ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 107

DOE Review Release 12/27/2010

Document Title:

Lawrence Livermore National Laboratory – 

Occupational Internal Dose

Document Number: ORAUT-TKBS-0035-5 

Revision: 02 

Effective Date: 12/13/2010

Type of Document: TBD

Supersedes: Revision 01 

Subject Expert(s): Stephen Spanos, Donald E. Bihl, and Mutty M. Sharfi

Site Expert(s): Thomas R. LaBone

Approval: Signature on File Approval Date: 12/06/2010

Ralph W. Kenning, Document Owner

Concurrence: Signature on File Concurrence Date: 12/07/2010

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 12/07/2010

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 12/09/2010

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 12/13/2010

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 2 of 107

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.
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 3 of 107

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

11

5.1 Introduction ............................................................................................................................13

5.1.1 Purpose......................................................................................................................14

5.1.2 Scope.........................................................................................................................14

5.1.3 Special Exposure Cohort Petition Information for LLNL .............................................. 15

5.2 Source Term ..........................................................................................................................15

5.2.1 Plutonium ...................................................................................................................16

5.2.2 Uranium......................................................................................................................17

5.2.3 Tritium ........................................................................................................................18

5.2.3.1 Metal Tritides ............................................................................................... 19

5.2.3.2 Organically Bound Tritium............................................................................ 19

5.2.4 Other Limited-Exposure Radionuclides.......................................................................20

5.3 Measurement Methods .......................................................................................................... 20

5.3.1 Analysis of Gross Alpha Measurements .....................................................................22

5.3.2 Analysis of Gross Beta Measurements .......................................................................24

5.3.2.1 Reactor Mixtures (1974–1980)..................................................................... 24

5.3.2.2 Weapons Residue (1974–1992)................................................................... 25

5.3.2.3 Aged Fission Products and Research-Related Radionuclides 

(Post 1992) .................................................................................................. 26

5.3.3 Measurement Types and Detection Levels ................................................................. 27

5.3.4 Reporting Formats and Codes....................................................................................31

5.3.5 Assessment of Intake for Unmonitored Workers.........................................................32

5.3.6 Unmonitored Intakes of LLNL Radionuclides Other than Primary 

Radionuclides.............................................................................................................32

5.4 Significant Incidents with Internal Dose Potential ................................................................... 34

5.5 Attributions and Annotations .................................................................................................. 34

References ........................................................................................................................................ 37

Glossary ............................................................................................................................................44

ATTACHMENT A, ENVIRONMENTAL RELEASES BY BUILDING ...................................................49

ATTACHMENT B, INTERNAL COWORKER .....................................................................................65

B1. Overview................................................................................................................................65

B2. Limitations..............................................................................................................................69

B3. Discussion of the Datasets..................................................................................................... 69

B3.1 Uranium......................................................................................................................69

B3.2 Gross Alpha, Plutonium, and Americium/Curium ........................................................71

B3.3 Gross Beta and MFP..................................................................................................73

B3.4 Analysis......................................................................................................................75
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B4. Intake Modeling......................................................................................................................77

B4.1 Assumptions...............................................................................................................77

B4.2 Bioassay Fitting and Intakes.......................................................................................77

B4.2.1 Uranium ....................................................................................................... 78

B4.2.2 Plutonium and Americium ............................................................................ 79

B4.2.3 Gross Beta/MFP .......................................................................................... 83

ATTACHMENT C, SCALING OF COWORKER DOSE INTAKES CONSISTENT WITH 

 TYPE AND DURATION OF EXPOSURES........................................................ 107
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LIST OF TABLES

TABLE TITLE

5-1 Isotopic information for aging of weapons-grade plutonium....................................................

PAGE

17

5-2 Isotopic information for various uranium enrichment levels .................................................... 17

5-3a Y-12 recommended recycled uranium contaminant ratios defaults based on 

upper levels of expected ranges............................................................................................. 18

5-3b Fernald recycled uranium contaminant ratios for natural uranium .......................................... 18

5-4 Routine in vitro sampling frequencies..................................................................................... 21

5-5 In vivo sampling frequencies .................................................................................................. 22

5-6 Bioassay in vitro detection levels............................................................................................ 27

5-7 Bioassay in vivo detection levels ............................................................................................ 28

5-8 Bioassay record codes........................................................................................................... 31

5-9 Radionuclides other than primary radionuclides ..................................................................... 33

5-10 Input parameters for significant incidents ............................................................................... 34

A-1 Environmental releases by building........................................................................................ 49

B-1 Americium coworker sample statistic data.............................................................................. 66

B-2 Beta coworker sample statistic data ....................................................................................... 67

B-3 Plutonium coworker sample statistic data............................................................................... 67

B-4 Uranium coworker sample statistic data................................................................................. 68

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

B-6 Time intervals for statistical analysis of gross beta urine samples .......................................... 74

B-7 Uranium urine bioassay data for 50th and 84th percentiles in units of μg/L and 

mg/day along with the period and effective bioassay date, 1958 through 1996 ...................... 75

B-8 Uranium mass and activity intake rates for four different periods............................................ 78

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

periods...................................................................................................................................78

B-10 Fiftieth- and 84th-percentiles for 239Pua urine bioassay data for 1957 through 

1996.......................................................................................................................................79

B-11 Fiftieth- and 84th-percentile Type M 239Pu intake rates from coworker plutonium 

urine bioassay data for 1957 through 1996 ............................................................................ 80

B-12 Fiftieth- and 84th-percentile Type S 239Pu intake rates from coworker plutonium 

urine bioassay data for 1957 through 1996 ............................................................................ 80

B-13 50th- and 95th-percentile Type M 239Pu intake rates from coworker plutonium 

urine bioassay data for 1957 through 1996 ............................................................................ 80

B-14 50th- and 95th-percentile Type S 239Pu intake rates from coworker plutonium 

urine bioassay data for 1957 through 1996 ............................................................................ 81

B-15 Fiftieth- and 84th-percentiles for 241Am urine bioassay data for 1957 through 

1996.......................................................................................................................................81

B-16 Fiftieth- and 84th-percentile Type M 241Am intake rates from coworker plutonium 

urine bioassay data for 1957 through 1996 ............................................................................ 82

B-17 50th- and 95th-percentile Type M 241Am intake rates from coworker plutonium 

urine bioassay data for 1957 through 1996 ............................................................................ 82

B-18 Fiftieth- and 84th-percentiles for gross beta urine bioassay data for 1974 through 

1995.......................................................................................................................................83

B-19 Fiftieth- and 84th-percentile intake rates of Type F 90Sr for 1974 through 1980...................... 84

B-20 50th- and 95th-percentile Type F 90Sr intake rates for 1974 through 1980 ............................. 84

B-21 Fiftieth- and 84th-percentile intake rates of Type F 103Ru for 1974 through 1992.................... 84

B-22 Fiftieth- and 84th-percentile intake rates of Type S 141Ce and Type F 89Sr for 

1974 through 1992................................................................................................................. 85

B-23 50th- and 95th-percentile Type F 103Ru intake rates for 1974 through 1992 ........................... 85
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B-24 50th- and 95th-percentile Type S 141Ce and Type F 89Sr intake rates for 1974 

through 1992..........................................................................................................................85

B-25 Fiftieth- and 84th-percentile intake rates of Type F 90Sr, 90Y, 32P, 137Cs, and 14C 

for after 1992..........................................................................................................................86

B-26 50th- and 95th-percentile Type F 90Sr, 90Y, 32P, 137Cs, and 14C intake rates for after 

1992.......................................................................................................................................86

C-1 Levels of contact for certain job categories .......................................................................... 107
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 7 of 107

LIST OF FIGURES

FIGURE TITLE

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

inhalation intake of Type F natural uranium from January 1958 through 

December 1973 .....................................................................................................................

PAGE

86

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

inhalation intake of Type F natural uranium from January 1974 through 

December 1987 .....................................................................................................................87

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

inhalation intake of Type F natural uranium from January 1988 through 

December 1996 .....................................................................................................................87

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

separate chronic inhalation intakes of Type F natural uranium............................................... 87

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

inhalation intake of Type F natural uranium from January 1958 through 

December 1973 .....................................................................................................................88

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

inhalation intake of Type F natural uranium from January 1974 through 

December 1987 .....................................................................................................................88

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

inhalation intake of Type F natural uranium from January 1988 through 

December 1996 .....................................................................................................................88

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

separate chronic inhalation intakes of Type F natural uranium............................................... 89

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

inhalation intake of Type M natural uranium from January 1958 through 

December 1973 .....................................................................................................................89

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

inhalation intake of Type M natural uranium from January 1974 through 

December 1987 .....................................................................................................................89

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

inhalation intake of Type M natural uranium from January 1988 through 

December 1996 .....................................................................................................................90

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

separate chronic inhalation intakes of Type M natural uranium .............................................. 90

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

inhalation intake of Type M natural uranium from January 1958 through 

December 1973 .....................................................................................................................90

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

inhalation intake of Type M natural uranium from January 1974 through 

December 1987 .....................................................................................................................91

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

inhalation intake of Type M natural uranium from January 1988 through 

December 1996 .....................................................................................................................91

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

separate chronic inhalation intakes of Type M natural uranium .............................................. 91

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

inhalation intake of Type S natural uranium from January 1959 through 

December 1973 .....................................................................................................................92
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B-18 Predicted and observed 50th-percentile urinary excretion assuming a chronic 

inhalation intake of Type S natural uranium from January 1974 through 

December 1987 .....................................................................................................................92

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

inhalation intake of Type S natural uranium from January 1988 through 

December 1996 .....................................................................................................................92

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

separate chronic inhalation intakes of Type S natural uranium............................................... 93

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

inhalation intake of Type S natural uranium from January 1958 through 

December 1973 .....................................................................................................................93

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

inhalation intake of Type S natural uranium from January 1974 through 

December 1987 .....................................................................................................................93

B-23 Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type S natural uranium from January 1988 through 

December 1996 .....................................................................................................................94

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

separate chronic inhalation intakes of Type S natural uranium............................................... 94

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

inhalation intake of Type M plutonium from January 1957 through December 

1960.......................................................................................................................................94

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

inhalation intake of Type M plutonium from January 1961 through December 

1964.......................................................................................................................................95

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

inhalation intake of Type M plutonium from January 1965 through December 

1968.......................................................................................................................................95

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

inhalation intake of Type M plutonium from January 1969 through December 

1981.......................................................................................................................................95

B-30 Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type M plutonium from January 1957 through December 

1996.......................................................................................................................................96

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

inhalation intake of Type M plutonium from January 1957 through December 

1963.......................................................................................................................................96

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

inhalation intake of Type M plutonium from January 1964 through December 

1967.......................................................................................................................................97

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

inhalation intake of Type M plutonium from January 1968 through December 

1981.......................................................................................................................................97

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

inhalation intake of Type M plutonium from January 1982 through December 

1990.......................................................................................................................................97

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

inhalation intake of Type M plutonium from January 1991 through December 

1996.......................................................................................................................................98

B-36 Predicted and observed 50th-percentile urinary excretion assuming a chronic 

inhalation intake of Type S plutonium from January 1957 through December 

1960.......................................................................................................................................98
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B-37 Predicted and observed 50th-percentile urinary excretion assuming a chronic 

inhalation intake of Type S plutonium from January 1961 through December 

1965.......................................................................................................................................98

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

inhalation intake of Type S plutonium from January 1966 through December 

1968.......................................................................................................................................99

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

inhalation intake of Type S plutonium from January 1969 through December 

1981.......................................................................................................................................99

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

inhalation intake of Type S plutonium from January 1982 through December 

1990.......................................................................................................................................99

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

inhalation intake of Type S plutonium from January 1991 through December 

1996..................................................................................................................................... 100

B-42 Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type S plutonium from January 1957 through December 

1963..................................................................................................................................... 100

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

inhalation intake of Type S plutonium from January 1964 through December 

1966..................................................................................................................................... 100

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

inhalation intake of Type S plutonium from January 1967 through December 

1969..................................................................................................................................... 101

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

inhalation intake of Type S plutonium from January 1970 through December 

1981..................................................................................................................................... 101

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

inhalation intake of Type S plutonium from January 1982 through December 

1990..................................................................................................................................... 101

