart H) reporting process 

[Harrach et al. 1994; Gallegos et al. 1996; Gallegos and Biermann 1997; Gallegos et al. 1998; Biermann et al. 1999; 

Gallegos et al. 2000, 2001; Harrach et al. 2002, 2003]. As part of this required reporting, LLNL must monitor emissions 

by radionuclide and total activity at all building release points. This table lists only buildings for which data were 

reported; this is not a complete list of all buildings. For a listing of major buildings and activities, see ORAUT [2005b]. To 

develop the table, the isotopic listing of each building in the relevant reports was captured in a building-specific 

spreadsheet. The annual emissions were totaled for each isotope and then for all isotopes in the listing. Activity fractions 

were obtained by dividing the activity total by the “all isotopes” total [ORAUT 2005c]. The activity fraction for each 

radionuclide in the table is that fraction of the total activity released to the environment (i.e., the sum of the fractions for 

each building is equal to “1" or 100% of the mixture). However, those radionuclides with activity fractions equal to “zero”, 

those with half-lives too short to contribute significant dose (i.e., isotopes of nitrogen and oxygen), and those that 

contribute submersion dose only were subsequently deleted from the listing.

b. Source: LLNL [1987]. 
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 61 of 113

 

ATTACHMENT B 

INTERNAL DOSIMETRY CO-EXPOSURE DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Overview................................................................................................................................ 66

B.2 Limitations.............................................................................................................................. 70

B.3 Discussion of the Datasets..................................................................................................... 70

B.3.1 Uranium...................................................................................................................... 70

B.3.2 Gross Alpha, Plutonium, and Americium/Curium ........................................................ 71

B.3.3 Gross Beta and Mixed Fission and Activation Products.............................................. 74

B.3.4 Analysis...................................................................................................................... 75

B.4 Intake Modeling...................................................................................................................... 77

B.4.1 Assumptions ............................................................................................................... 77

B.4.2 Bioassay Fitting and Intakes....................................................................................... 77

B.4.2.1 Uranium ....................................................................................................... 78

B.4.2.2 Plutonium and Americium ............................................................................ 79

B.4.2.3 Gross Beta for Mixed Fission and Activation Products ................................. 83

LIST OF TABLES

TABLE TITLE PAGE

B-1 Americium co-exposure sample statistic data......................................................................... 67

B-2 Beta co-exposure sample statistic data.................................................................................. 67

B-3 Plutonium co-exposure sample statistic data.......................................................................... 68

B-4 Uranium co-exposure sample statistic data............................................................................ 69

B-5 Time intervals for statistical analysis of uranium urine samples.............................................. 71

B-6 Time intervals for statistical analysis of gross beta urine samples .......................................... 75

B-7 Uranium urine bioassay data for 50th- and 84th-percentiles with the period and effective 

bioassay date, 1958 to 1996 .................................................................................................. 75

B-8 Uranium mass and activity intake rates for three different periods.......................................... 79

B-9 50th- and 95th-percentile uranium mass and activity intake rates for three different 

periods...................................................................................................................................79

B-10 50th- and 84th-percentiles for 239Pu urine bioassay data, 1957 to 1996 ................................. 80

B-11 50th- and 84th-percentile intake rates type M 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 81

B-12 50th- and 84th-percentile intake rates of type S 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 81

B-13 50th- and 84th-percentile intake rates of type SS 239Pu from co-exposure plutonium 

urine bioassay data, 1957 to 1996 ......................................................................................... 81

B-14 50th- and 95th-percentile intake rates of type M 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 82

B-15 50th- and 95th-percentile intake rates of type S 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 82

B-16 50th- and 95th-percentile intake rates of type SS 239Pu from co-exposure plutonium urine 

bioassay data, 1957 to 1996 .................................................................................................. 82
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 62 of 113

ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

B-17 50th- and 84th-percentile excretion rates for 241Am urine bioassay data, 1957 to 1996 .......... 82

B-18 50th- and 95th-percentile intake rates of type M 241Am from co-exposure americium urine 

bioassay data, 1957 to 1996 .................................................................................................. 83

B-19 50th- and 84th-percentile excretion rates for gross beta urine bioassay data, 1974 to 

1995.......................................................................................................................................84

B-20 50th- and 84th-percentile intake rates of type F 90Sr, 1974 to 1980........................................ 84

B-21 50th- and 95th-percentile intake rates of type F 90Sr, 1974 to 1980........................................ 84

B-22 50th- and 84th-percentile intake rates of type F 103Ru, 1974 to 1992...................................... 85

B-23 50th- and 84th-percentile intake rates of type S 141Ce and type F 89Sr, 1974 to 1992............. 86

B-24 50th- and 95th-percentile intake rates of type F 103Ru, 1974 to 1992...................................... 86

B-25 50th- and 95th-percentile intake rates of type S 141Ce and type F 89Sr, 1974 to 1992............. 86

B-26 50th- and 84th-percentile intake rates of type F 90Sr, 90Y, 32P, 137Cs, and 14C, after 1992 ....... 87

B-27 50th- and 95th-percentile intake rates of type F 90Sr, 90Y, 32P, 137Cs, and 14C, after 1992 ....... 87

LIST OF FIGURES

FIGURES TITLE PAGE

B-1 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1958 to December 1973 ........................................................... 87

B-2 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1974 to December 1987 ........................................................... 87

B-3 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1988 to December 1996 ........................................................... 88

B-4 Predicted and observed 50th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type F NU .................................................................................. 88

B-5 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1958 to December 1973 ........................................................... 88

B-6 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1974 to December 1987 ........................................................... 89

B-7 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1988 to December 1996 ........................................................... 89

B-8 Predicted and observed 84th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type F NU .................................................................................. 89

B-9 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1958 to December 1973 .......................................................... 90

B-10 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1974 to December 1987 .......................................................... 90

B-11 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1988 to December 1996 .......................................................... 90

B-12 Predicted and observed 50th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type M NU ................................................................................. 91

B-13 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1958 to December 1973 .......................................................... 91

B-14 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1974 to December 1987 .......................................................... 91

B-15 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1988 to December 1996 .......................................................... 92
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

B-16 Predicted and observed 84th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type M NU ................................................................................. 92

B-17 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1959 to December 1973........................................................... 92

B-18 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1974 to December 1987........................................................... 93

B-19 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1988 to December 1996........................................................... 93

B-20 Predicted and observed 50th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type S NU.................................................................................. 93

B-21 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1958 to December 1973........................................................... 94

B-22 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1974 to December 1987........................................................... 94

B-23 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1988 to December 1996........................................................... 94

B-24 Predicted and observed 84th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type S NU.................................................................................. 95

B-25 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1957 to December 1959 ................................................ 95

B-26 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1960 to December 1964 ................................................ 95

B-27 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1965 to December 1966 ................................................ 96

B-28 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1967 to December 1981 ................................................ 96

B-29 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1982 to December 1990 ................................................ 96

B-30 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1991 to December 1996 ................................................ 97

B-31 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1957 to December 1959 ................................................ 97

B-32 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1960 to December 1963 ................................................ 97

B-33 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1964 to December 1966 ................................................ 98

B-34 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1967 to December 1981 ................................................ 98

B-35 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1982 to December 1990 ................................................ 98

B-36 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1991 to December 1996 ................................................ 99

B-37 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1957 to December 1959................................................. 99

B-38 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1960 to December 1964................................................. 99

B-39 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1965 to December 1966............................................... 100

B-40 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1967 to December 1981............................................... 100
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B-41 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1982 to December 1990............................................... 100

B-42 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1991 to December 1996............................................... 101

B-43 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1957 to December 1959............................................... 101

B-44 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1960 to December 1963............................................... 101

B-45 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1964 to December 1966............................................... 102

B-46 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1967 to December 1981............................................... 102

B-47 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1982 to December 1990............................................... 102

B-48 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1991 to December 1996............................................... 103

B-49 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1957 to December 1959 ............................................ 103

B-50 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1960 to December 1964 ............................................ 103

B-51 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1965 to December 1966 ............................................ 104

B-52 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1967 to December 1981 ............................................ 104

B-53 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1982 to December 1990 ............................................ 104

B-54 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation

intake of type SS plutonium, January 1991 to December 1996 ............................................ 105

B-55 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1957 to December 1959 ............................................ 105

B-56 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1960 to December 1963 ............................................ 105

B-57 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1964 to December 1966 ............................................ 106

B-58 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1967 to December 1981 ............................................ 106

B-59 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1982 to December 1990 ............................................ 106

B-60 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1991 to December 1996 ............................................ 107

B-61 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1957 to December 1960 .................................................... 107

B-62 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1961 to December 1965 .................................................... 107

B-63 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1966 to December 1970 .................................................... 108

B-64 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1973 to December 1979 .................................................... 108

B-65 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1957 to December 1965 .................................................... 108
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B-66 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1966 to December 1970 .................................................... 109

B-67 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1971 to December 1996 .................................................... 109

B-68 Predicted and observed 50th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 90Sr, January 1974 to December 1992........................................ 109

B-69 Predicted and observed 84th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 90Sr, January 1974 to December 1992........................................ 110

B-70 Predicted and observed 50th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 103Ru, January 1974 to December 1992...................................... 110

B-71 Predicted and observed 84th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 103Ru, January 1974 to December 1992...................................... 110

B-72 Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F 90Sr, January 1993 to December 1995 ........................................................ 111

B-73 Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F 90Sr, January 1993 to December 1995 ........................................................ 111
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

B.1 OVERVIEW

ORAUT-OTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment [ORAUT 

2005a] describes the general process for analyzing bioassay data for the assignment of doses based 

on co-exposure results.

Bioassay results for LLNL were obtained as a copy of the site’s MAPPER database converted to a

spreadsheet with a cover letter explaining characteristics of the database and pointing out possibilities 

of data recording errors that might exist in the database [Mansfield 2006a]. As sent to the ORAU 

Team, the MAPPER database had approximately 34,900 records. It contained no in vivo counting 

results and had only a few of the many tritium urinalysis results obtained throughout the operation of 

the Laboratory. The in vivo counting results were not available electronically and, therefore, were not 

analyzed as a co-exposure database. (In vivo results for individual energy employees are included in 

the files sent to NIOSH from LLNL.)