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

inhalation intake of Type S plutonium from January 1991 through December 

1996..................................................................................................................................... 102

B-48 Predicted and observed 50th-percentile urinary excretion assuming a chronic 

inhalation intake of Type M 241Am from January 1957 through December 1960................... 102

B-49 Predicted and observed 50th-percentile urinary excretion assuming a chronic 

inhalation intake of Type M 241Am from January 1961 through December 1965................... 102

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

inhalation intake of Type M 241Am from January 1966 through December 1970................... 103

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

inhalation intake of Type M 241Am from January 1973 through December 1979................... 103

B-52 Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type M 241Am from January 1957 through December 1965................... 103

B-53 Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type M 241Am from January 1966 through December 1970................... 104

B-54 Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type M 241Am from January 1971 through December 1996................... 104

B-55 Predicted and observed 50th-percentile gross beta urinary excretion assuming a 

chronic inhalation intake of Type F 90Sr from January 1974 through December 

1992..................................................................................................................................... 104

B-56 Predicted and observed 84th-percentile gross beta urinary excretion assuming a 

chronic inhalation intake of Type F 90Sr from January 1974 through December 

1992..................................................................................................................................... 105
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B-57 Predicted and observed 50th-percentile gross beta urinary excretion assuming a 

chronic inhalation intake of Type F 103Ru from January 1974 through December 

1992..................................................................................................................................... 105

B-58 Predicted and observed 84th-percentile gross beta urinary excretion assuming a 

chronic inhalation intake of Type F 103Ru from January 1974 through December 

1992..................................................................................................................................... 105

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

inhalation intake of Type F 90Sr from January 1993 through December 1995....................... 106

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

inhalation intake of Type F 90Sr from January 1993 through December 1995....................... 106
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 11 of 107

ACRONYMS AND ABBREVIATIONS

AMAD activity median aerodynamic diameter

AVLIS Atomic Vapor Laser Isotope Separation

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

D-38 depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

F fast (solubility rate)

g gram

HEPA high-efficiency particulate air

HEU highly enriched uranium

HPGe hyperpure germanium

hr hour

HT elemental tritium or tritiated gas

HTO tritiated water vapor

ICP-MS inductively coupled plasma–mass spectrometry

in. inch

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

KPA kinetic phosphorescence analysis

L liter

LANL Los Alamos National Laboratory

LEGe low energy Germanium detector 

LEU low enriched uranium

LLNL Lawrence Livermore National Laboratory

LPTR Livermore Pool-Type Reactor

LSC liquid scintillation counting

M moderate (solubility rate)

M&E Mechanical and Electrical (Engineering Divisions)

MDA minimum detectable amount

MEQ muscle equivalent 

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission product

mg milligram
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 12 of 107

mm millimeter

MT metal tritide

NaI sodium iodide

nCi nanocurie

NESHAP National Emission Standard for Hazardous Air Pollutants

NIOSH National Institute for Occupational Safety and Health 

NTS Nevada Test Site

OBT organically bound tritium 

ORAU Oak Ridge Associated Universities

pCi picocurie

POC probability of causation

R&D research and development

S slow (solubility rate)

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TRU transuranic

U.S.C. United States Code

WB whole-body

WBC whole-body counter

yr year

§ section or sections

μCi microcurie

μg microgram

μm micrometer
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 13 of 107

5.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010):

] guidelines established under 

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

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

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2010).

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

conventional structures

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

reasons

 

1

 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 14 of 107

5.1.1

This TBD documents 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.

Purpose

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

LLNL for dose reconstructions for NIOSH in relation to the 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

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

ingestion, inhalation, or absorption of radioactive materials in the course of work. This section 

provides site-specific information to the dose reconstructor for use in reconstructing radiation doses 

for workers that were employed at LLNL, via multiple pathways of exposure including inhalation, 

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

intakes inside buildings. The dose reconstructor calculates intakes from information in the files, using 

information from this section to interpret data. 

Scope 

Two classes of LLNL employees have been added to the Special Exposure Cohort (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 from 1957 to the present;

• Internal dose from mixed fission products (MFPs) from 1974 to the present;

• Internal dose from americium from 1957 to the present;

• Internal dose from plutonium from 1957 to the present;

• Internal dose from uranium from 1958 to the present;

• Internal dose from other radionuclides present at LLNL may be reconstructed from 1957/1958

to the present. The dose reconstructor should refer to the respective section on when the 

internal dose for that particular radionuclide can be reconstructed. 

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

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

The remainder of this section describes the SEC, 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.
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 15 of 107

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

workers. Attachment B provides coworker intakes. Attachment C describes scaling of coworker dose 

intakes consistent with type and duration of exposures.

5.1.3

Classes Added to the SEC

Special Exposure Cohort Petition Information for LLNL

NIOSH has determined, with concurrence from the Secretary of Health and Human Services (HHS), 

that MFP doses at LLNL cannot be reconstructed with sufficient accuracy between 1950 and 1973, 

inclusive. For this reason, the following class of LLNL employees has been added to the SEC (Leavitt

2008). 

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.

NIOSH has subsequently determined that MFP doses for workers not monitored for radiation 

exposure at LLNL also cannot be reconstructed with sufficient accuracy between 1950 and 1973, 

inclusive. For this reason, the following class of LLNL employees has been added to the SEC 

(Sebelius 2010). 

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. 

The SEC class includes all workers during the SEC period. Because of the identified dose 

reconstruction infeasibility, all dose reconstructions for all workers having employment during the SEC 

period are considered partial dose reconstructions. If monitoring data are available for workers

included 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 the HHS has determined that 

exposure to mixed fission and activation products during the SEC period 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 

radioactive materials; persons who were assigned to operate processes or laboratory research had a 

potential for exposure (ORAUT 2005a; Mansfield 1989). Administrative personnel and others 

assigned to support the individual buildings had a lesser potential for exposure given the presence of 

engineering and administrative controls. For the purposes of dose reconstruction, it can be assumed 

that internal source terms were introduced at LLNL’s opening on September 2, 1952.
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 16 of 107

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

intakes of radionuclides was urine bioassay, with the primary focus on 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, MFPs, a variety of tracer 

radionuclides, iodine, and tritium (Mansfield 2000). Air monitoring in workplaces has been a common 

surveillance method. A brief description of the building process was provided in ORAUT (2005b, 

Table 2-2). The buildings that were assigned to handle plutonium are highlighted. 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 provided building location information for a worker’s 

claim. 

In vivo methodologies began on an investigational basis in 1964, focusing on high-energy (i.e., 

greater-than-200-keV) gamma emitters. In the 1970s, there were attempts 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 2000).

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

that quality control steps were an integral part of the process (Miller 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.2

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 non-nuclear technologies and materials testing has been ongoing since 

at least 1976 (LLL ca. 1978). LLNL conducted above-ground (atmospheric) and underground tests at 

both the Pacific Proving Ground, and at the Nevada Test Site from 1953 to 1992. Although the tests 

were conducted offsite, test planning and preparation were done at LLNL. In 1971, LLNL managed 

the CANNIKIN test event 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

If a monitoring result refers to “weapons-grade plutonium” or plutonium, dose reconstructors can use 

the isotopic mix listed in Table 5-1 for material aging times (Mansfield 2000). Until the fifth year of site 

operation, fresh plutonium should be assumed. For years 5 through 9, assume a 5-year-old 

plutonium mixture. After these times, 10-year-old plutonium should be assumed. For assessing 

Super S solubility, refer to Estimating Doses for Plutonium Strongly Retained in the Lung (ORAUT

2010a). 

Plutonium

 

2 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).
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 17 of 107

Table 5-1. Isotopic information for aging of weapons-grade 

plutonium (Mansfield 2000). 

Mixture designation Fresh 5-yr 10-yr 20-yr

Years of aging 0 5 10 20

Specific activity in mixture (Ci/g)

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

Activity ratios

Pu-239+240:Am-241 N/A 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

5.2.2

If a monitoring result refers to “uranium,” dose reconstructors can use the isotopic mix listed in 

Table 5-2 for enrichment levels (Mansfield 2000). LLNL received uranium metal from the Y-12 plant 

that was known to contain recycled uranium contaminants, including technetium, neptunium, thorium, 

and plutonium [1]. ORAUT (2006a) discusses the typical concentration of recycled uranium 

contaminants. Table 5-3a provides the list of recycled uranium ratios to be applied for uranium 

material received from Y-12. The recycled uranium ratios listed in Table 5-3a reflect upper bound

values. Given that there is uncertainty in the specific batches of uranium received by LLNL, the upper 

bound values in Table 5-3a should be used for dose reconstruction. LLNL also used natural uranium 

from Fernald for the Atomic Vapor Laser Isotope Separation (AVLIS) project. Table 5-3b provides the 

list of recycled uranium ratios to be applied for uranium associated with the AVLIS project. In the 

absence of specific information, the Y-12 recycled uranium ratios in Table 5-3a should be used.

Uranium

Table 5-2. Isotopic information for various uranium enrichment levels (Mansfield 2000)a. 

Mixture Radionuclide

Mass 

fraction

Activity 

fraction

Specific activity

of mix (Ci/g)

DU U-238 9.975E-1 9.02E-1 3.72E-7 

DU U-235 2.50E-3 1.45E-2

DU U-234 5.00E-6 8.4-E-2

U-nat. U-238 9.93E-1 4.77E-1 7.00E-7

U-nat. U-235 7.20E-3 2.23E-2

U-nat. U-234 5.60E-5 5.00E-1

AVLIS <5% enriched U-238 9.50E-1 4.11E-1 7.78E-7 

AVLIS <5% enriched U-235 5.00E-2 1.39E-1

AVLIS <5% enriched U-234 5.60E-5 4.50E-1

95% HEU U-238 3.96E-2 1.98E-4 6.72E-5

95% HEU U-235 9.50E-1 3.06E-2

95% HEU U-234 1.04E-2 9.69E-1

a. When U238, U28, or U38 was reported, these were units of total U mass (Mansfield 2006a).
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 18 of 107

Table 5-3a. Y-12 recommended recycled uranium contaminant ratios defaults based on upper levels 

of expected ranges (ORAUT 2006a).

Isotope Default levels in pCi/μg U Default levels in dpm/μg U

Default levels in pCi/pCi or 

dpm/dpm total U

Highly 

Enriched 

Uranium & 

Oralloy

Depleted, 

Natural, 

and 2% 

Enriched 

Uranium

Highly 

Enriched 

Uranium & 

Oralloy

Depleted, 

Natural, 

and 2% 

Enriched 

Uranium

Highly 

Enriched 

Uranium & 

Oralloy

Depleted, 

Natural, and 

2% Enriched 

Uranium

Tc-99 4 0.3 0.88 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 0.444 2.00E-03

Pu-239 0.2 (Oralloy 

only)

0.003 0.444

(Oralloy only)

0.00666 2E-03 (Oralloy 

only)

1.50E-03

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

isotopic mix equivalent to natural uranium (see Attachment B) in almost all cases at LLNL. The 

Table 5-3b. Fernald recycled uranium contaminant ratios for 

natural uranium. (ORAUT 2004).

Isotope

Recycled uranium contaminant ratios

for natural uranium

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

uranium in question would be either depleted or natural uranium. Highly enriched uranium (HEU) 

which came from Y-12 was used in Building 231, and the Oralloy Shop in Building 321C. HEU was 

restricted at LLNL, both in areas and amounts used, due to criticality concerns. HEU machining was 

done in Building 321C in either enclosures/hoods or close capture ventilation systems. Building 231 

is an experimental, manufacturing, assembly, test, and materials-handling facility. Building 231 also 

housed research and development activities. Building 241, which conducted materials science 

research, had an old glove box line in Room 1600 which contained enriched uranium. As result, 

internal exposure potential to HEU in Building 241 would have been remote. Building 327 is a 

radiography facility which conducted nondestructive testing of materials, 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. 

5.2.3

Tritium was encountered in several forms: tritium oxide as water or gas (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 mid1950s. 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. Tritiumfilled 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 Arthur 1975), 

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

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 

Tritium
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was filtered because experimentation with metal tritides 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 used in Buildings 212 and 292. Operations at Building 212 were the forerunner to 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 were terminated in 1987. Operations at Building 292 began in 1978 and were terminated in 

1987. This facility took over tritium operations that previously had been conducted at Building 212. 

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

would have had the potential of exposure to the MT targets would be chemists, researchers, health 

physics technicians, and accelerator operators and mechanical technicians who performed work on 

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 of a hood 

which was exhausted through a HEPA Filter. The target was inserted in one side of the hood and the 

target was changed through glove ports on the other side of the hood. Individuals involved in the 

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

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

the target change operation. 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 used on 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 anti-contamination 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 their 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). 

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

available information, there is no indication that MT solubility 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 facilities listed above, the 

claimant telephone interview 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.

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 claimant 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 20 of 107

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

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 activity of the source) that are not addressed above has 

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

given the requirements of 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 2000). This is also evident by the radionuclides assessed by lung counting or 

whole-body (WB) counting listed in Table 5-7 in Section 5.3.3. The dosimetry records would provide 

this information to the dose reconstructor as part of the case-specific information. In addition, 

Attachment C provides a method to scale the intakes provided to present a credible exposure to 

individuals exposed to rarely encountered radionuclides.

Other Limited-Exposure 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-in. by 2-in. sodium iodide detector for detection and quantification of 

relatively high (i.e., greater-than-200-keV) gamma photons. It had some capability for locating the 

source of radionuclide deposition in subjects. This device was calibrated with a water tank phantom 

spiked with 137Cs and 40K (Hickman undated). 

In April 1965, thin crystal studies began (Schmidt and Anderson 1965). In 1965 and 1966, LLNL 

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

that time (Anderson 1964-1967).

Workers were selected for WB or specific organ counting based on 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, with routine monitoring occurring 

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

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

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

involved 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 for involved employees appear to have been increased to quarterly, with 

weekly sampling for Building 331 (for tritium) and monthly sampling for Building 251 (Balanda 1966). 

Workers were selected for internal monitoring based on the potential of internal exposure through the 

1980s, in conjunction with the program supervisor and the area Health Physicist (Mansfield 1989). 

The selection of workers for internal monitoring based on the potential of internal exposure continues 

at LLNL (Mansfield 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 established a routine bioassay monitoring program as early as 1971 where all Site 300 

personnel received an employment start and termination WB count. In addition, persons who worked 
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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 (NAG 1971). Starting in 1980 at 

Site 300, all persons working at, or who can be expected to spend a considerable portion of their time 

in, areas where an airborne uranium hazard might exist will require an annual lung count and a 

semiannual urine analysis. Other schedules might be required if radionuclides other than uranium are 

used and considered a potential internal hazard (Straume 1980). Tables 5-4 and 5-5, compiled from 

a variety of LLNL documents (cited in the table footnotes), 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 2000). 

Table 5-4. Routine in vitro sampling frequencies.

a,b

Radionuclide Period Building

Measurement 

type Frequency

Ar-37 1980c 151 Urine Weekly, monthly

Gross alpha, beta (LSC) 2000–present Site-wide Urine As applicable

HTO 1989c Site-wide Urine Weekly, monthly

HTO 2000–present Site-wide Urine Quarterly, monthly, biweekly, or weekly

Kr-85 1980c 151 Urine Weekly, monthly

N-13 1980c 151 Urine Weekly, monthly

Workplace-specific 1966c 101 Urine Quarterly, annual

Workplace-specific 1966c 102 Urine Annual

Workplace-specific 1966c 110 Urine Annual

Workplace-specific 1966c 112 Urine Annual

Workplace-specific 1966c 114C Urine Annual

Workplace-specific 1966c 117 Urine Quarterly, annual

Workplace-specific 1966c 121 Urine Semiannual

Workplace-specific 1966c 125 Urine Semiannual

Workplace-specific 1966c 127 Urine Semiannual

Workplace-specific 1966c 170 Urine Annual

Workplace-specific 1966c 171 Urine Annual

Workplace-specific 1966c 172 Urine Weekly, monthly

Workplace-specific 1966c 173B Urine Annual

Workplace-specific 1966c 190 Urine Semiannual, monthly

Workplace-specific 1966c 193 Urine Annual

O-15 1980c 151 Urine Weekly, monthly

P-32 1989c Site-wide Urine Monthly

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

Pu 2000–present Site-wide Urine Semiannual

TRU 2000–present Site-wide Urine Annual, semiannual

U 1980c 231 Urine Quarterly

U 1983c 251 Urine Quarterly

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

U 2000–present Site-wide Urine Quarterly, monthly

Weapons-grade Pu 1989c Site-wide Urine Semiannual

a. Sources: Mansfield (2000); Balanda (1966, Appendix A); LLNL (1987a); Ozaki (1980); Wilson (1982); Leahy (1983); 

LRL (1961, 1964); Gibson (1985).

b. These are typical in vitro monitoring frequencies, and 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 beginning date for the stated 

bioassay period.
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Table 5-5. In vivo sampling frequencies.

a, b

Radionuclide Period Building

Measurement 

type Frequency

As detected 2000–present Site-wide WB scans Annual

I-131 1989-1999c Site-wide Thyroid Monthly

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

MFPs 1964–present Site-wide WB scans Not available

Pu, Am 1970–present Site-wide Lung count Annual

Pu, weapons-grade Pu 1989c Site-wide Lung Annual

Pu-239 1961–present Site-wide Wound Incident-specific

U 1985c Site 300 Lung count Annual

U, Th 1987 Bldg. 321 Lung count Annual

U, Th 2000–present Site-wide Lung count Annual, semiannual

a. Sources: Mansfield (2000); Balanda (1966, Appendix A); LLNL (1987a); Ozaki (1980); Wilson (1982); Leahy (1983); 

LRL (1961, 1964); Gibson (1985).

b. These are typical in vivo frequencies, and 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 beginning date for the stated 

bioassay period.

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: (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/or (2) In the 

case of suspected uptakes or for follow-up, mass spectrometry is 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 activity for the same sample, the lower of the two values 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

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 (Miller 1979). The results were reported as gross alpha, but all three involved 

the chemical separation of several different elements, depending on the method that was used (Bihl 

2006a). The records are not clear about which method was used and require some review by the 

dose reconstructor. The methods were capable of separating uranium from the sample before 

analysis to reduce the potential interference (Miller 1979). 

Analysis of Gross Alpha Measurements

The three urine procedures involving alpha (not including uranium) are described below (Miller 1979):

1. Analytical Method for Determining Gross Alpha Activity in Urine. 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 captures plutonium.

2. Determination of Plutonium and Americium in Urine by Separation. This method is specific for 

plutonium and americium. It involves precipitation with ammonium hydroxide, dissolution in 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 23 of 107

nitric, and then adsorption on an anion exchange column. The americium does not stick on 

the column so it comes 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. Meanwhile, the 

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

3. Urinalysis for Curium by Electrodeposition. This method has the same first steps as procedure 

1, but has a third precipitation using lanthanum hydroxide. This separates curium from all the 

other alpha emitters. The precipitate is electroplated, and then can be counted using “pulseheight analysis,” low background proportional counting, or autoradiography.

Gross alpha results were a mix of procedure 1 or 2. The database does not indicate which procedure 

was used. Sometimes the gross alpha analysis might have been specific for americium, other times it 

included natural thorium and plutonium. To account for either possibility [4]:

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

sequentially), assume there is no plutonium in the gross alpha result and assess it as 241Am.

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.

ii. Use the plutonium intake to project the plutonium content of the urine samples on 

the dates of the gross alpha results. 

iii. Subtract the plutonium component from each of the gross alpha samples.

iv. Use the adjusted sample results to assess a 241Am intake. If the individual also had 

241Am results (from step 1), include those in the assessment (do not

b. If there are no plutonium results for the individual:

subtract 

plutonium content from them).

i. If the work location is Building 251

1. assess the gross alpha results twice, once assuming plutonium and once 

assuming 241Am.

2. Select the nuclide yielding the largest dose to the organ of interest.

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

assuming plutonium.

Exceptions occur if there is something in the file that shows the worker being 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 close capture ventilation 

systems. Natural thorium, along with depleted and natural uranium was also used at Site 300 in high 

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

was getting regular uranium analyses, and then had a gross alpha bioassay analysis, this suggests 

that the worker had been working with the 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 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 24 of 107

appropriate. If there is a WB count, the thorium intake calculated from the gross alpha urinalysis 

could be compared against the detection limit of the WB count. 

From mid-2002 to late 2003, bioassay records contained an analyte termed “GA-X.” 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 GA1. The new GA-1 region spans 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 spans 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 [5].

5.3.2

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

included some separation chemistry that removed 40K (Bihl 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, most of the 

transition metals unless they form strong amine complexes (such as cobalt, nickel, copper, zinc, and 

silver), all the rare earths, and all the actinides. 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 (Bihl 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 

represented many potential nuclides. Where gross beta samples were reported in units of cpm, the 

dose reconstructor should assume a conversion of 0.4 cpm/dpm (ORAUT 2009).

Analysis of Gross Beta Measurements

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

• Fission and activation products 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 up 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. After 1980 only weapons residue should be 

assumed for the gross beta results up through 1992. Reactor mixtures should only be assumed up 

through 1980, since 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 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 by the ratio of the WBC radionuclide intake to the specific gross beta 

radionuclide intake. 

5.3.2.1 Reactor Mixtures (1974–1980)

Fission and Activation Product Assignment for Internal Dose-Related Gross Beta and Gross Gamma 

Analyses, ORAUT-OTIB-0054 (ORAUT 2007b) discusses how to use gross beta urine bioassay 

results to assign intakes of fission and activation products. Of the three types of gross beta analysis 

discussed in OTIB-0054, the gross beta analysis used at LLNL is considered to be best described as 

“Gross beta analysis that includes all fission and activation products except radioiodines and beta 

emitters naturally in urine such as 40K” (see Section 7.1 of OTIB-0054). The default assumption for 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 25 of 107

this is 10-day-old reactor fuel (Table 7-1 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 reactor fuel. 

2. Determine the associated radionuclides using the ratios for estimating intakes for intakes 

based on intake activity relative to 90Sr for 10-day-old reactor fuel in Table 7-3 in OTIB-0054.

5.3.2.2 Weapons Residue (1974–1992)

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 the Nevada Test Site (NTS) (Bihl 

2007). The fission/activation product mixtures 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 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 Sedan 

(1962), Schooner (1968), Sulky (1964), Cabriolet (1968), and Palanquin (1965) (Bihl, Brackett, and 

Smith 2007). 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, presumably, some build-up of contamination occurred in hoods and ventilation systems. 

Therefore, exposure was probably a mix of contamination from current cores and older contamination 

from previous cores. Therefore, the analysis compared ratios for 100 days after each shot as being 

representative of the age of contamination accessible to the LLNL workers (Bihl, Brackett, and Smith 

2007).

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:

 

Tungsten-185 0.58

Tungsten-181 0.36

Ruthenium-103 0.0070

Tungsten-188 0.0064

Yttrium-91 0.0051

Iron-55 0.0046

Niobium-95 0.0044

Strontium-89 0.0041

Cerium-141 0.0036

Zirconium-95 0.0033

Ruthenium-106 0.0022

Beryllium-7 0.0019

Manganese-54 0.0017

Cerium-144 0.0013
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 26 of 107

Cobalt-58 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 (Bihl, Brackett, and Smith 2007). It is unlikely the tungsten isotopes or 7Be were 

source terms for workers other than for the NTS core analysis program; for instance, 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 ET1) 

per unit activity in urine than 185W (Bihl, Brackett, and Smith 2007). 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 betaemitting radionuclide for all five shots. But 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. 

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% 103Ru type F, 25% 141Ce type S, and 

25% 89Sr type F. 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 needed for beta yield. Additional intakes of 141Ce type S and 89Sr type F should be included at 50% 

each of the 103Ru intakes. Tungsten-181 is not a beta emitter so 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/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. Ce-141 (type S) intake = 0.5 × 103Ru intake.

b. Sr-89 (type F) intake = 0.5 × 103Ru intake.

5.3.2.3 Aged Fission Products and Research-Related Radionuclides (Post 1992)

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

use that information for determining the internal dose. 