The principal radionuclides with potential for intakes at LLNL were tritium, isotopes of plutonium in a 

weapons-grade mixture, isotopes of uranium in mixtures typical of DU, NU, or slightly enriched 

uranium (to 5%), various TRU radionuclides, activation products, and fission products [ORAUT 

2005b]. The MAPPER database had small numbers of analyses for other radionuclides, such as 

241Am, curium, and TRU materials, but the numbers were too small for statistical analysis as an 

individual co-exposure dataset. LLNL generally used the gross alpha analysis to screen for intakes of 

TRU radionuclides with more specific follow-up analyses for the radionuclides of concern.

Because of general applicability and number of measurements, the urinalysis data chosen for this coexposure study were for plutonium, uranium, gross alpha, gross beta, and an analysis labeled MFP

(mixed fission product). Two previous bioassay laboratory managers confirmed that the gross beta 

results and the MFP results were from the same procedure and should be treated as the same 

[ORAUT 2006b; Mansfield 2006b]. 

The statistical analyses of the bioassay data for each radionuclide were performed in accordance with 

the internal dosimetry co-exposure technical information bulletin [ORAUT 2005a] and the procedure 

ORAUT-PROC-0095, Generating Summary Statistics for Coworker Bioassay Data [ORAUT 2006b]. 

The resultant values were input to the IMBA and Internal Dosimetry Tool (IDOT) computer programs 

and a fit to the data for each of the four analyzed groups at the 50th- and 84th-percentile values was 

performed to obtain intake rates for assignment of dose distributions. The 50th-percentile intakes 

were converted to 95th-percentile intakes based on the following methodology:

(B-1) 1 645 - - . 95th percentile intake 50th percentile inta = ke GSD ×

The tables listing the 50th- and 95th-percentile intakes are the values that the dose reconstructor 

should use when assigning co-exposure intakes. The 50th-percentile intakes are assigned as 

lognormal distributions with the associated GSD. The 95th-percentile intakes are assigned as 

constant distributions. 

Co-exposure intakes for aged fission products and research-related radionuclides from 1994 to 1995, 

plutonium from 1991 to 1996, uranium from 1988 to 1996, and americium from 1971 to 1996 can be 

extended to later years. Building 251 transitioned from a Category II Nonreactor Nuclear Facility in 

2002 to a radiological facility in 2005 when equipment and inventory were removed [Gray et al. 2006]. 

During this transition of Building 251, there were no major contamination incidents and no radiation 

overexposures (in fact, doses were far lower than predicted). Individual and collective doses were 
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

maintained under the as low as reasonably achievable policy. Collective annual WB doses were at 

least 3 times lower than the policy goals and more than 10 times lower than conservative dose 

projections. Individual annual external WB doses were less than 150 mrem [Mitchell et al. 2006]. 

Based on this information about Building 251, extending co-exposure intakes to later years is 

applicable.

Attachment C provides guidance on assigning the 50th- or 95th-percentile intakes based on job 

category and level of contact with the radioactive material.

Tables B-1 through B-4 list bioassay sample data statistics that were used to determine the coexposure intakes.

Table B-1. Americium co-exposure sample statistic data. 

Year

Number of 

samples

Number of 

employees

1957 89 78

1958 125 84

1959 133 83

1960 88 62

1961 116 79

1962 121 61

1963 98 59

1964 134 83

1965 139 78

1966–1967 83 54

1968 89 58

1969 106 68

1970 122 85

1971–1972 78 35

1973–1975 51 35

1976–1978 112 75

Year

Number of 

samples

Number of 

employees

1979–1981 127 90

1982 117 64

1983 127 64

1984 140 65

1985 157 92

1986 210 110

1987 160 96

1988 141 81

1989 165 79

1990 173 91

1991 205 104

1992 220 119

1993 319 144

1994 325 138

1995 300 130

1996 173 99

Table B-2. Beta co-exposure sample statistic data. 

Year

Number of 

samples

Number of 

employees

1957–1963 50 41

1964–1965 75 60

1966–1973 43 42

1974–1975 112 57

1976 70 35

1977 96 46

1978–1979 144 75

1980–1981 148 91

1982 120 69

1983 152 79

1984 200 103

1957–1963 50 41

Year

Number of 

samples

Number of 

employees

1985 199 112

1986 265 137

1987 227 143

1988 225 123

1989 218 103

1990 250 125

1991 298 136

1992 287 152

1993 407 179

1994 357 153

1995 298 126
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Table B-3. Plutonium co-exposure sample statistic data. 

Year

Number of 

samples

Number of 

employees

1957 90 79

1958 125 84

1959 133 83

1960 88 62

1961 116 79

1962 121 61

1963 126 73

1964 140 87

1965 154 82

1966 102 57

1967 84 57

1968 112 66

1969 133 84

1970 133 87

1971 182 111

1972 175 95

1973 106 75

1974 113 83

1975 137 85

1976 135 94

Year

Number of 

samples

Number of 

employees

1977 169 107

1978 140 102

1979 119 76

1980 118 80

1981 194 141

1982 314 185

1983 340 180

1984 402 205

1985 384 214

1986 464 253

1987 450 268

1988 492 285

1989 529 300

1990 399 233

1991 377 212

1992 443 260

1993 522 261

1994 493 244

1995 447 223

1996 252 168

The MAPPER database was used for the determination of the co-exposure study datasets, rather 

than the data from individual claims. The cover letter from LLNL transmitting the MAPPER database 

identified 11 potential issues with the data [Mansfield 2006a]. These issues are addressed below.

• Employee identification. While it is helpful to have information linking all of an individual’s 

samples for a given radionuclide for reviewing the data, investigating outliers, and ensuring 

that there is no bias due to a few individuals, errors in a fraction of the identifiers will not affect 

the co-exposure statistics.

• Sample dates. Sample dates are used only in grouping results into datasets. The shortest 

statistical analysis interval is 3 months, so the day of the month that a sample was collected is 

not necessary information for the co-exposure study.

• Analyte. Obvious typographical errors in the analyte were corrected. A list of sample numbers 

associated with likely erroneous results was provided in the cover letter transmitting the 

MAPPER database, and those involving 238Pu were not included in the 239Pu analysis. Tritium 

co-exposure data were not evaluated.

• Results less than zero. The cover letter notes that these obvious problems were corrected 

before sending the dataset to the ORAU Team.

• Result error. Result error has no impact on a co-exposure analysis; all results are used as 

reported.

• Units. As noted by LLNL, the errors were largely typographical errors and the correct values 

are apparent.
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Table B-4. Uranium co-exposure sample statistic data. 

Year

Number of 

samples

Number of 

employees

1958 74 67

1959–1961 87 70

1962–1964 96 44

1965 107 88

1966 113 90

1967 51 31

1968 141 87

1969 95 92

1970 118 101

1971 35 33

1972 272 172

1973 172 138

1974 139 120

1975 S1 146 134

1975 S2 159 128

1976 S1 139 136

1976 S2 95 92

1977 S1 142 133

1977 S2 151 122

1978 S1 155 148

1978 S2 131 120

1979 S1 213 149

1979 S2 204 163

1980 S1 199 108

1980 S2 141 109

1981 S1 157 126

1981 S2 168 131

1982 S1 245 163

1982 S2 271 194

1983 S1 278 195

1983 S2 317 219

1984 S1 286 191

1984 S2 230 165

1985 S1 219 159

1985 S2 213 153

1986 S1 238 178

1986 S2 257 194

1987 Q1 220 153

Year

Number of 

samples

Number of 

employees

1987 Q2 254 171

1987 Q3 221 139

1987 Q4 224 151

1988 Q1 273 150

1988 Q2 256 154

1988 Q3 236 129

1988 Q4 175 119

1989 Q1 175 126

1989 Q2 195 147

1989 Q3 188 131

1989 Q4 230 146

1990 Q1 211 142

1990 Q2 192 121

1990 Q3 210 120

1990 Q4 238 141

1991 Q1 298 187

1991 Q2 286 153

1991 Q3 284 143

1991 Q4 214 134

1992 Q1 313 204

1992 Q2 326 197

1992 Q3 279 168

1992 Q4 278 174

1993 Q1 332 202

1993 Q2 304 188

1993 Q3 324 201

1993 Q4 316 200

1994 Q1 247 183

1994 Q2 152 122

1994 Q3 148 120

1994 Q4 120 98

1995 Q1 189 153

1995 Q2 114 98

1995 Q3 151 110

1995 Q4 194 146

1996 Q1 197 122

1996 Q2 354 206

1996 Q3 182 134

• Media. There are only eight samples with a media code R or S; all were assumed to be urine 

samples. Neither the sample with media code B nor the two samples with a result of 1,747.8 

were included the analyses, both because the same sample number was attributed to two 

individuals and because the result appears to be far outside the range of other urine values, 

indicating a very unusual circumstance or error.

• Priority. Priority of a sample is irrelevant to the co-exposure analysis.

• Exposure area. Because there are limited samples available for the site, no analysis by work 

area was performed. This field was therefore not used in the study.
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

• Result status. All comments were reviewed and samples indicated to be invalid (e.g., spiked, 

cross contaminated, etc.) were excluded from the analysis.

• Assumptions about sample volume. Discussions were held with LLNL personnel, including the 

former and present bioassay laboratory managers, about this issue. A resolution favorable to 

claimants was adopted; see the discussion in Section B.3.2 for details.

B.2 LIMITATIONS

NIOSH has determined, with concurrence from the Secretary of HHS, that MFP doses at LLNL cannot 

be reconstructed between 1950 and 1973, inclusive. For this reason, a class of LLNL employees has 

been added to the SEC. Therefore, only a limited assessment of internal dose can be performed for 

periods before 1974 for workers who were monitored for some type of radiation exposure covered 

under EEOICPA [Leavitt 2008]. If monitoring data are available for workers who are included in the 

SEC, dose is to be assigned as appropriate based on available monitoring data. However, such dose 

reconstructions are still considered partial dose reconstructions.

NIOSH has further determined that internal doses from 233U at LLNL cannot be reconstructed with 

sufficient accuracy between 1958 and 1989, inclusive. For this reason, a class of LLNL employees 

has been added to the SEC [Burwell 2016]. If monitoring data are available for workers who are 

included in the SEC, dose is to be assigned as appropriate based on available monitoring data.

However, such dose reconstructions are still considered partial dose reconstructions.