If no specific beta bioassay results are available, intakes should be modeled as 90Sr type F 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 post 1992, the radionuclides in 

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

Attachment A that would have been measured by the gross beta procedure, 90Sr and 137Cs was 

considered representative. They tended to be among the more significant radionuclides in many of 

the beta-exposure facilities and one is a bone seeker and one is a whole-body 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 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 27 of 107

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, as this is favorable to the claimant. In addition, the listing in Attachment A for the 

waste operations buildings (513, 514, 514 tank farms, 613, 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 Sr-90, 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 also be included. Therefore, the 

following method is used to assign internal dose from gross beta results from aged fission products 

and research-related radionuclides:

1. Determine the 90Sr (type F) intake assuming 55 percent of the gross beta results.

2. P-32 (type F, M) intake = 90Sr intake.

3. C-14 (type F, M, S) intake = 90Sr intake.

4. Cs-137 (type F) intake = 90Sr intake.

5. Y-90 (type F) intake = 90Sr intake.

5.3.3

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

methods. These methods are listed in Tables 5-6 and 5-7, and the documents are cited in the table 

footnotes. 

Measurement Types and Detection Levels

Table 5-6. Bioassay in vitro detection levels.

a

Measurement 

type

Analysis 

method

Selected

radionuclided Period MDA

Fecal Radiometric alpha 

spectrometry

Pu-239, 238 1977–1980 1 nCi/sample

Fecal Co-precipitation and 

proportional counting

Pu-239, 238 1990–1999 0.04 dpm

Urine Co-precipitation, 

separation 

(gross alpha method)

Am 1957–1989 0.718 dpm/sample

Urine Co-precipitation, 

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 

alphab

Po-210 1955–1978 0.1 pCi/L

Urine 24-hr sample, alpha 

spec

Po-210 2000–present 0.01 dpm
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 28 of 107

Measurement 

type

Analysis 

method

Selected

radionuclided Period MDA

Urine Co-precipitation Gross alpha 1957–1989 0.78 dpm/sample

Urine Co-precipitation Gross alpha 1990–1999 0.365 dpm/sample

Urine Co-precipitation, 

separation and 

proportional counting 

Pu-total alpha 1957–1989 0.051 dpm/sample

Urine Co-precipitation, 

separation and 

proportional counting

Pu-total alpha 1990–1999 0.012 dpm/samplef

Urine 24-hr sample, alpha 

spec

Pu 2000–present 0.01 dpm

Urine Gross beta MFP 1957–1989 600 dpm/Lg

Urine Gross beta MFP 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 U depleted, natural 

U, U AVLIS

1958–1973 5 μg/L

Urine Fluorometric U depleted, natural 

U, U AVLIS

1974–1978 1 μg/L

Urine Fluorometric U depleted, natural 

U, U AVLIS

1979–1989 0.3 μg/L

Urine Spot sample KPA U depleted, natural 

U, 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) b

HEU 1955–1970 100 dpm/Le

Urine Extraction/alpha 

proportional countingb

HEU 1971–1976 15 dpm/L

Urine Extraction/alpha 

proportional countingb

HEU 1977–1986 4 pCi/L

Urine Alpha spectrometryc HEU 1987–1999 0.3 pCi/L

a. Sources: Mansfield (2000); LRL (1960, 1964, 1970); Griffith (1980); Berger (2005); Schmidt and Anderson (1965); 

ORAUT (2009, 2007a), LLL ca. 1978, Miller (1979). Dupzyk (1990), Simpson (1989), Szalinski . 

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

c. Based on ICRP (1989) information.

d. Other nuclides might have been detected by the method; these might be reported in workers’ files. 

e. Based on information from Los Alamos National Laboratory (LANL). Decision level of 50 dpm/L reported. Assume the 

MDA is 2 times the decision level. 

f. MDA range of 0.012 to 0.051 dpm/sample was indicated. DRs should use available sample specific information. 

g. Gross beta method MDA was 3.84 cpm/sample with a detector efficiency range of 20 to 31 percent, and 60 percent

sample recovery. This results in a MDA range of approximately 400 to 600 dpm/sample depending upon detector 

efficiency (Miller 1979). DRs should use available sample specific information.

Table 5-7. Bioassay in vivo detection levels.

a

Measurement 

type

Analysis 

method Radionuclide Period MDA

Lung Phoswichb Pu-239, Am-241 1966–1969 400 nCi, 2 nCi

Lung Phoswichc 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 - 5.25 cm 

CWT)
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 29 of 107

Measurement 

type

Analysis 

method Radionuclide Period MDA

Lung Dual Phoswich Pu-239 1980–present >16.3 nCi

(2.1 cm CWT)

Lung -- “Pure” Pu-239 1984–1985 9 - 200 nCi (1.45 - 5.95 cm 

CWT)

Lung HPGe arrays “Pure” Pu-239 2000–present 200 nCi

Lung -- “Pure” Pu-238 1984–1985 4 - 100 nCi (1.5 - 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 -- Weapons-grade Pu 1984–1985 1 - 4 nCi (1.45 - 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 -- DU, natural U Circa 1977 5 nCi

Lung -- DU 1984–1985 0.5 - 2 nCi

Lung -- Natural U 1984–1985 0.7 - 2 nCi

Lung -- 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, natural U 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, U-235 2000–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 -- I-131 1984–1985 1 nCi

Thyroid count -- 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 -- 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 K40)

WB scan Shadow shield counter 150- to 300-keV 

gamma emitters

1964–present 500 gammas per second 

between 150 and 300 keV

WB Scan -- Th-232 1970–present 1 nCi (based on Tl-208 with no 

correction for branching ratio)

WB scan -- Natural Th (Tl-208) 2000–present 1 nCi

WB scan -- I-131 (TB) 1970–1995 1 nCi

WB scan -- Cs-137 1970–1995 1 nCi

WB scan -- Co-60 1970–1995 1 nCi

WB scan -- Ce-144 1970–present 8 nCi

WB scan -- Ce-141 1970–present 2 nCi

WB scan -- Ru-103 1970–present 3 nCi

WB scan -- Ru-106 1970–1995 5 nCi
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 30 of 107

Measurement 

type

Analysis 

method Radionuclide Period MDA

WB scan -- Zr-95 1970–present 1 nCi

WB scan -- Mn-54 1970–1995 1 nCi

WB scan -- Zn-65 1970–present 1 nCi

WB scan -- Na-22 1970–1995 1 nCi

WB scan -- K-40 1996–1999 4.46 nCi

WB scan -- Mn-54 1996–1999 0.63 nCi

WB scan -- Fe-59 1996–1999 0.98 nCi

WB scan -- Co-60 1996–1999 0.66 nCi

WB scan -- Ru-106 1996–1999 6.53 nCi

WB scan -- I-131 1996–1999 1 nCi

WB scan -- Cs-134 1996–1999 0.7 nCi

WB scan -- Cs-137 1996–1999 0.89 nCi

WB scan -- Co-57 2000–2005 0.97 nCi

WB scan -- Co-60 2000–2005 0.51 nCi

WB scan -- Cs-137 2000–2005 0.7 nCi

WB scan -- I-131 2000–2005 0.66 nCi

WB scan -- K-40 2000–2005 2.6 nCi

WB scan -- Na-22 2000–2005 0.44 nCi

WB scan -- K-40 2006 2.6 nCi

WB scan -- Cs-137 2006 0.72 nCi

WB scan -- Co-60 2006 0.43 nCi

WB scan -- I-131 2006 0.67 nCi

WB scan -- Na-22 2006 0.42 nCi

WB scan -- Pa-233 2006 1.5 nCi

WB scan -- Ac-228 2007–present 2.74 nCi

WB scan -- Ba-133 2007–present 0.85 nCi

WB scan -- Bi-207 2007–present 0.47 nCi

WB scan -- Bi-212 2007–present 7.71 nCi

WB scan -- Bi-214 2007–present 1.82 nCi

WB scan -- Cf-249 2007–present 0.83 nCi

WB scan -- Cm-243 2007–present 3.59 nCi

WB scan -- Co-57 2007–present 0.97 nCi

WB scan -- Co-60 2007–present 0.37 nCi

WB scan -- Cs-134 2007–present 0.56 nCi

WB scan -- Cs-137 2007–present 0.7 nCi

WB scan -- Eu-152 2007–present 1.75 nCi

WB scan -- Eu-154 2007–present 1.25 nCi

WB scan -- Eu-155 2007–present 2.58 nCi

WB scan -- I-131 2007–present 0.68 nCi

WB scan -- K-40 2007–present 2.52 nCi

WB scan -- Na-22 2007–present 0.4 nCi

WB scan -- Np-237 2007–present 6.38 nCi

WB scan -- Np-239 2007–present 4.73 nCi

WB scan -- Pa-231 2007–present 21.83 nCi

WB scan -- Pa-233 2007–present 1.6 nCi

WB scan -- Pa-234 2007–present 3.93 nCi

WB scan -- Pb-212 2007–present 2.23 nCi

WB scan -- Pb-214 2007–present 2.36 nCi

WB scan -- Po-210 2007–present 3.40E+04 nCi

WB scan -- Ra-224 2007–present 20.6 nCi

WB scan -- Ra-226 2007–present 23.9 nCi

WB scan -- Th-231 2007–present 11.6 nCi
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 31 of 107

Measurement 

type

Analysis 

method Radionuclide Period MDA

WB scan -- Th-234 2007–present 15.8 nCi

WB scan -- U-235 2007–present 1.61 nCi

a. Sources: Mansfield (2000); LRL (1964, 1970); Schmidt and Anderson (1965); Griffith (1980); Berger (2005); ORAUT 

(2009, 2007a); LLL (ca. 1978, 1979); King (1977); Author unknown (ca. 1985); Sundsmo and Hickman (2009a,b,c); 

Hickman (2008).

b. Based on LANL information.

c. Based on ICRP (1989) information.

In some cases, there is evidence that LLNL implemented monitoring by a particular methodology, but 

site-specific detection levels for that methodology were not available. In these cases, detection levels 

for comparable LANL methods were used to complete Tables 5-6 and 5-7 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, thus increasing the likelihood for shared procedures 

(Berger 2005).

In a few instances, neither LLNL nor LANL detection levels were available, requiring a surrogate 

source of information. ICRP Publication 54 (ICRP 1989) specified nominal detection levels for a 

variety of measurement methods that were comparable to those used at LLNL. 

A review of in vivo bioassay records provided by LLNL indicates that the data typically includes the 

radionuclide assessed, 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 5-7. 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

There are a variety of codes on urine bioassay records generated at or for LLNL. Table 5-8 is a 

summary of the codes known at present, along with their interpretation.

Reporting Formats and Codes

Uncertainties associated with bioassay measurements were not stated in the records.

Table 5-8. Bioassay record codes.

Column no. Type of entry Acceptable entries

1–6 Employee number 6-digit integer

8–27 Employee name Last name, comma, other names and/or initials

28–32 Nuclide analyzed MFP, ALPHA, PU239, PU238, U238, AM241, CM244, or other nuclides 

(maximum of 5 characters)

34–39 Date Month, day, year (2 digits each)

41–48 Results Floating point number between 9999999.0 and 0.000 (F8.3 format)

50–51 Error One standard deviation as a percentage of the result

53 Sample type U = urine, B = blood, f = feces, n = nose swipe

55 Sample type R = routine; S = special

57–60 Address Building or trailer number (e.g., B253 or 2532)

61–63 L code 3-digit L code

65 Frequency W = weekly; M = monthly; Q = quarterly; S = semiannually; A = 

annually

67–70 Location of exposure Building number (e.g., B253)
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 32 of 107

Column no. Type of entry Acceptable entries

72–76 Analysis number Integer value of analysis number assigned to sample by Bioassay 

Laboratory

79–80 Method Analytical method used to process samples: 1 = gross alpha technique 

(since 1957); 2 = gross beta technique (since 1974); 3 = plutonium 

separation technique (since 1963); 4 = fluorometric uranium technique 

(since 1974); 5 – 99 = reserved for future revised techniques

5.3.5

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 may not have been monitored, but should be assigned an intake based on their 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 site-wide requirement whereas 

bioassay was determined based on the potential of exposure by facility supervisors and Health 

Physics. Dose reconstructors should refer to the coworker data in Attachment B to assign intakes for 

unmonitored workers. Attachment C provides guidance on assigning the 50th- or 95th-percentile 

coworker intakes based on job category, and level of contact with the radioactive material.

Assessment of Intake for Unmonitored Workers

5.3.6

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.

Unmonitored Intakes of LLNL Radionuclides Other than Primary Radionuclides

Actinium-227 was used in Building 251. Building 251 conducted basic research in the chemical and 

nuclear behavior of the transuranic elements starting in 1955. Building 251 was restricted to “an as 

needed basis” (LLL 1980). Given the nature of the work conducted, access and radiological controls, 

and an in vitro and in vivo bioassay monitoring program in place for workers who worked 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 1981). Given the containment and 

restrictive use controls, 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 from an operational mode to standby mode 

in 1995. During the standby mode 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 (LLNL 2006a). 

Thorium-232 (natural thorium) was associated primarily with Site 300, and two facilities, Buildings 231 

and 321C. Building 231 is a large experimental, manufacturing, assembly, test, and materialshandling facility, and Building 321C machined depleted uranium (DU) and thorium. Machining only 

occurred in 321C. Site 300 was a high-explosive test facility that opened in mid-1955. Depleted 

uranium, natural uranium, 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 conducted, access and radiological controls, and an in vitro and in vivo

bioassay monitoring program in place for workers who worked 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. 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 33 of 107

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 of 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 maximum doses can be assigned when 

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, any of the radionuclides associated with this building listed 

above, should not be assigned after 1995. 

The following table 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 coworker intakes described in the table below will not be assessed for the period prior to 

1957 for plutonium and americium, and prior to 1958 for uranium, as they are included in the 

designated SEC period.

Table 5-9. Radionuclides other than primary radionuclides. 

Radionuclide Building

Likely exposed 

individuals

Period of potential 

exposure

Dose estimation 

techniques 

(in order of preference)

Actinium-227 251 Chemists, researchers, 

health physics technicians, 

technicians, and 

maintenance

1978 - 1995 1. Base on bioassay 

results.

2. Assign dose based on 

coworker intakes for 

Pu-239 assuming Ac227 solubility types F, 

M, and S.

Neptunium237251, 281 Chemists, researchers, 

health physics technicians, 

technicians, and 

maintenance

1955 – 1995 (Building 

251)

1952 - present (Building 

281) 

1. Base on bioassay 

results.

2. Assign dose based on 

coworker intakes for 

Pu-239 assuming Np237 solubility type M.

Curium-244 251, 332 Chemists, researchers, 

health physics technicians, 

technicians, and 

maintenance

1955 – 1995 (Building 

251)

1961 - present (Building 

332 

1. Base on bioassay 

results.

2. Assign dose based on 

coworker intakes for 

Am-241 assuming Cm244 solubility type M.

Thorium-232

(natural 

thorium)

231, 

321C, 

Site 300

Chemists, researchers, 

machinists, health physics 

technicians, and 

technicians

1952 – present (Buildings

231, 321C) 

1955 – present (Site 300)

1. Base on bioassay

results.

2. Assign dose assuming 

a natural thorium mass 

intake equal to the 

coworker uranium 

mass intake. Assume 

a natural thorium 

specific activity of 

4.88x102 dpm/mg to 

determine the total 

thorium activity intake. 

Assume 50 percent of 

the total thorium 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 34 of 107

activity intake is Th232 and 50 percent of 

the total thorium 

activity intake is Th228. Assume solubility 

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 5-10 as input to an incident-specific assessment. This list is not allencompassing, 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.

Table 5-10. Input parameters for significant incidents. 

Incident date Incident description Facility Other information

March 26, 1963 Criticality Bldg. 110 Potential for internal and external exposure (I, Kr, 

Xe); monitored participants listed by name in 1963 

reference (see below).

November 6, 

1975

Radium contamination Bldg. 343, 

Rm. 1005

Crystal was removed from radium dial, resulting in 

spread of contamination.

November 23, 

1971

Pu-contaminated squib 

valve

Bldg. 343 Direct and indirect bioassays performed on 14 

individuals

Sources: LRL (ca. 1963a,b); Leahy (1975); AEC (1971).

5.5 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations to assist in the process of worker dose 

reconstruction. These callouts are listed in this section with information that identifies the source and 

justification for each item. Conventional references are provided in the next section that link data, 

quotations, and other information to documents available for review on the ORAU Team servers.

Thomas LaBone served as a Site Expert for this document and he performed the intake modeling. 

Mr. LaBone was previously employed by the Savannah River Site and he performed a radiological 

assessment at LLNL that involved management, direction, or implementation of radiation protection 

and/or health physics program policies, procedures or practices related to atomic weapons activities 

at the site. This revision has been overseen by a Document Owner who is fully responsible for the 

content, including all findings and conclusions. In all cases where information or prior studies or 

writings are included or relied upon by Mr. LaBone, those materials are fully attributed to the source. 

Mr. LaBone's Disclosure Statement is available at http://www.oraucoc.org. 

[1] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. March 2007. 

The LLNL Technical Basis Manual for Internal Dosimetry (Mansfield 2000) discusses recycled 

uranium in the DOE complex, so the internal dosimetrists at LLNL are aware of the issues.

[2] Thomas, Bill R. Integrated Environmental Management. Health Physicist. March 2007. 

Results for gross alpha in urine were reviewed in the claimant files. The frequency of sample 

collection was dependant on the particular building where the claimant was assigned. The 

frequency ranged from samples several times per year to a single sample per year.
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 35 of 107

[3] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. March 2007. 

In Bihl (2006), when asked if the procedures were in place from 1957 to 1996, I. Dupzyk 

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] Szalinski, Paul A. Integrated Environmental Management. 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.

[5] Thomas, Bill R. Integrated Environmental Management. Health Physicist. March 2007. 

The information in this paragraph comes from Explanatory Notes on “Gross Alpha” Results in 

LLNL claimant files (Author unknown 2006) for Summary of Bioassay Results that were 

reported after approximately 2002. 

[6] Bihl, Donald E. BPNNL. Principal Health Physicist. April 2007.

Personal communication with M. H. Chew, October 16, 2006.

[7] LaBone, Thomas R. ORAU Team. Deputy Principal Internal Dosimetrist. 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 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 Dosimetrist. 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 Dosimetrist. 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. BPNNL. Principal Health Physicist. February 2007.

The minimum geometric standard deviation (GSD) of 3 is established in ORAUT-OTIB-0060, 

Internal Dose Reconstruction (ORAUT 2007c). It reflects the overall uncertainty associated 

with biokinetic modeling as well as usual radiochemical analysis, and indicates that even 

though the spread in coworker 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 Dosimetrist. 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 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 36 of 107

unknown dates. See, for instance, ANSI/HPS-N13.39-2001, Design of Internal Dosimetry 

Programs (HPS 2001). 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 37 of 107

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Sundsmo, M., and D. Hickman, 2009a, collection of lung counter data for 1996 to 1999, Lawrence 

Livermore National Security, Lawrence Livermore National Laboratory, Livermore, California. 

[SRDB Ref ID: 72493]

Sundsmo, M., and D. Hickman, 2009b, collection of minimum detectable activity data for 2005, 

Lawrence Livermore National Security, Lawrence Livermore National Laboratory, Livermore, 

California. [SRDB Ref ID: 72505]

Sundsmo, M., and D. Hickman, 2009c, collection of minimum detectable activity data for 2006, 

Lawrence Livermore National Security, Lawrence Livermore National Laboratory, Livermore, 

California. [SRDB Ref ID: 72506]

Sundsmo, M., and D. Hickman, 2009d, collection of lung counter data for 2002 to 2003, Lawrence 

Livermore National Security, Lawrence Livermore National Laboratory, Livermore, California. 

[SRDB Ref ID: 72498]
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 43 of 107

Sundsmo, M., and D. Hickman, 2009e, collection of minimum detectable activity data for 2007, 

Lawrence Livermore National Security, Lawrence Livermore National Laboratory, Livermore, 

California. [SRDB Ref ID: 72506]

Szalinski, P., undated, Explanatory Notes from Summary of Bioassay Results from Individual 

Dosimetry Records, Lawrence Livermore National Security, Lawrence Livermore National 

Laboratory, Livermore, California. [SRDB Ref ID: 31513]

Surano, K. A., A. H. Bierrnann, R. J. Harrach, F. J. Gouveia, R. L. Berger, B. C. Fields, G. M. 

Gallegos, and P. J. Tate, 1995, LLNL NESHAPs 1994 Annual Report, UCRL-ID-113867-95, 

University of California, Lawrence Livermore National Laboratory, Livermore, California, June. 

[SRDB Ref ID: 23053]

Wilson, R. L., 1982, "Establishing Criteria for Routine Uranium Bioassay in Building 231," 

memorandum to Distribution, University of California, Lawrence Livermore National 

Laboratory, Livermore, California, November 2. [SRDB Ref ID: 15115]
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 44 of 107

GLOSSARY

absorption type 

Categories for materials according to their rate 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 retained or retained 

relatively long the respiratory tract (slow solubilization). Also called solubility type. 

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. 

acute exposure 

Radiation exposure to the body delivered in a short period. See chronic exposure. 

air sampling 

Collection of samples of the ambient atmosphere to detect or measure the presence of 

radioactive material in the air.

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.

beta radiation 

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

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

beta particle is a positron.

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. 

bremsstrahlung 

Electromagnetic radiation released as a result of inelastic scattering of a moving charged 

particle within the nucleus of an atom. X-rays produced in a typical medical X-ray tube 

frequently originate from inelastic scattering of accelerated electrons in the anode material. 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 45 of 107

chronic exposure 

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years). 

See acute exposure.

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. There can be multiple claims filed per worker case.

claimant 

Individual who has filed for compensation under the Energy Employees Occupational Illness 

Compensation Program. This individual can be the energy employee (worker), a survivor, or 

the legal representative of the energy employee.

cohort 

Group of individuals selected for inclusion in a study. 

contamination 

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 x 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

depleted uranium (DU) 

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. 

dose 

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays. 

dose equivalent (H)

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dosimeter 

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. 

electron-volt (eV) 

Unit equal to the energy of one electron moving through a potential difference of 1 volt 

(1.602 × 10-19 joules). The common units in nuclear physics and radiology are kiloelectronvolts (thousands) and megaelectron-volts (millions).

Energy Employees Occupational Illness Compensation Program Act of 2000, as 

amended (EEOICPA; 42 U.S.C. § 7384 et seq.) 

Law that provides for evaluation of cause and potential compensation for energy employees 

who have certain types of cancer.
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 46 of 107

exchange period 

Period (weekly, biweekly, monthly, etc.) for routine exchange of dosimeters. Also called 

exchange frequency.

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.

Integrated Modules for Bioassay Analysis (IMBA) 

Computer program that uses bioassay results and other information to calculate intakes of 

radionuclides and subsequent doses.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds. 

internal dose 

Dose received from radioactive material in the body.

internal dose assessment

Estimation of an intake of radioactive material and the consequent radiation dose based on 

bioassay or other measurements in the work environment. 

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.
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 47 of 107

in vivo bioassay

The measurements of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body. 

ionizing radiation 

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons.

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.

lung solubility type (F, M, or S)

See absorption type.

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. 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 (U, U-nat, 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 microcurie

per gram). See uranium.

occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 48 of 107

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

probability of causation (POC, PC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Act, the percent likelihood, at the 99th percentile, that a worker incurred a 

particular cancer from occupational exposure to radiation.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation.

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 average American receives 360 millirem a year from background radiation. 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.

routine monitoring

Monitoring carried out at regular intervals during normal operations. See special monitoring.

sievert

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

special monitoring

Monitoring in addition to the routine monitoring program carried out for special purposes such 

as estimating the amount of radionuclide deposited in a person after a known or suspected 

accidental intake or after a known or suspected environmental release. See routine 

monitoring.

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. 

whole-body exposure 

Exposure of the entire body to radiation rather than an isolated part. When an ingested 

radioisotope is uniformly distributed throughout the body tissues rather than being 

concentrated in certain parts, the irradiation is a whole-body exposure.