B.3 DISCUSSION OF THE DATASETS

B.3.1 Uranium

The MAPPER database contained uranium urinalysis data from 1958 to 1996. Nearly all of the results 

were reported as 238U in micrograms per liter. The LLNL cover letter indicates that the results actually 

represent total uranium [Mansfield 2006a]. According to Section 5.2, most uranium exposure was to 

DU or NU. The AVLIS project converted kilogram quantities of NU to slightly enriched uranium, up to 

5%, as a demonstration plant. The uranium for the AVLIS project came from the Feed Materials 

Production Center. The AVLIS project operated in Buildings 175, 177, and the 482 complex from 1973 

through 1999. HEU that came from Y-12 was used throughout the site’s history in only a few selected 

buildings. HEU was used in Building 231 and the Oralloy Shop in Building 321C. Another project 

involving HEU in the 1990s involved the United States and Russian HEU Purchase Agreement. The 

program monitors the Russian process of converting weapons-usable HEU into low-enriched uranium.

A Transparency Monitoring Office in Russia was established in 1996 by DOE and is staffed in part by 

LLNL workers. When identified in MAPPER, results from workers on these projects were excluded 

from the uranium dataset because of the higher specific activity of uranium on that project (31 results).

Uranium results were left in units of micrograms per liter for the statistical analysis and curve fitting.

The uranium dataset has no negative (i.e., <0) results. MDAs for the uranium analysis methods were 

not provided in the LLNL cover letter. Table 5-8 lists the MDA as 5 μg/L through 1973, 1 μg/L for 1974 

through 1978, 0.3 μg/L for 1979 through 1989, and 0.05 μg/L for 1990 through 1999. Most of the 

results less than 1 μg/L were recorded as zero before 1974. In 1974, the database shows a distinctly 

different pattern, with an overall decrease in the general magnitude of the results, a lower reporting 

level at around 0.01 μg/L, and sufficient results between 0.01 to 1 μg/L to support a reasonably 

normal or lognormal pattern implying that a change to a more sensitive analysis technique took place.

The zero results were included in the ranking of the data but not the curve fitting. In several years 
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

before 1974, there were a few results recorded with values between 1 and 0 μg/L. Because there 

were just a few of these and not in all years, they seemed to be anomalous recordings relative to the 

1-μg/L reporting level. Because they might have been misrecorded data, such as misplaced decimals 

or incorrect unit conversion, and because they were clearly anomalous when included in the 

lognormal plots of the results, these results were excluded from the final annual datasets. There were 

37 of these results throughout the period from 1962 through 1973, in comparison with 1,200 results 

for the co-exposure study during this period.

The uranium dataset had entries from recounts on the same sample and entries showing multiple 

same-day samples from a given worker. The recounted samples were identified as having identical 

information, including sample number, but different results and different uncertainty. The multiple 

same-day samples were indicated by having the same worker identification and sample dates, but the 

sample numbers were different, often by one digit. Because two counts on a single sample or two 

samples from a given worker on the same day produces a bias in comparison with other workers’ 

samples that were given only a single count, the multicounted samples were replaced by the average 

of the multiple counts.

Duplicate entries that were obvious and samples that were marked by LLNL as baselines or quality 

control samples were excluded.

Table B-5 lists the time intervals for separate statistical analyses. Intervals were chosen to obtain 

approximately 100 results or more per interval. However, 1967 had only 51 results and 1971 had only 

34 results. (The median and 84th-percentile values for 1971 were also much lower than surrounding 

years, perhaps indicating that there was little work with uranium at the site in 1971.) The statistical 

parameters for each interval (e.g., 50th and 84th percentiles) were assigned to the midpoint of the 

interval.

Table B-5. Time intervals for statistical 

analysis of uranium urine samples. 

Calendar year Analysis interval

1958 Year

1959−1961 Treated as a single interval

1962–1964 Treated as a single interval

1965–1974 Year

1975–1986 6 months

1987−1996 3 months

B.3.2 Gross Alpha, Plutonium, and Americium/Curium

In one analytical procedure used at LLNL, plutonium, americium, neptunium, curium, actinium, and 

thorium were extracted from the urine matrix and counted by a gross alpha technique. This is referred 

to as a “gross alpha” analysis because any measured activity in the urine sample was an indication 

that one or more of these radionuclides was present. In MAPPER these are called “ALPHA” results.

In another procedure the radionuclides were separated from the urine matrix with anion exchange 

resin. Plutonium, americium, and curium were retained on the resin, with the plutonium subsequently 

eluted off. This plutonium fraction of the eluate is referred to as “PU239” in MAPPER and in this 

attachment. Pulse height analysis was used for plutonium analysis starting in 1967 or 1968 (the first 

238Pu results show up in 1968) [ORAUT 2006b], so as of 1968 a PU239 result is interpreted to mean 

239+240Pu.
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

In an interview [ORAUT 2006b] with the [position redacted], it was pointed out that the eluate from the 

anion exchange column that did not contain plutonium was often collected and analyzed by a gross 

alpha counting technique. This effluent would contain americium and curium and the results were also 

reported as ALPHA in MAPPER. Note that the two different ALPHA analyses in MAPPER potentially 

report different radionuclides, so to avoid confusion this analysis is referred to here as “AM/CM,” while 

the results from the procedure described in the first paragraph are referred to as “GA.” A pair of 

PU239 and AM/CM analyses is referred to as a “split” analysis. There was no reliable way to 

determine from the ALPHA results in MAPPER which procedure was used, although it is likely that all 

splits used the anion exchange procedure [ORAUT 2006b], and this will be assumed here.

In summary, four types of analytical results are evaluated here:

• GA. Nonspecific analysis that can potentially report any combination of plutonium, americium, 

neptunium, curium, actinium, and thorium.

• PU239. Analysis specific for 239Pu. Might or might not have an associated AM/CM result.

• AM/CM. Nonspecific analysis that can potentially report any combination of alpha-emitting 

isotopes of americium and curium. Always has an associated PU239 result.

• AM241. Analysis that is specific for 241Am.

And, to be clear:

• ALPHA. Analyte type in MAPPER that could be either GA or AM/CM.

All PU239 and AM/CM urinalysis results were recorded in dpm which, according to the LLNL cover 

letter, should be interpreted as dpm/24 hours with the exception of samples that were split for AM/CM 

and PU239 analyses [Mansfield 2006a] (see additional discussion on splits below). Most GA 

urinalysis results were recorded in dpm, which should also be interpreted as dpm/24 hours. Some GA 

sample results in 1957 and 1958 had units of cpm; because the appropriate conversion factor 

between cpm and dpm was not known, the cpm results were excluded from this analysis.

There were no negative (less than zero) PU239 or AM/CM results in the dataset, but numerous 

results were recorded as zero and some small positive results were below the reported MDA for the 

time. For example, in 1990, the recorded PU239 results descend in a somewhat orderly way down to 

0.001 dpm; then there are 3 results at 1.11 × 10-7

, 9.3 × 10-8, and 7.7 × 10-9 dpm. All of these types of 

extremely small results were indicated in the database as not detected (code 99). However, when 

included as recorded they distorted the curve fitting and so were treated as zero for curve fitting 

purposes. (Another way to look at this problem is that it appears that most of the time results below 

0.001 dpm were recorded as zero, but the policy was not rigorous and some of these very small 

results were entered. It is also possible these samples were analyzed by a method different from all 

the others and so were recorded differently.) Therefore, all PU239 and AM/CM results less than 

0.001 dpm were censored in the analysis (i.e., the rank of the sample was represented in the 

probability plot but the numerical value was not in fitting the lognormal distribution to data for each 

time interval). There were no negative gross alpha results, except in 1996. A lower recording level of 

about 0.01 dpm appears to have been used fairly consistently from 1957 to 1995 with a few results 

recorded in the 0.001- to 0.009-dpm range intermittently throughout the years. As with PU239, all 

ALPHA results equal to or less than 0.001 dpm were censored in the analysis.
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Starting in 1976, some samples were split and given two analyses, one recorded as PU239 and one 

recorded as ALPHA (called AM/CM here). According to the LLNL cover letter (Mansfield 2006a), the 

recorded results were for the split sample, not the total sample, and to obtain a 24-hour result the 

split-sample results should be doubled. However, in a conference call interview with the former 

[position redacted], present [position redacted], and other LLNL personnel, the former [position 

redacted] said that during [pronoun redacted] time in the [building redacted] plutonium/alpha split 

samples were reported as “per total sample volume” [ORAUT 2006b]. The present [position redacted]

confirmed that the interpretation in the cover letter (that splits should be considered as representing 

half the volume) was correct for recent times [Mansfield 2006a; ORAUT 2006b]. No one was sure of 

the exact time when the change in policy occurred. There was agreement that the change occurred in 

the mid- to late 1980s. The former [position redacted] was present in the [building redacted] until 1986 

or 1987. As a consequence, the data were analyzed twice: once with split-sample results being 

doubled starting in 1987 and the other with split-sample results being doubled starting in 1988.

Review of the plutonium data was inconclusive, but continuity of the geometric means for the gross 

alpha results was better for the 1988 assumption. Therefore, a PU239 and an ALPHA result in 

MAPPER on the same day are considered to be “splits.” Splits have different sample numbers, and 

the ALPHA (AM/CM) portion of the split does not contain PU239. Therefore, all PU239 and ALPHA 

(AM/CM) results from splits from 1988 to 1996 were multiplied by a factor of 2.

In later years, it appeared that many samples were recounted and some workers had multiple sameday samples. Multiple results on the same day are “recounts” if they have the same sample number.

The final result for a recount is the mean of the multiple results, which is assigned to one sample, 

discarding the others. Multiple results on the same day are partial samples if they have different 

sample numbers. These results are assumed to be expressed in dpm per partial sample, so the final 

result for this sample is the sum of the results, which is assigned to one sample, discarding the others.

Incident samples are problematic for performing group statistics and applying group median values to 

unmonitored workers because of the large number of samples that often were obtained for a single 

individual after a known acute intake. LLNL sent a list of known large acute intakes or intakes that 

occurred to LLNL employees while working at other sites; this list was used to remove sample results 

from known acute intakes of plutonium. There were 19 samples excluded from the ALPHA results 

from a non-DOE facility. There were 25 samples excluded from the PU239 results from other DOE 

sites. There were 18 samples excluded from the PU239 results from wounds. Other sample results 

removed from the dataset before analysis included duplicate entries and those marked as invalid for 

some reason.

For statistical analysis the data were divided into two categories:

• All PU239 and GA results are considered to be 239Pu.

• All AM/CM, AM-241, and GA results are considered to be americium/curium.