X-ray radiation 

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom. 
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 49 of 107

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 1 of 16

Table A-1. Environmental releases by building (1993–2001).

a

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

131 M&E Divisions Th-232 2.29E-10

U-234 8.47E-02

U-235 1.18E-02

U-238 9.04E-01

132 Analytical & 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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 2 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

151 Isotope Sciences, Environmental Services Laboratory 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

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 U-234 4.38E-01

U-235 1.24E-01

U-238 4.38E-01

212 Physics and Space (rotating target neutron source) H-3 1

222 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 Pu-238 1.09E-02

Pu-239 8.78E-01

Am-241 1.25E-04

Am-243 1.11E-01

226 H-3 1.00E+00

U-238 9.67E-06

227 U-234 5.05E-01

U-235 2.26E-02

U-238 4.73E-01
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 3 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

231 Safeguards and engineering Th-232 1.3E-02

U-234 8.5E-02

U-235 1.2E-02

U-238 8.9E-01

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

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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 4 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

253 Pu-241 6.5E-04

Pu-242 9.9E-10

Am-241 3.1E-06

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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 5 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

255 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

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 rotating target neutron source 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

321A Milling and shaping U-234 1.8E-01

U-235 2.4E-02

U-238 7.9E-01
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 54 of 107

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 6 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

321C Milling, machining, and shaping U-234 6.5E-01

U-235 8.1E-03

U-238 3.4E-01

322 M&E U-234 8.1E-01

U-235 1.2E-02

U-238 1.8E-01

327 M&E 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

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 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
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 55 of 107

ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 7 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

378 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 H-3 1

391 H-3 1

412W Ni-59 3.50E-07

Ni-63 1.00E+00

446 C-14 1

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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 8 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

513 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

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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 9 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

514 Waste processing 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

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 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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 10 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

514 Tank Farm 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

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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 11 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

514 Tank Farm 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

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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 12 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

612 Waste storage and repackaging 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

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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 13 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

612 Waste storage and repackaging 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

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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 14 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

612 Waste storage and repackaging Am-244 4.45E-06

Cm-244 1.05E-04

Cf-249 3.39E-06

Cf-250 2.24E-12

612 Yard 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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 15 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

612 Yard 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
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ATTACHMENT A

ENVIRONMENTAL RELEASES BY BUILDING (1993–2001)

Page 16 of 16

(Current) 

building no. Building activity Radionuclide

Activity 

fraction

625 Waste operations Cm-243 9.78E-09

Cm-244 2.34E-09

Cm-245 1.38E-09

2561 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 H-3 1

Site 300 Pit 7 H-3 1

Site 300-810A U-234 8.38E-02

U-235 1.16E-02

U-238 9.05E-01

Site 300-810B U-234 8.47E-02

U-235 1.13E-02

U-238 9.04E-01

Site 300-850 H-3 9.6E-01

U-234 3.2E-03

U-235 4.4E-04

U-238 3.4E-02

Site 300-851 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

H-3 1

a. This table was derived from data reported for calendar years 1992 through 2001 as part of the National Emission 

Standards for Hazardous Air Pollutants (NESHAPs) (40 CFR Part 61, Subpart H) reporting process (Harrach et al. 1994; 

Surano et al. 1995; 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 its required reporting under NESHAPs, LLNL must 

monitor emissions by radionuclide and total activity at all building release points. This table lists only buildings for which 

NESHAPs 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 NESHAPs 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 (Berger

2005). 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. LLNL (1987b).
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 65 of 107

TABLE OF CONTENTS

SECTION TITLE

B1. Overview................................................................................................................................

PAGE

65

B2. Limitations..............................................................................................................................69

B3. Discussion of the Datasets..................................................................................................... 69

B3.1 Uranium......................................................................................................................69

B3.2 Gross Alpha, Plutonium, and Americium/Curium ........................................................71

B3.3 Gross Beta and MFP..................................................................................................73

B3.4 Analysis......................................................................................................................75

B4. Intake Modeling......................................................................................................................77

B4.1 Assumptions...............................................................................................................77

B4.2 Bioassay Fitting and Intakes.......................................................................................77

B4.2.1 Uranium ....................................................................................................... 78

B4.2.2 Plutonium and Americium ............................................................................ 79

B4.2.3 Gross Beta/MFP .......................................................................................... 83

B1.

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 coworker results.

OVERVIEW

Bioassay results for LLNL were obtained as a copy of the site’s MAPPER database converted to an 

Excel spreadsheet, along 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 coworker 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 depleted uranium, natural uranium 

or slightly enriched uranium (to 5%), various transuranic radionuclides, activation products, and fission 

products (ORAUT 2005b). The MAPPER database had small numbers of analyses for other 

radionuclides, such as 241Am, curium, and transuranic (TRU) materials; however, the numbers were 

too small for statistical analysis as an individual coworker data set. Generally, LLNL used the gross 

alpha analysis to screen for intakes of TRU radionuclides with follow-up analyses being more specific 

for the radionuclide of concern. 

The urinalysis data chosen for this coworker study because of general applicability and number of 

measurements 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 

ATTACHMENT B

INTERNAL COWORKER

Page 1 of 42
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 66 of 107

results and the MFP results were from the same procedure and should be treated as the same (Bihl 

2006b; Mansfield 2006b).

The statistical analyses of the bioassay data for each radionuclide were performed in accordance with 

the internal dosimetry coworker 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 Integrated Modules for Bioassay Analysis (IMBA) computer 

program 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:

95th-percentile intake = 50th-percentile intake × GSD1.645

The tables listing the 50th- and 95th-percentile intakes are the values that the dose reconstructor 

should use when assigning coworker intakes. The 50th-percentile intakes are assigned as a 

lognormal distribution with the associated GSD listed in each table. The 95th-percentile intakes are 

assigned as constant distribution.

Coworker intakes can be extended out to later years 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. Building 251 transitioned from Category II Nonreactor Nuclear Facility 

in 2002 to a radiological facility in 2005 where equipment and inventory were removed (LLNL 2006a). 

During this transition of Building 251, there were no major contamination incidents, no radiation overexposures (in fact, doses were far lower than dose predictions). Individual and collective doses were 

maintained ALARA. Collective annual whole body doses were at least three times lower than ALARA 

goals and more than 10 times lower than conservation dose projections. Individual annual external 

whole body doses were less than 150 mrem (LLNL 2006b). Based on this information about Building 

251, extending coworker intakes out 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 for the coworker intakes determined. 

Table B-1. Americium coworker sample statistic data. 

Year

Number of 

samples

Number of 

employees Year

Number of 

samples

Number of 

employees

1957 89 78 1979–81 127 90

1958 125 84 1982 117 64

1959 133 83 1983 127 64

1960 88 62 1984 140 65

1961 116 79 1985 157 92

1962 121 61 1986 210 110

1963 98 59 1987 160 96

1964 134 83 1988 141 81

1965 139 78 1989 165 79

1966–67 83 54 1990 173 91

1968 89 58 1991 205 104

1969 106 68 1992 220 119

1970 122 85 1993 319 144

ATTACHMENT A

Page 1 of 42

ATTACHMENT B

INTERNAL COWORKER

Page 2 of 42

ATTACHMENT B

INTERNAL COWORKER

Page 2 of 42
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 67 of 107

1971–72 78 35 1994 325 138

1973–75 51 35 1995 300 130

1976–78 112 75 1996 173 99

The MAPPER database was used for the determination of the coworker study datasets, and not the 

individual claims received by DOL. 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 coworker statistics.

Table B-2. Beta coworker sample statistic data.

Year

Number of 

samples

Number of 

employees Year

Number of 

samples

Number of 

employees

1957–63 50 41 1986 265 137

1964–65 75 60 1987 227 143

1966–73 43 42 1988 225 123

1974–75 112 57 1989 218 103

1976 70 35 1990 250 125

1977 96 46 1991 298 136

1978–79 144 75 1992 287 152

1980–81 148 91 1993 407 179

1982 120 69 1994 357 153

1983 152 79 1995 298 126

1984 200 103 1996 171 98

1985 199 112

Table B-3. Plutonium coworker sample statistic data.

Year

Number of 

samples

Number of 

employees Year

Number of 

samples

Number of 

employees

1957 90 79 1977 169 107

1958 125 84 1978 140 102

1959 133 83 1979 119 76

1960 88 62 1980 118 80

1961 116 79 1981 194 141

1962 121 61 1982 314 185

1963 126 73 1983 340 180

1964 140 87 1984 402 205

1965 154 82 1985 384 214

1966 102 57 1986 464 253

1967 84 57 1987 450 268

1968 112 66 1988 492 285

1969 133 84 1989 529 300

1970 133 87 1990 399 233

1971 182 111 1991 377 212

1972 175 95 1992 443 260

1973 106 75 1993 522 261

1974 113 83 1994 493 244

1975 137 85 1995 447 223

ATTACHMENT B

INTERNAL COWORKER

Page 3 of 42
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 68 of 107

Year

Number of 

samples

Number of 

employees Year

Number of 

samples

Number of 

employees

1976 135 94 1996 252 168

• Sample Dates: Sample dates are used only in grouping results together into data sets. 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 coworker study.

• Analyte: Obvious typographical errors in the analyte were corrected. A list of samples 

numbers associated with likely erroneous results was provided in the cover letter transmitting 

the MAPPER database by the site and those involving 238Pu were not included in the 239Pu 

analysis. Tritium coworker data were not evaluated.

• Results Less Than Zero: The cover letter notes that these obvious problems were corrected 

before sending the database to the ORAU Team.

Table B-4. Uranium coworker sample statistic data.

Year

Number of 

samples

Number of 

employees Year

Number of 

samples

Number of 

employees

1958 74 67 1987 Q2 254 171

1959–61 87 70 1987 Q3 221 139

1962–64 96 44 1987 Q4 224 151

1965 107 88 1988 Q1 273 150

1966 113 90 1988 Q2 256 154

1967 51 31 1988 Q3 236 129

1968 141 87 1988 Q4 175 119

1969 95 92 1989 Q1 175 126

1970 118 101 1989 Q2 195 147

1971 35 33 1989 Q3 188 131

1972 272 172 1989 Q4 230 146

1973 172 138 1990 Q1 211 142

1974 139 120 1990 Q2 192 121

1975 S1 146 134 1990 Q3 210 120

1975 S2 159 128 1990 Q4 238 141

1976 S1 139 136 1991 Q1 298 187

1976 S2 95 92 1991 Q2 286 153

1977 S1 142 133 1991 Q3 284 143

1977 S2 151 122 1991 Q4 214 134

1978 S1 155 148 1992 Q1 313 204

1978 S2 131 120 1992 Q2 326 197

1979 S1 213 149 1992 Q3 279 168

1979 S2 204 163 1992 Q4 278 174

1980 S1 199 108 1993 Q1 332 202

1980 S2 141 109 1993 Q2 304 188

1981 S1 157 126 1993 Q3 324 201

1981 S2 168 131 1993 Q4 316 200

1982 S1 245 163 1994 Q1 247 183

1982 S2 271 194 1994 Q2 152 122

1983 S1 278 195 1994 Q3 148 120

1983 S2 317 219 1994 Q4 120 98

1984 S1 286 191 1995 Q1 189 153

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Year

Number of 

samples

Number of 

employees Year

Number of 

samples

Number of 

employees

1984 S2 230 165 1995 Q2 114 98

1985 S1 219 159 1995 Q3 151 110

1985 S2 213 153 1995 Q4 194 146

1986 S1 238 178 1996 Q1 197 122

1986 S2 257 194 1996 Q2 354 206

1987 Q1 220 153 1996 Q3 182 134

• Result Error: Result error has no impact on a coworker analysis; all results are used as 

reported.

• Units: As noted by LLNL, the errors were largely typographical errors and the correct values 

are apparent.

• Media: There are only eight samples with a media code = R or S; all were assumed to be 

urine samples. The sample with a media code of B was not included in the analyses nor were 

the two samples with a result of 1747.8, 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 coworker analysis.

• Exposure Area: Because there are limited samples available for the site, no analysis by work 

area was performed. This field is therefore not used in the study.

• Result Status: All comments were reviewed and samples indicated to be invalid (e.g., spiked, 

cross-contaminated) were excluded from the analysis.

• Assumptions about Sample Volume: Discussions were held with LLNL personnel, including 

the former and present bioassay laboratory managers, regarding this issue. A favorable to 

claimant resolution was adopted; see the discussion in the coworker study for details.

B2.

NIOSH has determined, with concurrence from the Secretary of HHS, that mixed fission product 

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 before 1974 for workers monitored for some type of radiation exposures under the 

EEOICPA (Leavitt 2008). If monitoring data are available for workers 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. 

LIMITATIONS

B3.

B3.1 URANIUM

DISCUSSION OF THE DATASETS

The MAPPER database contained uranium urinalysis data from 1958 to 1996. Nearly all of the 

results were reported as 238U in units of μg/L. The LLNL cover letter indicates that the results actually 

represent total uranium (Mansfield 2006a). According to Section 5.2, most uranium exposure was to 

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depleted or natural uranium. The AVLIS project converted kilogram quantities of natural uranium to 

slightly enriched uranium, up to 5%, as a demonstration plant. The uranium used for the AVLIS

project came from Fernald. The AVLIS project operated in Buildings 175, 177, and the 482 complex 

from 1973 through 1999. HEU which came from Y-12 was used through the site’s history in only a 

few selected buildings. HEU was used in Building 231, and the Oralloy Shop in Building 321C. There 

was another project involving highly enriched uranium in the 1990s, involving the United States and 

Russian HEU Purchase Agreement. The program monitors the Russian process of converting 

weapons-usable HEU into low enriched uranium (LEU). A Transparency Monitoring Office in Russia

was established in 1996 by DOE and is staffed in part by Livermore 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 

μg/L 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 5-6 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 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, as 

compared to 1200 results used for the coworker 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, 

however, 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 when compared to 

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 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. 

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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

B3.2 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 activity measured 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 activity was separated from its 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 report. 

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. 

In an interview (Bihl 2006b) with the former bioassay laboratory manager, 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 (Bihl 2006b) and this will be assumed here. 

In summary, there are three types of analytical results evaluated here:

• GA – a nonspecific analysis that can potentially report any combination of plutonium, 

americium, neptunium, curium, actinium, and thorium.

• PU239 – an analysis specific for 239Pu. Might or might not have an associated AM/CM result.

• AM/CM – a nonspecific analysis that can potentially report any combination of alpha-emitting 

isotopes of americium and curium. Always has an associated PU239 result.

• AM241 – an analysis that is specific for 241Am. 

And, just to make things clear

ALPHA – an analyte type in MAPPER that could be either GA or AM/CM.

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All PU239 and AM/CM urinalysis results were recorded in units of 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 units of 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 that were below the reported MDA for 

the time. For example, in 1990, the recorded PU239 results descend rather orderly 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.) Thus, 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 throughout (1957–1995) 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.

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 

bioassay laboratory manager, present bioassay laboratory manager, and other LLNL personnel, the 

former manager said that during her time in the laboratory plutonium/alpha split samples were 

reported as “per total sample volume” (Bihl 2006b). The present bioassay laboratory manager 

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; Bihl 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 bioassay laboratory manager was present in the bioassay lab 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. Thus, 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. Thus, all PU239 and ALPHA (AM/CM) results 

from splits reported 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.

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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 

samples for some reason. 

For statistical analysis the data were broken down 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 here, 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 + 1967, 1971 + 1972, 1973 + 1974+ 1975, 1976 +

1977 + 1978, and 1979 + 1980 + 1981. The results of the statistical analysis (the 50th- and 84thpercentile excretion rates) were assigned to the midpoint of the analysis interval, which is July 1 for an 

annual interval.

B3.3 GROSS BETA AND MFP

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 (Miller 1979; Bihl 2006b). As stated above, the gross beta and MFP 

results in MAPPER were treated as a single dataset. The gross beta procedure consisted of wetashing 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 

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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/day. 

Duplicate entries that were obvious and samples marked by LLNL as baselines or quality control 

samples were excluded. No samples were excluded due to known large acute intakes. There was a 

single high outlier in the dataset in 1985, which 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 on the gross beta results resulted in poor fits and large geometric 

standard deviations (GSDs) for many of the years. This situation appeared to be due in part because 

there was 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/day 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. 

The entire dataset for 1996 was suspect. The 1996 dataset had a definite low bias, with negative 

numbers running into the thousands and with a total of 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 radiochemist; his reply was, “The LLNL procedure 

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will recover and measure nearly everything except 40K, 137Cs, 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 are 

not completely recovered by this method; they include cobalt, nickel, copper, zinc, and silver. Nickel 

is probably not recovered at all. Iron should be recovered and measured completely. Ruthenium is 

probably mostly recovered by this method” (Soderquist 2007).

B3.4 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 2006b). The 50th- and 84th-percentile values 

were calculated using the method described in ORAUT (2006a). Table B-7 shows the statistical 

analysis results for uranium.

Table B-7. Uranium urine bioassay data for 50th and 84th percentiles in units of μg/L and mg/day 

along with the period and effective bioassay date, 1958 through 1996.

Period

Effective 

bioassay date

50tha

(μg/L)

84thb

(μg/L)

50tha

(mg/day)

84thb

(mg/day)

Employees 

monitored

1958 6/30/1958 6.89E+00 1.39E+01 9.64E-03 1.95E-02 67

1959–61 6/30/1960 6.13E+00 1.48E+01 8.58E-03 2.07E-02 70

1962–64 6/30/1963 5.77E+00 1.35E+01 8.08E-03 1.89E-02 44

1965 6/30/1965 7.61E+00 1.54E+01 1.07E-02 2.16E-02 88

1966 6/30/1966 1.62E+00 6.24E+00 2.26E-03 8.74E-03 90

1967 6/30/1967 6.91E+00 1.40E+01 9.67E-03 1.96E-02 31

1968 6/30/1968 5.42E+00 9.34E+00 7.58E-03 1.31E-02 87

1969 6/30/1969 3.68E+00 7.73E+00 5.15E-03 1.08E-02 92

1970 6/30/1970 1.91E+00 5.41E+00 2.68E-03 7.58E-03 101

1971 6/30/1971 9.16E-01 3.89E+00 1.28E-03 5.45E-03 33

1972 6/30/1972 8.32E+00 2.03E+01 1.16E-02 2.84E-02 172

1973 6/30/1973 4.31E+00 1.33E+01 6.04E-03 1.87E-02 138

1974 6/30/1974 4.57E-02 2.61E-01 6.41E-05 3.66E-04 120

1975 S1c 3/31/1975 6.25E-02 1.87E-01 8.75E-05 2.62E-04 134

1975 S2c 9/30/1975 7.56E-02 2.95E-01 1.06E-04 4.13E-04 128

1976 S1c 3/31/1976 3.08E-02 2.09E-01 4.32E-05 2.93E-04 136

1976 S2c 9/30/1976 2.75E-02 1.79E-01 3.86E-05 2.50E-04 92

1977 S1c 3/31/1977 2.11E-02 1.03E-01 2.95E-05 1.44E-04 133

1977 S2c 9/30/1977 2.42E-02 1.90E-01 3.39E-05 2.66E-04 122

1978 S1c 3/31/1978 3.27E-02 1.65E-01 4.58E-05 2.32E-04 148

1978 S2c 9/30/1978 3.32E-02 1.72E-01 4.64E-05 2.40E-04 120

1979 S1c 3/31/1979 2.14E-02 1.07E-01 2.99E-05 1.50E-04 149

1979 S2c 9/30/1979 3.71E-02 1.74E-01 5.19E-05 2.44E-04 163

1980 S1c 3/31/1980 2.48E-02 2.75E-01 3.47E-05 3.85E-04 108

1980 S2c 9/30/1980 8.37E-03 6.44E-02 1.17E-05 9.02E-05 109

1981 S1c 3/31/1981 9.34E-03 8.28E-02 1.31E-05 1.16E-04 126

1981 S2c 9/30/1981 8.42E-02 3.66E-01 1.18E-04 5.12E-04 131

1982 S1c 3/31/1982 3.96E-02 1.98E-01 5.55E-05 2.77E-04 163

1982 S2c 9/30/1982 8.18E-02 3.61E-01 1.15E-04 5.06E-04 194

1983 S1c 3/31/1983 2.84E-02 2.23E-01 3.98E-05 3.12E-04 195

1983 S2c 9/30/1983 3.69E-02 2.04E-01 5.17E-05 2.85E-04 219

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Period

Effective 

bioassay date

50tha

(μg/L)

84thb

(μg/L)

50tha

(mg/day)

84thb

(mg/day)

Employees 

monitored

1984 S1c 3/31/1984 2.78E-02 1.55E-01 3.90E-05 2.17E-04 191

1984 S2c 9/30/1984 4.76E-02 1.59E-01 6.66E-05 2.22E-04 165

1985 S1c 3/31/1985 3.33E-02 2.18E-01 4.66E-05 3.05E-04 159

1985 S2c 9/30/1985 2.86E-02 1.10E-01 4.00E-05 1.54E-04 153

1986 S1c 3/31/1986 2.01E-02 1.37E-01 2.82E-05 1.92E-04 178

1986 S2c 9/30/1986 5.12E-03 5.04E-02 7.16E-06 7.05E-05 194

1987 Q1c 2/15/1987 6.01E-03 5.42E-02 8.41E-06 7.59E-05 153

1987 Q2c 5/15/1987 1.12E-02 6.62E-02 1.56E-05 9.27E-05 171

1987 Q3c 8/15/1987 2.04E-02 1.05E-01 2.85E-05 1.47E-04 139

1987 Q4c 11/15/1987 1.90E-02 1.23E-01 2.66E-05 1.72E-04 151

1988 Q1c 2/15/1988 1.38E-02 1.15E-01 1.93E-05 1.61E-04 150

1988 Q2c 5/15/1988 6.78E-02 1.58E-01 9.49E-05 2.22E-04 154

1988 Q3c 8/15/1988 6.88E-02 1.66E-01 9.64E-05 2.32E-04 129

1988 Q4c 11/15/1988 1.04E-01 2.75E-01 1.45E-04 3.84E-04 119

1989 Q1c 2/15/1989 7.95E-02 1.88E-01 1.11E-04 2.63E-04 126

1989 Q2c 5/15/1989 1.04E-01 2.58E-01 1.46E-04 3.61E-04 147

1989 Q3c 8/15/1989 7.69E-02 2.16E-01 1.08E-04 3.02E-04 131

1989 Q4c 11/15/1989 1.04E-01 2.83E-01 1.46E-04 3.97E-04 146

1990 Q1c 2/15/1990 5.56E-02 1.42E-01 7.78E-05 1.99E-04 142

1990 Q2c 5/15/1990 4.67E-02 1.19E-01 6.54E-05 1.67E-04 121

1990 Q3c 8/15/1990 4.64E-02 1.02E-01 6.49E-05 1.42E-04 120

1990 Q4c 11/15/1990 5.33E-02 8.95E-02 7.46E-05 1.25E-04 141

1991 Q1c 2/15/1991 7.23E-02 1.35E-01 1.01E-04 1.89E-04 187

1991 Q2c 5/15/1991 8.44E-02 1.59E-01 1.18E-04 2.22E-04 153

1991 Q3c 8/15/1991 6.28E-02 1.11E-01 8.79E-05 1.55E-04 143

1991 Q4c 11/15/1991 3.92E-02 7.68E-02 5.48E-05 1.08E-04 134

1992 Q1c 2/15/1992 3.84E-02 7.15E-02 5.37E-05 1.00E-04 204

1992 Q2c 5/15/1992 3.67E-02 6.04E-02 5.13E-05 8.45E-05 197

1992 Q3c 8/15/1992 4.19E-02 7.38E-02 5.86E-05 1.03E-04 168

1992 Q4c 11/15/1992 3.11E-02 4.56E-02 4.35E-05 6.38E-05 174

1993 Q1c 2/15/1993 3.58E-02 5.45E-02 5.01E-05 7.63E-05 202

1993 Q2c 5/15/1993 4.38E-02 6.67E-02 6.14E-05 9.34E-05 188

1993 Q3c 8/15/1993 4.34E-02 6.85E-02 6.07E-05 9.58E-05 201

1993 Q4c 11/15/1993 5.26E-02 7.10E-02 7.36E-05 9.94E-05 200

1994 Q1c 2/15/1994 4.76E-02 6.45E-02 6.67E-05 9.03E-05 183

1994 Q2c 5/15/1994 4.32E-02 6.54E-02 6.05E-05 9.16E-05 122

1994 Q3c 8/15/1994 5.07E-02 7.46E-02 7.09E-05 1.04E-04 120

1994 Q4c 11/15/1994 5.34E-02 7.86E-02 7.47E-05 1.10E-04 98

1995 Q1c 2/15/1995 6.47E-02 9.70E-02 9.06E-05 1.36E-04 153

1995 Q2c 5/15/1995 5.13E-02 7.84E-02 7.18E-05 1.10E-04 98

1995 Q3c 8/15/1995 4.63E-02 6.61E-02 6.48E-05 9.26E-05 110

1995 Q4c 11/15/1995 4.71E-02 7.43E-02 6.59E-05 1.04E-04 146

1996 Q1c 2/15/1996 5.24E-02 9.03E-02 7.34E-05 1.26E-04 122

1996 Q2c 5/15/1996 4.86E-02 8.39E-02 6.81E-05 1.18E-04 206

1996 Q3c 8/15/1996 5.53E-02 8.90E-02 7.75E-05 1.25E-04 134

a. 50th-percentile value of the fitted line.

b. 84th-percentile value of the fitted line.
Document No. ORAUT-TKBS-0035-5 Revision No. 02 Effective Date: 12/13/2010 Page 77 of 107

c. S1 means first 6 months of the year, S2 means second 6 months of the year, Q1 means months 1-3 of the year, Q2 

means months 4-6 of the year, Q3 means months 7-9 of the year, Q4 means months 10-12 of the year.