Note that the GA results play a dual role, being considered to measure 239Pu on one hand and 

americium/curium on the other. In general, the time interval for separate statistical analyses was the 

calendar year with the goal of having about 100 or more results per analysis period. There were 

sufficient 239Pu results to permit analyses for each year from 1957 to 1996, but a number of years had 

to be combined for the americium/curium: 1966 and 1967; 1971 and 1972; 1973, 1974, and 1975;

1976, 1977, and 1978; and 1979, 1980, and 1981. The results of the statistical analysis (the 50th- and 

84th-percentile excretion rates) were assigned to the midpoint of the analysis interval, which is July 1 

for an annual interval.
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

B.3.3 Gross Beta and Mixed Fission and Activation Products

There is only one procedure in the 1979 bioassay procedures manual for gross beta analysis, “Gross 

Beta Determination of Fission Products in Urine,” and, according to the former bioassay laboratory, 

that was the only procedure used [Dupzyk 1979; ORAUT 2006b]. As stated above, the gross beta and 

MFP-coded results in MAPPER were treated as a single dataset. The gross beta procedure consisted 

of wet-ashing to white salts, dissolution in nitric acid, coprecipitation with ammonium hydroxide, 

plating on stainless-steel planchets, and gross beta counting.

MAPPER had gross beta or MFP data starting in 1957. In general, the results were recorded as dpm 

or dpm/L, but some results were recorded as cpm, cpm/L, μCi, or μCi/L. The cpm and cpm/L results 

were excluded from the statistical analysis to avoid having to estimate the cpm-to-dpm conversion 

factor. It was also not certain that the unit was entered correctly; that is, the magnitude of the result 

might have been correct, but the unit should have been entered as dpm. The μCi and μCi/L samples 

were also excluded because the magnitudes of the results were not obviously consistent with the μCi 

unit and error in the recording of the units was considered possible. Results reported as dpm/L were 

converted to dpm/24-hour sample by multiplying by 1.4 L/d reference excretion [ICRP 1975].

The only negative numbers in the database were recorded in 1996. Negative numbers and zeros 

were included in the ranking of the results but were not included in the fitting of the lognormal curve.

Starting in 1984, there were situations where two samples were listed for the same worker in the 

same day. Some of these had the same sample numbers but different results. These were assumed 

to be recounts on the same sample. To reduce bias from multiple counts on a single sample, the 

average of the same-day results were used. Also starting in 1984, there were situations where two or 

more samples were listed for the same worker in the same day but had different sample numbers. It is 

not known if these were actually different voidings from the worker or different aliquots of the same 

sample. To reduce bias, these were reduced to a single result. If the results were reported as dpm/L, 

the average of the two results was used before normalization to 24-hour samples; if the results were 

reported as dpm, the two results were added to get the total dpm/d.

Duplicate entries that were obvious and samples that were marked by LLNL as baselines or quality 

control samples were excluded. No samples were excluded due to known large acute intakes. A 

single high outlier in the dataset in 1985 appeared to be an error (5,718,000 dpm/L without a follow-up 

sample or recount and marked as “negative” in the MAPPER field that shows the dosimetrist’s 

disposition of the result). This result was excluded.

The initial statistical analyses of the gross beta results resulted in poor fits and large GSDs for many 

of the years. This situation appeared to be due in part to a scattering of results between 1 to 0 dpm/d 

that were not consistent with a lognormal or even a normal distribution. These results were few in 

number and were scattered over several decades (e.g., a few between 0.1 to 1 dpm and a few 

between 0.01 to 0.1 dpm). An analysis of the 99-code break points (i.e., the bioassay result below 

which almost all samples were labeled as nondetections) for the gross beta analyses implied that the 

decision level for the analysis was in the tens of dpm/d before 1982 and hundreds of dpm/L for 1982 

through 1996. Because these random small results were well below the detection or decision level 

and appeared to be outliers relative to the rest of the distribution, all results less than 1 dpm/d were 

treated as if they were zero. As a consequence, they were included in the ranking but were not part of 

the fitting of the lognormal curve.
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

The entire dataset for 1996 was suspect. The dataset had a definite low bias, with negative numbers 

running into the thousands and with 46 results <0, 101 results recorded exactly as zero, and only 4 

results >0. The distribution for this year was the mirror image of a lognormal distribution. Statistical 

parameters were not generated for this year.

Table B-6 lists the time intervals for separate statistical analyses. Intervals were chosen to obtain 

approximately 100 results or more per interval (although in some years consideration was given to 

how many results were >0). Before 1974, there were too few samples to perform meaningful statistics 

so the intakes were determined only for 1974 through 1995. The statistical parameters for each 

interval (e.g., 50th and 84th percentiles) were assigned to the midpoint of the interval.

Table B-6. Time intervals for statistical analysis of gross 

beta urine samples. 

Calendar year Analysis interval

1957−1973 Intakes not determined (168 samples 

total for the 17-yr period)

1974–1975 Treated as a single interval 

1976 Year (70 samples) 

1977 Year

1978−1979 Treated as a single interval

1980–1981 Treated as a single interval

1982–1995 Year

1996 Not analyzed

The procedure did not state exactly which elements would have been quantitatively captured on the 

planchet. This question was posed to a Hanford [position redacted]; [pronoun redacted] reply was 

[ORAUT 2006c, p. 2]: 

The LLNL procedure will recover and measure nearly everything except K-40, Cs-137, 

tritium, carbon, sulfur, and halogens. It will recover and measure strontium, barium, 

radium, most of the transition metals, all the rare earths, and all the actinides. Some of 

the transition metals that form strong ammine complexes probably aren't completely 

recovered by this method; they include Co, Ni, Cu, Zn, and Ag. Nickel is probably not 

recovered at all. Iron should be recovered and measured completely. Ruthenium is 

probably mostly recovered by this method.

B.3.4 Analysis

For each of the radionuclide groups above, a lognormal distribution for the data in the intervals 

specified in Tables B-1 and B-2 was assumed [ORAUT 2006d]. The 50th- and 84th-percentile values 

were calculated using the method described in ORAUT [2012]. Table B-7 shows the statistical 

analysis results for uranium.

Table B-7. Uranium urine bioassay data for 50th- and 84th-percentiles (μg/L and mg/d) with the period 

and effective bioassay date, 1958 to 1996. 

Period

Effective 

bioassay date

50tha percentile 

(μg/L)

84thb percentile 

(μg/L)

50tha percentile 

(mg/d)

84thb percentile 

(mg/d)

1958 06/30/1958 6.89E+00 1.39E+01 9.64E–03 1.95E–02

1959–61 06/30/1960 6.13E+00 1.48E+01 8.58E–03 2.07E–02

1962–64 06/30/1963 5.77E+00 1.35E+01 8.08E–03 1.89E–02

1965 06/30/1965 7.61E+00 1.54E+01 1.07E–02 2.16E–02
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Period

Effective 

bioassay date

50tha percentile 

(μg/L)

84thb percentile 

(μg/L)

50tha percentile 

(mg/d)

84thb percentile 

(mg/d)

1966 06/30/1966 1.62E+00 6.24E+00 2.26E–03 8.74E–03

1967 06/30/1967 6.91E+00 1.40E+01 9.67E–03 1.96E–02

1968 06/30/1968 5.42E+00 9.34E+00 7.58E–03 1.31E–02

1969 06/30/1969 3.68E+00 7.73E+00 5.15E–03 1.08E–02

1970 06/30/1970 1.91E+00 5.41E+00 2.68E–03 7.58E–03

1971 06/30/1971 9.16E–01 3.89E+00 1.28E–03 5.45E–03

1972 06/30/1972 8.32E+00 2.03E+01 1.16E–02 2.84E–02

1973 06/30/1973 4.31E+00 1.33E+01 6.04E–03 1.87E–02

1974 06/30/1974 4.57E–02 2.61E–01 6.41E–05 3.66E–04

1975 S1c 03/31/1975 6.25E–02 1.87E–01 8.75E–05 2.62E–04

1975 S2c 09/30/1975 7.56E–02 2.95E–01 1.06E–04 4.13E–04

1976 S1c 03/31/1976 3.08E–02 2.09E–01 4.32E–05 2.93E–04

1976 S2c 09/30/1976 2.75E–02 1.79E–01 3.86E–05 2.50E–04

1977 S1c 03/31/1977 2.11E–02 1.03E–01 2.95E–05 1.44E–04

1977 S2c 09/30/1977 2.42E–02 1.90E–01 3.39E–05 2.66E–04

1978 S1c 03/31/1978 3.27E–02 1.65E–01 4.58E–05 2.32E–04

1978 S2c 09/30/1978 3.32E–02 1.72E–01 4.64E–05 2.40E–04

1979 S1c 03/31/1979 2.14E–02 1.07E–01 2.99E–05 1.50E–04

1979 S2c 09/30/1979 3.71E–02 1.74E–01 5.19E–05 2.44E–04

1980 S1c 03/31/1980 2.48E–02 2.75E–01 3.47E–05 3.85E–04

1980 S2c 09/30/1980 8.37E–03 6.44E–02 1.17E–05 9.02E–05

1981 S1c 03/31/1981 9.34E–03 8.28E–02 1.31E–05 1.16E–04

1981 S2c 09/30/1981 8.42E–02 3.66E–01 1.18E–04 5.12E–04

1982 S1c 03/31/1982 3.96E–02 1.98E–01 5.55E–05 2.77E–04

1982 S2c 09/30/1982 8.18E–02 3.61E–01 1.15E–04 5.06E–04

1983 S1c 03/31/1983 2.84E–02 2.23E–01 3.98E–05 3.12E–04

1983 S2c 09/30/1983 3.69E–02 2.04E–01 5.17E–05 2.85E–04

1984 S1c 03/31/1984 2.78E–02 1.55E–01 3.90E–05 2.17E–04

1984 S2c 09/30/1984 4.76E–02 1.59E–01 6.66E–05 2.22E–04

1985 S1c 03/31/1985 3.33E–02 2.18E–01 4.66E–05 3.05E–04

1985 S2c 09/30/1985 2.86E–02 1.10E–01 4.00E–05 1.54E–04

1986 S1c 03/31/1986 2.01E–02 1.37E–01 2.82E–05 1.92E–04

1986 S2c 09/30/1986 5.12E–03 5.04E–02 7.16E–06 7.05E–05

1987 Q1c 02/15/1987 6.01E–03 5.42E–02 8.41E–06 7.59E–05

1987 Q2c 05/15/1987 1.12E–02 6.62E–02 1.56E–05 9.27E–05

1987 Q3c 08/15/1987 2.04E–02 1.05E–01 2.85E–05 1.47E–04

1987 Q4c 11/15/1987 1.90E–02 1.23E–01 2.66E–05 1.72E–04

1988 Q1c 02/15/1988 1.38E–02 1.15E–01 1.93E–05 1.61E–04

1988 Q2c 05/15/1988 6.78E–02 1.58E–01 9.49E–05 2.22E–04

1988 Q3c 08/15/1988 6.88E–02 1.66E–01 9.64E–05 2.32E–04

1988 Q4c 11/15/1988 1.04E–01 2.75E–01 1.45E–04 3.84E–04

1989 Q1c 02/15/1989 7.95E–02 1.88E–01 1.11E–04 2.63E–04

1989 Q2c 05/15/1989 1.04E–01 2.58E–01 1.46E–04 3.61E–04

1989 Q3c 08/15/1989 7.69E–02 2.16E–01 1.08E–04 3.02E–04

1989 Q4c 11/15/1989 1.04E–01 2.83E–01 1.46E–04 3.97E–04

1990 Q1c 02/15/1990 5.56E–02 1.42E–01 7.78E–05 1.99E–04

1990 Q2c 05/15/1990 4.67E–02 1.19E–01 6.54E–05 1.67E–04

1990 Q3c 08/15/1990 4.64E–02 1.02E–01 6.49E–05 1.42E–04

1990 Q4c 11/15/1990 5.33E–02 8.95E–02 7.46E–05 1.25E–04

1991 Q1c 02/15/1991 7.23E–02 1.35E–01 1.01E–04 1.89E–04

1991 Q2c 05/15/1991 8.44E–02 1.59E–01 1.18E–04 2.22E–04
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Period

Effective 

bioassay date

50tha percentile 

(μg/L)

84thb percentile 

(μg/L)

50tha percentile 

(mg/d)

84thb percentile 

(mg/d)

1991 Q3c 08/15/1991 6.28E–02 1.11E–01 8.79E–05 1.55E–04

1991 Q4c 11/15/1991 3.92E–02 7.68E–02 5.48E–05 1.08E–04

1992 Q1c 02/15/1992 3.84E–02 7.15E–02 5.37E–05 1.00E–04

1992 Q2c 05/15/1992 3.67E–02 6.04E–02 5.13E–05 8.45E–05

1992 Q3c 08/15/1992 4.19E–02 7.38E–02 5.86E–05 1.03E–04

1992 Q4c 11/15/1992 3.11E–02 4.56E–02 4.35E–05 6.38E–05

1993 Q1c 02/15/1993 3.58E–02 5.45E–02 5.01E–05 7.63E–05

1993 Q2c 05/15/1993 4.38E–02 6.67E–02 6.14E–05 9.34E–05

1993 Q3c 08/15/1993 4.34E–02 6.85E–02 6.07E–05 9.58E–05

1993 Q4c 11/15/1993 5.26E–02 7.10E–02 7.36E–05 9.94E–05

1994 Q1c 02/15/1994 4.76E–02 6.45E–02 6.67E–05 9.03E–05

1994 Q2c 05/15/1994 4.32E–02 6.54E–02 6.05E–05 9.16E–05

1994 Q3c 08/15/1994 5.07E–02 7.46E–02 7.09E–05 1.04E–04

1994 Q4c 11/15/1994 5.34E–02 7.86E–02 7.47E–05 1.10E–04

1995 Q1c 02/15/1995 6.47E–02 9.70E–02 9.06E–05 1.36E–04

1995 Q2c 05/15/1995 5.13E–02 7.84E–02 7.18E–05 1.10E–04

1995 Q3c 08/15/1995 4.63E–02 6.61E–02 6.48E–05 9.26E–05

1995 Q4c 11/15/1995 4.71E–02 7.43E–02 6.59E–05 1.04E–04

1996 Q1c 02/15/1996 5.24E–02 9.03E–02 7.34E–05 1.26E–04

1996 Q2c 05/15/1996 4.86E–02 8.39E–02 6.81E–05 1.18E–04

1996 Q3c 08/15/1996 5.53E–02 8.90E–02 7.75E–05 1.25E–04

a. 50th-percentile value of the fitted line.

b. 84th-percentile value of the fitted line.

c. S1 means first 6 months of the year, S2 means second 6 months of the year, Q1 means months 1 to 3 of the year, Q2 

means months 4 to 6, Q3 means months 7 to 9, and Q4 means months 10 to 12.

B.4 INTAKE MODELING

B.4.1 Assumptions

All urinalysis results were assumed to be representative of a full-day (24-hour) urinary excretion or 

were normalized to be representative of a full day before creating the lognormal plots with the 

exception of the uranium data, which were normalized to daily excretion before intake modeling [7].

Each result that was used in the intake calculation was assumed to have a normal distribution and a 

uniform absolute error of 1 was applied to all results, which weighted all results equally [8]. A chronic 

exposure pattern was assumed unless the data clearly showed a sharp, short-term increase in 

excretion. Intakes were assumed to be from inhalation using a default breathing rate of 1.2 m3

/hr and 

a 5-μm AMAD particle size distribution [9].

When calculating doses to individuals from bioassay data, a GSD of 3 has been used to account for 

biological variation and uncertainty in the models. It was considered inappropriate to assign a value 

less than 3 for the co-exposure data. Therefore, a GSD of 3 was assigned for any intake period in 

which the calculated GSD was <3 [10].

B.4.2 Bioassay Fitting and Intakes

IMBA and IDOT were used to fit the bioassay results to a series of chronic inhalations. The intake 

assumptions were based on observed patterns in the bioassay data. Intervals with constant chronic 

intake rates were chosen by selecting periods during which the bioassay results were of similar 

magnitudes. A new chronic intake period was started where the data indicated a significant sustained 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

change in the results. The effective bioassay dates that were used in IMBA and IDOT to calculate the 

intake rates were the midpoints of the sampling periods [11]. The 50th- and 84th-percentile excretion 

values were fit as independent datasets producing separate 50th- and 84th-percentile intakes 

[ORAUT 2005a]. The GSDs were calculated as the ratio of the 84th- to the 50th-percentile intakes for 

each intake interval [ORAUT 2005a]. 

This TBD does not specify inhalation absorption types for the various radionuclides or groups of 

radionuclides; therefore, the bioassay results were entered into IMBA and IDOT with assumed lung 

absorption types chosen to be consistent with International Commission on Radiological Protection

(ICRP) Publication 68 [ICRP 1995] or ORAUT-OTIB-0049 [ORAUT 2020]. The plots at the end of this 

attachment show the resultant 50th-percentile intakes. The bioassay data that were used in the fits 

are shown as solid blue dots (●) (dark spots when printed) and data that are not used in the fits are 

shown as red dots (●) (light dots when printed).

Some materials have very long radiological half-lives and/or are retained in the body for long periods.

In such cases, the excretion results for different chronic intake periods are not independent. For 

example, an intake in the 1950s could contribute to urinary excretion in the 1980s and later. To avoid 

potential underestimation of intakes for people who worked at LLNL for relatively short periods, each 

chronic intake of long-lived or long-retained material was independently fit using only the bioassay 

results from the single intake period. This fitting method will result in a best estimate of dose if the 

person worked during only one period and a potential overestimate if an individual worked during 

multiple periods.

B.4.2.1 Uranium

The 50th- and 84th-percentile uranium urinary excretion concentrations in micrograms per liter were 

converted to milligrams per day assuming a daily urine volume of 1.4 L/d [ICRP 1975]. These data are 

shown in Table B-7. The intakes were determined from the milligrams per day excretion values. For 

ease of use by dose reconstructors, the intakes were also converted to activity using the specific 

activity of NU (682 pCi/mg), which is favorable to claimants in comparison with the specific activity of 

DU. While workers at LLNL might not have been chronically exposed to NU, chronic intakes will 

approximate a series of acute intakes with unknown intake dates. Plots of the fits of models to urinary 

excretion are shown in Figures B-1 through B-24 at the end of this attachment.

The intake rates, GSDs, and periods in which they are applicable are given in Table B-8. Table B-9 

lists the 50th- and 95th-percentile co-exposure uranium intakes to be assigned for dose 

reconstruction.

There were no bioassay results for uranium before 1958 in the MAPPER database and, according to 

the laboratory manager circa 1958, few or none might have been obtained, so workers exposed to 

uranium during 1952 through 1957 who did not have subsequent uranium bioassay would be 

considered unmonitored. Therefore, unmonitored uranium doses will not be assessed for the period 

before 1958, which is included in the designated SEC period.
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Table B-8. Uranium mass and activity intake rates for three different periods.

a

Type Dates

50th 

percentile 

(mg/d)

84th

percentile

(mg/d)

50th 

percentile 

(pCi/d)

84th

percentile

(pCi/d) GSD

F 01/01/1958–12/31/1973 2.51E–02 5.81E–02 1.71E+01 3.96E+01 2.32a

F 01/01/1974–12/31/1987 1.65E–04 8.52E–04 1.13E–01 5.81E–01 5.16

F 01/01/1988–12/31/1996 2.87E–04 3.73E–04 1.96E–01 2.54E–01 1.30a

M 01/01/1958–12/31/1973 1.03E–01 2.39E–01 7.04E+01 1.65E+02 2.32a

M 01/01/1974–12/31/1987 6.76E–04 3.49E–03 4.61E–01 2.38E+00 5.16

M 01/01/1988–12/31/1996 1.20E–03 1.582E–03 8.16E–01 1.08E+00 1.32a

S 01/01/1958–12/31/1973 1.24E+00 2.93E+00 8.49E+02 1.99E+03 2.35a

S 01/01/1974–12/31/1987 8.55E–03 4.19E–02 5.83E+00 2.86E+01 4.90

S 01/01/1988–12/31/1996 1.75E–02 3.01E–02 1.19E+01 2.05E+01 1.72a

a. Use the default GSD of 3.0 instead of the calculated GSD.

Table B-9. 50th- and 95th-percentile uranium mass and activity intake rates for three different 

periods.a

Type Dates

50th 

percentile 

(mg/d)

50th 

percentile 

(pCi/d) GSD

95th

percentile

(mg/d)

95th percentile

(pCi/d)

F 01/01/1958–12/31/1973 2.51E–02 1.71E+01 3.00 1.53E–01 1.04E+02

F 01/01/1974–12/31/1987 1.65E–04 1.13E–01 5.16 2.45E–03 1.68E+00

F 01/01/1988–12/31/1996 2.87E–04 1.96E–01 3.00 1.75E–03 1.19E+00

M 01/01/1958–12/31/1973 1.03E–01 7.04E+01 3.00 6.28E–01 4.29E+02

M 01/01/1974–12/31/1987 6.76E–04 4.61E–01 5.16 1.01E–02 6.86E+00

M 01/01/1988–12/31/1996 1.20E–03 8.16E–01 3.00 7.31E–03 4.97E+00

S 01/01/1958–12/31/1973 1.24E+00 8.49E+02 3.00 7.56E+00 5.17E+03

S 01/01/1974–12/31/1987 8.55E–03 5.83E+00 4.90 1.17E–01 7.96E+01

S 01/01/1988–12/31/1996 1.75E–02 1.19E+01 3.00 1.07E–01 7.25E+01

a. Use the intakes in pCi/d assuming all the activity is from 234U, which is favorable to claimants.

B.4.2.2 Plutonium and Americium

Table B-10 provides the 50th- and 84th-percentile values for 239Pu urinary excretion per day. Plots of 

the fits of models to urinary excretion are shown in Figures B-25 through B-60. The 50th- and 84thpercentile intakes based on urinary excretion modeled as 239Pu are summarized in Tables B-11 

through B-13. Type M, S, or SS might have been present in the LLNL workplace, so the dose 

reconstructor should use whichever is more favorable to the claimant. All GSDs are larger than 3, so 

they can be used directly from the table. The plutonium intakes in Tables B-11 through B-13 can be 

considered pure 239Pu; that is, other isotopes in a plutonium mixture should be based on isotopic 

ratios from Table 5-2. Tables B-14 through B-16 list the 50th- and 95th-percentile co-exposure

plutonium intakes to be assigned for dose reconstruction.

For assessing type SS, refer to ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly 

Retained in the Lung [ORAUT 2020].

Table B-17 provides the 50th- and 84th-percentile values for 241Am urinary excretion per day. Plots of 

the fits of models to urinary excretion are shown in Figures B-61 through B-67. Intakes were modeled 

assuming all of the americium/curium results were 241Am. The biokinetic model for curium is basically 

the same as that for americium and produces doses comparable to those from 241Am for longer-lived 

isotopes of curium [ICRP 1995]. Therefore, using 241Am as the modeled radionuclide is favorable to 

claimants. The 50th- and 84th-percentile intakes based on urinary excretion modeled as 241Am are 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

summarized in Table B-17. Table B-18 lists the 50th- and 95th-percentile co-exposure americium 

intakes to be assigned for dose reconstruction. The co-exposure americium intakes should be 

assigned for cases in which there was potential for internal exposure from americium or curium. 

Table B-10. 50th- and 84th-percentiles for 239Pua urine 

bioassay data, 1957 to 1996. 

Year

Effective 

bioassay 

date

50th-percentile 

excretion rate 

(dpm/d)

84th-percentile 

excretion rate 

(dpm/d)

1957 07/01/1957 4.470E–02 2.829E–01

1958 07/01/1958 9.435E–02 1.002E+00

1959 07/01/1959 1.362E–01 1.361E+00

1960 07/01/1960 7.974E–02 3.566E–01

1961 07/01/1961 3.559E–02 4.136E–01

1962 07/01/1962 4.280E–02 1.113E+00

1963 07/01/1963 2.858E–02 2.956E–01

1964 07/01/1964 3.106E–02 1.133E–01

1965 07/01/1965 1.571E–02 1.547E–01

1966 07/01/1966 1.007E–02 1.547E–01

1967 07/01/1967 6.051E–03 3.878E–02

1968 07/01/1968 4.899E–03 2.202E–02

1969 07/01/1969 1.462E–03 1.526E–02

1970 07/01/1970 2.788E–04a 6.192E–03a

1971 07/01/1971 2.427E–03 2.898E–02

1972 07/01/1972 3.008E–03 2.499E–02

1973 07/01/1973 4.306E–03 3.266E–02

1974 07/01/1974 2.651E–03 2.339E–02

1975 07/01/1975 1.395E–02 9.313E–02

1976 07/01/1976 3.893E–03 5.441E–02

1977 07/01/1977 3.085E–03 2.961E–02

1978 07/01/1978 1.541E–03 3.706E–02

1979 07/01/1979 7.989E–03 4.731E–02

1980 07/01/1980 1.759E–02 1.376E–01

1981 07/01/1981 5.392E–03 3.582E–02

1982 07/01/1982 4.136E–04 7.478E–03

1983 07/01/1983 1.650E–03 1.078E–02

1984 07/01/1984 8.048E–04 9.488E–03

1985 07/01/1985 6.502E–04 7.737E–03

1986 07/01/1986 9.787E–04 1.257E–02

1987 07/01/1987 2.037E–03 9.097E–03

1988 07/01/1988 1.207E–03 9.843E–03

1989 07/01/1989 3.082E–03 1.816E–02

1990 07/01/1990 1.092E–03 1.479E–02

1991 07/01/1991 2.890E–04 3.935E–03

1992 07/01/1992 1.308E–04 3.465E–03

1993 07/01/1993 2.425E–04 4.860E–03

1994 07/01/1994 2.670E–04 5.789E–03

1995 07/01/1995 2.991E–04 5.973E–03

1996 07/01/1996 1.357E–04 4.863E–03

a. The data for 1970 were excluded from the analysis because they are 

anomalously low.
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Separated americium was used in Building 251 as early as 1955 for tracer studies. Curium was used 

primarily in Building 332 and in Building 251 around the same time as americium. Americium coexposure intakes should only be assigned in the case for work in Building 251. This can be 

determined based on the worker’s job description, work location, claimant telephone interview, and 

DOE dosimetry records information. In the absence of specific information, the plutonium co-exposure

intakes, as part of a weapons-grade mixture, should be assigned.

There were no bioassay results for plutonium or gross alpha before 1957 in the MAPPER database 

and, according to the [position redacted] circa 1958, few or none might have been obtained [ORAUT 

2006b]. Therefore, unmonitored plutonium doses will not be assessed for the period before 1957, 

which is included in the designated SEC period.

Table B-11. 50th- and 84th-percentile intake rates of type M 239Pu from 

co-exposure plutonium urine bioassay data, 1957 to 1996. 

From To

50th percentile 

(dpm/d)

84th percentile 

(dpm/d) GSD

01/01/1957 12/31/1959 29.25 289 9.88

01/01/1960 12/31/1963 8.15 137.7 16.90

01/01/1964 12/31/1964 8.15 41.32 5.07

01/01/1965 12/31/1966 4.277 41.32 9.66

01/01/1967 12/31/1981 0.74 5.996 8.10

01/01/1982 12/31/1990 0.232 1.881 8.11

01/01/1991 12/31/1996 0.043 0.982 22.59

Table B-12. 50th- and 84th-percentile intake rates of type S 239Pu from 

co-exposure plutonium urine bioassay data, 1957 to 1996. 

From To

50th percentile

(dpm/d)

84th percentile 

(dpm/d) GSD

01/01/1957 12/31/1959 896.4 8,957 9.99

01/01/1960 12/31/1963 167.9 3,423 20.39

01/01/1964 12/31/1964 167.9 1,203 7.16

01/01/1965 12/31/1966 145 1,203 8.30

01/01/1967 12/31/1981 11.37 92.41 8.13

01/01/1982 12/31/1990 4.243 33.96 8.00

01/01/1991 12/31/1996 0.880 20.58 23.39

Table B-13. 50th- and 84th-percentile intake rates of type SS 239Pu 

from co-exposure plutonium urine bioassay data, 1957 to 1996. 

From To

50th percentile 

(dpm/d)

84th percentile 

(dpm/d) GSD

01/01/1957 12/31/1959 3,950 38,800 9.82

01/01/1960 12/31/1963 1,160 19,000 16.38

01/01/1964 12/31/1964 1,160 5,650 4.87

01/01/1965 12/31/1966 579 5,650 9.76

01/01/1967 12/31/1981 98.8 799 8.09

01/01/1982 12/31/1990 31.8 259 8.14

01/01/1991 12/31/1996 6.07 136 22.41
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Table B-14. 50th- and 95th-percentile intake rates of type M 239Pu 

from co-exposure plutonium urine bioassay data, 1957 to 1996. 

From To

50th percentile 

(dpm/d) GSD

95th percentile 

(dpm/d)

01/01/1957 12/31/1959 29.25 9.88 1266

01/01/1960 12/31/1963 8.15 16.9 853

01/01/1964 12/31/1964 8.15 5.07 118

01/01/1965 12/31/1966 4.277 9.66 178

01/01/1967 12/31/1981 0.74 8.10 23.1

01/01/1982 12/31/1990 0.232 8.11 7.25

01/01/1991 12/31/1996 0.043 22.59 7.33

Table B-15. 50th- and 95th-percentile intake rates of type S 239Pu 

from co-exposure plutonium urine bioassay data, 1957 to 1996. 

From To

50th percentile 

(dpm/d) GSD

95th percentile 

(dpm/d)

01/01/1957 12/31/1959 896.4 9.99 39,532

01/01/1960 12/31/1963 167.9 20.39 23,930

01/01/1964 12/31/1964 167.9 7.16 4,284

01/01/1965 12/31/1966 145 8.30 4,709

01/01/1967 12/31/1981 11.37 8.13 357.0

01/01/1982 12/31/1990 4.243 8.00 129.9

01/01/1991 12/31/1996 0.88 23.39 157.2

Table B-16. 50th- and 95th-percentile intake rates of type SS 239Pu 

from co-exposure plutonium urine bioassay data, 1957 to 1996. 

From To

50th percentile 

(dpm/d) GSD

95th percentile 

(dpm/d)

01/01/1957 12/31/1959 3,950 9.82 169,365

01/01/1960 12/31/1963 1,160 16.38 115,338

01/01/1964 12/31/1964 1,160 4.87 15,687

01/01/1965 12/31/1966 579 9.76 24,558

01/01/1967 12/31/1981 98.8 8.09 3,077

01/01/1982 12/31/1990 31.8 8.14 1,002

01/01/1991 12/31/1996 6.07 22.41 1,010

Table B-17. 50th- and 84th-percentile excretion rates for 241Am 

urine bioassay data, 1957 to 1996. 

Year

Effective 

bioassay date

50th percentile

(dpm/d)

84th percentile 

(dpm/d)

1957 07/01/1957 4.313E–02 2.785E–01

1958 07/01/1958 9.435E–02 1.002E+00

1959 07/01/1959 1.362E–01 1.361E+00

1960 07/01/1960 7.974E–02 3.566E–01

1961 07/01/1961 3.559E–02 4.136E–01

1962 07/01/1962 4.280E–02 1.113E+00

1963 07/01/1963 4.462E–02 4.147E–01

1964 07/01/1964 3.329E–02 1.106E–01

1965 07/01/1965 2.384E–02 1.910E–01

1966–1967 12/31/1966 9.702E–03 1.660E–01

1968 07/01/1968 5.216E–03 2.327E–02

1969 07/01/1969 1.179E–03 1.462E–02
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Year

Effective 

bioassay date

50th percentile

(dpm/d)

84th percentile 

(dpm/d)

1970 07/01/1970 8.473E–05a 2.928E–03a

1971–1972 12/31/1971 2.371E–03 2.889E–02

1973–1975 07/01/1974 1.395E–02 1.283E–01

1976–1978 07/01/1977 3.564E–02 7.259E–01

1979–1981 07/01/1980 2.059E–02 1.807E–01

1982 07/01/1982 4.221E–02 2.041E–01

1983 07/01/1983 1.466E–02 9.062E–02

1984 07/01/1984 3.748E–02 2.872E–01

1985 07/01/1985 2.131E–02 1.345E–01

1986 07/01/1986 5.588E–02 2.278E–01

1987 07/01/1987 7.659E–02 5.604E–01

1988 07/01/1988 4.988E–02 3.161E–01

1989 07/01/1989 2.334E–02 2.421E–01

1990 07/01/1990 3.713E–02 2.369E–01

1991 07/01/1991 1.872E–02 1.024E–01

1992 07/01/1992 5.301E–02 2.260E–01

1993 07/01/1993 1.601E–02 1.010E–01

1994 07/01/1994 2.848E–02 2.101E–01

1995 07/01/1995 3.952E–02 1.900E–01

1996 07/01/1996 7.815E–02 3.557E–01

a. The data for 1970 were excluded from the analysis because they are 

anomalously low.

Table B-18. 50th- and 95th-percentile intake rates of type M 

241Am from co-exposure americium urine bioassay data, 1957 to 

1996. 

From To

50th percentile 

(dpm/d) GSD

95th percentile 

(dpm/d)

01/01/1957 12/31/1960 10.1 4.36 113.84

01/01/1961 12/31/1965 3.48 12.6 224.74

01/01/1966 12/31/1970 0.49 11.8 28.41

01/01/1971 12/31/1996 1.93 6.54 42.38

B.4.2.3 Gross Beta for Mixed Fission and Activation Products

Interpreting gross beta urinalyses requires:

• At least approximate knowledge of the ratio of the various possible MFAPs at the time of 

intake,

• How the mix of those radionuclides will be altered by human biokinetics and be manifested in 

urine, and

• How the latter mix will contribute to gross beta counts given the impact of the chemistry and 

counting technique.

Note: See Section 5.2 for source term information.
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Reactor Mixtures, 1974 to 1980

As shown in Figures B-68 and B-69, the gross beta urine data in Table B-19 were modeled as 100% 

type F 90Sr. The resulting intakes should be multiplied by the activity fractions of 90Sr in urine after 

intakes of 10-day old, 40-day-old, 180-day-old, and 1-year-old reactor fuel in Table 7-1a of OTIB-0054 

for the ATR, N Reactor, and TRIGA reactor types. Table B-20 lists the 50- and 84th-percentile intake 

rates of type F 90Sr for 1974 through 1980. The 50th- and 95th-percentile 90Sr intake rates 

(Table B-21) should also be adjusted for the indicator radionuclide activity fractions in Table 7-1a in 

OTIB-0054, and should be used in conjunction with the associated radionuclides’ ratios, excluding 131I,

for the ATR, N Reactor, and TRIGA reactor types for estimating intakes for relevant radionuclides 

based on intake activity relative to 90Sr for 10-day old, 40-day-old, 180-day-old, and 1-year-old reactor 

fuel from Table 7-3a through 7-3c, and 7-3f through 7-3i in OTIB-0054. As noted previously, the LPTR 

operated from December of 1957 to March 31, 1980.

Table B-19. 50th- and 84th-percentile excretion rates for 

gross beta urine bioassay data, 1974 to 1995. 

Year

Effective 

bioassay date

50th percentile 

(dpm/d)

84th percentile 

(dpm/d)

1974–1975 12/31/1974 19.22 199.1

1976 07/01/1976 17.74 136.8

1977 07/01/1977 18.13 103.8

1978–1979 12/31/1978 1.474 55.76

1980–1981 12/31/1980 13.04 165.8

1982 07/01/1982 22.56 236.8

1983 07/01/1983 96.75 636.4

1984 07/01/1984 9.806 99.19

1985 07/01/1985 20.17 145.1

1986 07/01/1986 17.59 203.3

1987 07/01/1987 64.88 395.0

1988 07/01/1988 125.1 777.1

1989 07/01/1989 62.62 544.2

1990 07/01/1990 62.73 455.0

1991 07/01/1991 14.79 145.5

1992 07/01/1992 7.883 76.81

1993 07/01/1993 1.128 19.69

1994 07/01/1994 0.1914 4.956

1995 07/01/1995 0.2789 4.694

Table B-20. 50th- and 84th-percentile intake rates of type F 

90Sr, 1974 to 1980.a

From To

50th percentile 

(dpm/d)

84th percentile 

(dpm/d) GSD

01/01/1974 03/31/1980 62.72 605.52 9.65

a. Intake rates are unadjusted for the indicator radionuclide activity fractions 

in Table 7.1a in OTIB-0054.

Table B-21. 50th- and 95th-percentile intake rates of type F 

90Sr, 1974 to 1980. 

From To

50th percentile 

(dpm/d) GSD

95th percentile 

(dpm/d)

01/01/1974 03/31/1980 62.72 9.65 2,611.88
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

To summarize, the following method is used to assign internal dose from gross beta co-exposure

intakes from reactor mixtures:

1. Determine the 90Sr reactor mixture co-exposure intake rate by multiplying the 90Sr co-exposure

intake rate in Table B-19 by the 90Sr average fraction of beta activity in urine sample for 10-

day-old, 40-day-old, 180-day-old, and 1-year-old reactor fuel for the ATR, N Reactor, and 

TRIGA reactor types in Table 7-1a in OTIB-0054.

2. Determine the associated radionuclides, excluding 131I, using the ratios for estimating intakes 

for intakes based on intake activity relative to 90Sr for 10-day-old, 40-day-old, 180-day-old, and 

1-year-old reactor fuel for the ATR, N Reactor, and TRIGA reactor types in Tables 7-3a 

through 7-3c, and Tables 7-3f through 7-3i in OTIB-0054.

3. Assign the highest dose determined for the above decay times.

The intake rates in Table B-21 are used in conjunction with OTIB-0054 to assign intakes of fission 

product mixtures from reactors based on the indicator radionuclide activity fractions and associated 

radionuclides ratios relative to 90Sr for 10-day old reactor fuel for the ATR, N Reactor, and TRIGA 

reactor types using Table 7.1a, and Tables 7-3a through 7.3c and Tables 7.3f through 7.3i in 

OTIB-0054.

Weapons Residue, 1974 to 1992

As shown in Figures B-70 and B71, the gross beta urinary excretion for 1974 through 1992 was 

modeled as type F 103Ru. Ruthenium-103 intakes were then assigned using 59% of the geometric 

means and 84th percentiles. Additional intakes of type S 141Ce and type F 89Sr were included at 50% 

each of the 103Ru intakes. Tungsten-181 is not a beta emitter and would not have been detected in the 

gross beta urinalysis [Kocher 1981]; however, the additional dose to each organ from using 103Ru 

compensates for the unmeasured intake or dose from 181W. Table B-22 lists the 50th- and 84thpercentile 103Ru intakes. Table B-23 lists the 50th- and 84th-percentile 141Ce, and 89Sr intakes.

Table B-24 lists the 50th- and 95th-percentile 103Ru intakes to be assigned for dose reconstruction.

Table B-25 lists the 50th- and 95th-percentile 141Ce and 89Sr intakes to be assigned for dose 

reconstruction.

Table B-22. 50th- and 84th-percentile intake rates of type F 

103Ru, 1974 to 1992. 

From To

50th percentile 

(dpm/d)

84th percentile

(dpm/d) GSD

01/01/1974 12/31/1982 62.72 605.52 9.65

01/01/1983 12/31/1983 411.48 2,706.57 6.58

01/01/1984 12/31/1986 62.72 605.52 9.65

01/01/1987 12/31/1987 269.51 1,975.44 7.33

01/01/1988 12/31/1988 531.74 3,304.41 6.21

01/01/1989 12/31/1990 269.51 1,975.44 7.33

01/01/1991 12/31/1992 62.72 605.52 9.65
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Table B-23. 50th- and 84th-percentile intake rates of type S 

141Ce and type F 89Sr, 1974 to 1992. 

From To

50th percentile 

(dpm/d)

84th percentile 

(dpm/d) GSD

01/01/1974 12/31/1982 31.36 302.76 9.65

01/01/1983 12/31/1983 205.74 1,353.28 6.58

01/01/1984 12/31/1986 31.36 302.76 9.65

01/01/1987 12/31/1987 134.76 987.72 7.33

01/01/1988 12/31/1988 265.87 1,652.21 6.21

01/01/1989 12/31/1990 134.76 987.72 7.33

01/01/1991 12/31/1992 31.36 302.76 9.65

Table B-24. 50th- and 95th-percentile intake rates of type F 

103Ru, 1974 to 1992. These intake rates are used to assign 

intakes of MFAP mixtures from weapons test residuals.

From To

50th percentile 

(dpm/d) GSD

95th percentile

(dpm/d)

01/01/1974 12/31/1982 62.72 9.65 2,611.88

01/01/1983 12/31/1983 411.48 6.58 9,127.05

01/01/1984 12/31/1986 62.72 9.65 2,611.88

01/01/1987 12/31/1987 269.51 7.33 7,139.55

01/01/1988 12/31/1988 531.74 6.21 10,723.46

01/01/1989 12/31/1990 269.51 7.33 7,139.55

01/01/1991 12/31/1992 62.72 9.65 2,611.88

Table B-25. 50th- and 95th-percentile intake rates of type S 

141Ce and type F 89Sr, 1974 to 1992. These intake rates are 

used to assign intakes of MFAP mixtures from weapons test 

residuals. Note that the indicated intake rate is assigned for 

each radionuclide.

From To

50th percentile 

(dpm/d) GSD

95th percentile 

(dpm/d)

01/01/1974 12/31/1982 31.36 9.65 1,305.94

01/01/1983 12/31/1983 205.74 6.58 4,563.52

01/01/1984 12/31/1986 31.36 9.65 1,305.94

01/01/1987 12/31/1987 134.76 7.33 3,569.91

01/01/1988 12/31/1988 265.87 6.21 5,361.73

01/01/1989 12/31/1990 134.76 7.33 3,569.91

01/01/1991 12/31/1992 31.36 9.65 1,305.94

Aged Fission Products and Research-Related Radionuclides, After 1992

As shown in Figures B-72 and B-73, the gross beta urinary excretion data for 1993 to 1995 were 

modeled as type F 90Sr. Intakes of 32P, 14C, 137Cs, and 90Y equal to the 90Sr intakes were then applied. 

Table B-26 lists the 50th- and 84th-percentile 90Sr, 90Y, 32P, 137Cs, and 14C intakes. Table B-27 lists the 

50th- and 95th-percentile 90Sr, 90Y, 32P, 137Cs, and 14C intakes to be assigned for dose reconstruction.

Because there were few bioassays before 1974, guidance on assigning intakes of beta/gammaemitting radionuclides before 1974 were not developed. Intakes and doses before 1974 are covered 

under the SEC designation [Leavitt 2008].
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Table B-26. 50th- and 84th-percentile intake rates of type F 

90Sr, 90Y, 32P, 137Cs, and 14C, after 1992. (Note that the 

indicated intake rate is assigned for each radionuclide.)a

From To

50th percentile 

(dpm/d)

84th percentile

(dpm/d) GSD

01/01/1993 12/31/1993 5.11 89.5 17.5

01/01/1994 12/31/1995 0.909 19.1 20.9

a. Refer to Section 5.3.2.3 for absorption types for Y-90, P-32, Cs-137, 

and C-14.

Table B-27. 50th- and 95th-percentile intake rates of type F 

90Sr, 90Y, 32P, 137Cs, and 14C, after 1992. (Note that the 

indicated intake rate is assigned for each radionuclide.)a

From To

50th percentile 

(dpm/d) GSD

95th percentile

(dpm/d)

01/01/1993 12/31/1993 5.11 17.5 566.52

01/01/1994 12/31/1995 0.909 20.9 134.96

a. Refer to Section 5.3.2.3 for absorption types for Y-90, P-32, Cs-137, 

and C-14.

Figure B-1. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1958 to December 1973 (blue dots). 

Figure B-2. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1974 to December 1987 (blue dots). 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-3. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1988 to December 1996 (blue dots). 

Figure B-4. Predicted and observed 50th-percentile urinary excretion assuming three separate chronic 

inhalation intakes of type F NU. 

Figure B-5. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1958 to December 1973 (blue dots). 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-6. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1974 to December 1987 (blue dots). 

Figure B-7. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F NU, January 1988 to December 1996 (blue dots). 

Figure B-8. Predicted and observed 84th-percentile urinary excretion assuming three separate chronic 

inhalation intakes of type F NU. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-9. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1958 to December 1973 (blue dots). 

Figure B-10. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1974 to December 1987 (blue dots). 

Figure B-11. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1988 to December 1996 (blue dots). 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-12. Predicted and observed 50th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type M NU. 

Figure B-13. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1958 to December 1973 (blue dots). 

Figure B-14. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1974 to December 1987 (blue dots). 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-15. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M NU, January 1988 to December 1996 (blue dots). 

Figure B-16. Predicted and observed 84th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type M NU (blue dots). 

Figure B-17. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1959 to December 1973 (blue dots). 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-18. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1974 to December 1987 (blue dots). 

Figure B-19. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1988 to December 1996 (blue dots). 

Figure B-20. Predicted and observed 50th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type S NU. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-21. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1958 to December 1973 (blue dots). 

Figure B-22. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1974 to December 1987 (blue dots). 

Figure B-23. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S NU, January 1988 to December 1996 (blue dots). 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-24. Predicted and observed 84th-percentile urinary excretion assuming three separate 

chronic inhalation intakes of type S NU. 

Figure B-25. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1957 to December 1959. 

Figure B-26. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1960 to December 1964. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-27. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1965 to December 1966. 

Figure B-28. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1967 to December 1981. 

Figure B-29. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1982 to December 1990. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-30. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1991 to December 1996. 

Figure B-31. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1957 to December 1959. 

Figure B-32. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1960 to December 1963. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-33. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1964 to December 1966. 

Figure B-34. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1967 to December 1981. 

Figure B-35. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1982 to December 1990. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-36. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M plutonium, January 1991 to December 1996. 

Figure B-37. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1957 to December 1959. 

Figure B-38. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1960 to December 1964. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-39. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1965 to December 1966. 

Figure B-40. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1967 to December 1981. 

Figure B-41. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1982 to December 1990. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-42. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1991 to December 1996. 

Figure B-43. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1957 to December 1959. 

Figure B-44. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1960 to December 1963. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-45. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1964 to December 1966. 

Figure B-46. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1967 to December 1981. 

Figure B-47. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1982 to December 1990. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-48. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type S plutonium, January 1991 to December 1996. 

Figure B-49. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1957 to December 1959. 

Figure B-50. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1960 to December 1964. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-51. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1965 to December 1966. 

Figure B-52. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1967 to December 1981. 

Figure B-53. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1982 to December 1990. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-54. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1991 to December 1996. 

Figure B-55. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1957 to December 1959. 

Figure B-56. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1960 to December 1963. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-57. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1964 to December 1966. 

Figure B-58. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1967 to December 1981. 

Figure B-59. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1982 to December 1990. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-60. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type SS plutonium, January 1991 to December 1996. 

Figure B-61. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1957 to December 1960. 

Figure B-62. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1961 to December 1965. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-63. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1966 to December 1970. 

Figure B-64. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1973 to December 1979. 

Figure B-65. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1957 to December 1965. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-66. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1966 to December 1970. 

Figure B-67. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type M 241Am, January 1971 to December 1996. 

Figure B-68. Predicted and observed 50th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 90Sr, January 1974 to December 1992. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-69. Predicted and observed 84th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 90Sr, January 1974 to December 1992. 

Figure B-70. Predicted and observed 50th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 103Ru, January 1974 to December 1992. 

Figure B-71. Predicted and observed 84th-percentile gross beta urinary excretion assuming a chronic 

inhalation intake of type F 103Ru, January 1974 to December 1992. 
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ATTACHMENT B

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued) 

Figure B-72. Predicted and observed 50th-percentile urinary excretion assuming a chronic inhalation 

intake of type F 90Sr, January 1993 to December 1995. 

Figure B-73. Predicted and observed 84th-percentile urinary excretion assuming a chronic inhalation 

intake of type F 90Sr, January 1993 to December 1995. 
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ATTACHMENT C 

SCALING OF CO-EXPOSURE DOSE INTAKES FOR TYPE AND DURATION OF EXPOSURES

Assigned Intakes Using Case-Specific Information

As indicated previously in Section 5.3, workers were selected for internal monitoring based on 

direction by the program supervisor and health physics staff based on the potential for internal 

exposure. Therefore, the assignment of co-exposure intakes to personnel without internal monitoring 

data should be uncommon. However, for those cases where co-exposure intakes need to be 

assigned, the following guidance should be used.

The major presumptive exposures at LLNL were made up of the primary radionuclides plutonium, 

uranium, and tritium. Other radionuclides might have resulted in intakes to personnel but, due to the 

small amounts of material, the limited operations, and the use of engineering controls such as hot 

cells, fume hoods, gloveboxes, and dry boxes, potential exposures are relatively unlikely. When 

exposures were possible, the intakes were likely to be small and highly dependent on work category.

Work locations are another important consideration when evaluating for a potential unmonitored 

intake. Although it is not possible in all cases to precisely locate an LLNL employee’s work areas, it is 

certainly true that the nonprimary radionuclides presented a limited hazard to most LLNL personnel. In 

this attachment, a method is presented to scale assigned intakes based on job category and likely 

duration of exposure to limit the margin of overestimation.

Level of Contact

For the purpose of dose reconstruction, the amount of exposure should be considered when 

assigning co-exposure dose intakes. For personnel with full contact with the material, the 95th 

percentile of the calculated co-exposure intakes should be assigned. For others, the 50th percentile is 

more appropriate, as listed in Table C-1. The suggested co-exposure intake percentiles listed in Table 

C-1 are suggested guidance. The decision of what co-exposure intake percentile or environmental 

intake to assign must be made by the dose reconstructor taking into account a worker’s job duties, 

work location(s), available dosimetry records, and any other information available in the claimant file.
Document No. ORAUT-TKBS-0035-5 Revision No. 04 Effective Date: 09/01/2020 Page 113 of 113

ATTACHMENT C

SCALING OF CO-EXPOSURE DOSE INTAKES FOR TYPE AND DURATION OF EXPOSURES 

(continued)

Table C-1. Levels of contact for certain job categories. 

Contact 

level Description Example job titles

Percentile of 

co-exposure

intake

Primary 

contact

Contact with material in unencapsulated 

and uncontained form. Normal contact 

with the material is under less controlled 

environments such as in fume hoods 

and inside chemical apparatus.

Chemists, researchers, metallurgists 95th

Secondary 

contact

Normal contact with the material is in 

containment or process equipment, 

occasional contact with airborne material 

during excursions. 

Operators, janitors, 

decommissioning and 

decontamination workers, health 

physics personnel routinely assigned

50th 

Incidental 

contact

Contact with contamination and airborne 

radioactive material only during an 

excursion or incident or as part of an 

exposure during a maintenance, repair, 

waste disposal, decontamination, or 

decommissioning activity.

Janitors, decommissioning and 

decontamination workers, 

maintenance, health physics 

personnel not routinely assigned 

construction activities associated 

with waste disposal or demolition of 

legacy structures

50th 

Contact 

unlikely, but 

presumed

Nonradiological job categories, 

construction activities not associated 

with waste or decontamination and 

decommissioning, but evidence supports 

a potential incidental exposure. 

Construction activities 50th

Contact 

unlikely, not 

presumed

Nonradiological job categories, and no 

evidence to suggest a presumptive 

exposure.

Administrative personnel, clerk, 

secretary, telephone operator, 

draftsman, computer specialista

Assume 

environmental 

internal dose 

a. These are just a few example job titles where internal exposure are considered unlikely. Refer to ORAUT-OTIB-0014, 

Technical Information Bulletin: Assignment of Environmental Internal Doses for Employees Not Exposed to Airborne 

Radionuclides in the Workplace [ORAUT 2004], for other examples of job categories that might have little or no potential 

for workplace internal exposures.