B4.

B4.1 ASSUMPTIONS

INTAKE MODELING

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 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 coworker data. Therefore, a GSD of 3 was assigned for any intake period in which

the calculated GSD was <3 [10]. 

B4.2 BIOASSAY FITTING AND INTAKES

The IMBA Expert ORAU-Edition, Version 4.0.9 computer program was 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 change in the results. The effective bioassay dates 

used in IMBA to calculate the intake rates were the midpoints of the sampling periods [11]. The 50thand 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 radionuclide or group of 

radionuclides; therefore, the bioassay results were entered into IMBA with assumed lung absorption 

types chosen to be consistent with International Commission on Radiological Protection (ICRP) 

Publication 68 (ICRP 1995). The plots at the end of this attachment show the resultant 50thpercentile intakes. The bioassay data 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 in only one period and a potential overestimate if an individual worked in multiple 

periods.

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B4.2.1

The 50th- and 84th-percentile uranium urinary excretion concentrations in μg/L were converted to 

mg/day assuming a daily urine volume of 1.4 L/d (ICRP 1975). These data are shown in Table B-8. 

The intakes were determined from the mg/d excretion values. For ease of use by dose 

reconstructors, the intakes were also converted to activity using the specific activity of natural uranium 

(682 pCi/mg), which is favorable to claimants relative to using the specific activity of depleted 

uranium. While workers at LLNL might not have been chronically exposed to natural uranium, 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; note that all figures are at the end 

of this attachment. 

Uranium

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 coworker uranium intakes to be assigned for dose reconstruction.

Table B-8. Uranium mass and activity intake rates for four different periods.

a

Type Dates

50th

percentile

(mg/d)

84th

percentile

(mg/d)

50th

percentile

(pCi/d)

84th

percentile

(pCi/d) GSD

F 1/1/58–12/31/73 2.51E-02 5.81E-02 1.71E+01 3.96E+01 2.32a

F 1/1/74–12/31/87 1.65E-04 8.52E-04 1.13E-01 5.81E-01 5.16

F 1/1/88–12/31/96 2.87E-04 3.73E-04 1.96E-01 2.54E-01 1.30a

M 1/1/58–12/31/73 1.03E-01 2.39E-01 7.04E+01 1.65E+02 2.32a

M 1/1/74–12/31/87 6.76E-04 3.49E-03 4.61E-01 2.38E+00 5.16

M 1/1/88–12/31/96 1.20E-03 1.582E-03 8.16E-01 1.08E+00 1.32a

S 1/1/58–12/31/73 1.24E+00 2.93E+00 8.49E+02 1.99E+03 2.35a

S 1/1/74–12/31/87 8.55E-03 4.19E-02 5.83E+00 2.86E+01 4.90

S 1/1/88–12/31/96 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 four 

different periods.

a

Type Dates

50th

percentile

(mg/d)

50th

percentile

(pCi/d) GSD

95th

percentile

(mg/day)

95th

percentile

(pCi/d)

F 1/1/58–12/31/73 2.51E-02 1.71E+01 3.00 1.53E-01 1.04E+02

F 1/1/74–12/31/87 1.65E-04 1.13E-01 5.16 2.45E-03 1.68E+00

F 1/1/88–12/31/96 2.87E-04 1.96E-01 3.00 1.75E-03 1.19E+00

M 1/1/58–12/31/73 1.03E-01 7.04E+01 3.00 6.28E-01 4.29E+02

M 1/1/74–12/31/87 6.76E-04 4.61E-01 5.16 1.01E-02 6.86E+00

M 1/1/88–12/31/96 1.20E-03 8.16E-01 3.00 7.31E-03 4.97E+00

S 1/1/58–12/31/73 1.24E+00 8.49E+02 3.00 7.56E+00 5.17E+03

S 1/1/74–12/31/87 8.55E-03 5.83E+00 4.90 1.17E-01 7.96E+01

S 1/1/88–12/31/96 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. 

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 

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considered unmonitored. Therefore, unmonitored uranium doses will not be assessed for the period 

before 1958, which is included in the designated SEC period. 

B4.2.2

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-47. The 50th- and 84thpercentile intakes based on urinary excretion modeled as 239Pu are summarized in Tables B-11 and 

B-12. Either absorption type M or S 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 and B-12 can be 

considered pure 239Pu; that is, other isotopes in a plutonium mixture should be based on isotopic 

ratios from Table 5-1. Tables B-13 and B-14 list the 50th- and 95th-percentile coworker plutonium 

intakes to be assigned for dose reconstruction.

Plutonium and Americium 

Table B-10. Fiftieth- and 84th-percentiles for 239Pua urine 

bioassay data for 1957 through 1996. 

Year

Effective 

bioassay 

date

50th-percentile 

excretion rate 

(dpm/d)

84th-percentile 

excretion rate 

(dpm/d)

1957 7/1/1957 4.470E-02 2.829E-01

1958 7/1/1958 9.435E-02 1.002E+00

1959 7/1/1959 1.362E-01 1.361E+00

1960 7/1/1960 7.974E-02 3.566E-01

1961 7/1/1961 3.559E-02 4.136E-01

1962 7/1/1962 4.280E-02 1.113E+00

1963 7/1/1963 2.858E-02 2.956E-01

1964 7/1/1964 3.106E-02 1.133E-01

1965 7/1/1965 1.571E-02 1.547E-01

1966 7/1/1966 1.007E-02 1.547E-01

1967 7/1/1967 6.051E-03 3.878E-02

1968 7/1/1968 4.899E-03 2.202E-02

1969 7/1/1969 1.462E-03 1.526E-02

1970 7/1/1970 2.788E-04 6.192E-03

1971 7/1/1971 2.427E-03 2.898E-02

1972 7/1/1972 3.008E-03 2.499E-02

1973 7/1/1973 4.306E-03 3.266E-02

1974 7/1/1974 2.651E-03 2.339E-02

1975 7/1/1975 1.395E-02 9.313E-02

1976 7/1/1976 3.893E-03 5.441E-02

1977 7/1/1977 3.085E-03 2.961E-02

1978 7/1/1978 1.541E-03 3.706E-02

1979 7/1/1979 7.989E-03 4.731E-02

1980 7/1/1980 1.759E-02 1.376E-01

1981 7/1/1981 5.392E-03 3.582E-02

1982 7/1/1982 4.136E-04 7.478E-03

1983 7/1/1983 1.650E-03 1.078E-02

1984 7/1/1984 8.048E-04 9.488E-03

1985 7/1/1985 6.502E-04 7.737E-03

1986 7/1/1986 9.787E-04 1.257E-02

1987 7/1/1987 2.037E-03 9.097E-03

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1988 7/1/1988 1.207E-03 9.843E-03

1989 7/1/1989 3.082E-03 1.816E-02

1990 7/1/1990 1.092E-03 1.479E-02

1991 7/1/1991 2.890E-04 3.935E-03

1992 7/1/1992 1.308E-04 3.465E-03

1993 7/1/1993 2.425E-04 4.860E-03

1994 7/1/1994 2.670E-04 5.789E-03

1995 7/1/1995 2.991E-04 5.973E-03

1996 7/1/1996 1.357E-04 4.863E-03

a. The data for 1970 were excluded from the analysis because they are

anomalously low.

Table B-11. Fiftieth- and 84th-percentile Type M 239Pu intake rates 

from coworker plutonium urine bioassay data for 1957 through 1996. 

From To

50th 

percentile

(dpm/day)

84th 

percentile

(dpm/day) GSD

1/1/1957 12/31/1960 22.97 123.02 5.36

1/1/1961 12/31/1963 8.09 123.02 15.20

1/1/1964 12/31/1964 8.09 26.19 3.24

1/1/1965 12/31/1967 1.88 26.19 13.95

1/1/1968 12/31/1968 1.88 6.32 3.37

1/1/1969 12/31/1981 0.82 6.32 7.69

1/1/1982 12/31/1990 0.23 1.88 8.11

1/1/1991 12/31/1996 0.04 0.98 22.59

Table B-12. Fiftieth- and 84th-percentile Type S 239Pu intake rates 

from coworker plutonium urine bioassay data for 1957 through 1996. 

From To

50th

percentile

(dpm/day)

84th

percentile

(dpm/day) GSD

1/1/1957 12/31/1960 580.54 2,280.00 3.93

1/1/1961 12/31/1963 126.67 2,280.00 18.00

1/1/1964 12/31/1965 126.67 1,203.20 9.50

1/1/1966 12/31/1968 47.40 163.17 3.44

1/1/1969 12/31/1969 13.41 163.17 12.16

1/1/1970 12/31/1981 13.41 120.09 8.95

1/1/1982 12/31/1990 4.24 33.96 8.01

1/1/1991 12/31/1996 0.88 20.58 23.38

Table B-13. 50th- and 95th-percentile Type M 239Pu intake rates from 

coworker plutonium urine bioassay data for 1957 through 1996. 

From To

50th

percentile

(dpm/d)

84th

percentile

(dpm/d) GSD

95th

percentile

(dpm/d)

1/1/1957 12/31/1960 22.97 123.02 5.36 363.61

1/1/1961 12/31/1963 8.09 123.02 15.2 711.35

1/1/1964 12/31/1964 8.09 26.19 3.24 55.95

1/1/1965 12/31/1967 1.88 26.19 13.95 143.54

1/1/1968 12/31/1968 1.88 6.32 3.37 13.87

1/1/1969 12/31/1981 0.82 6.32 7.69 23.51

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1/1/1982 12/31/1990 0.23 1.88 8.11 7.20

1/1/1991 12/31/1996 0.04 0.98 22.59 6.75

For assessing Super S solubility, refer to ORAUT-OTIB-0049, Estimating Doses for Plutonium 

Strongly Retained in the Lung (ORAUT 2010a). 

Table B-15 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-48 through B-54. 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). Thus, using 241Am as the modeled radionuclide is favorable to 

Table B-14. 50th- and 95th-percentile Type S 239Pu intake rates from 

coworker plutonium urine bioassay data for 1957 through 1996. 

From To

50th

percentile

(dpm/d) GSD

95th

percentile

(dpm/d)

1/1/1957 12/31/1960 580.54 3.93 5515.79

1/1/1961 12/31/1963 126.67 18 14709.47

1/1/1964 12/31/1965 126.67 9.5 5140.78

1/1/1966 12/31/1968 47.4 3.44 361.76

1/1/1969 12/31/1969 13.41 12.16 816.85

1/1/1970 12/31/1981 13.41 8.95 493.38

1/1/1982 12/31/1990 4.24 8.01 129.97

1/1/1991 12/31/1996 0.88 23.38 157.12

claimants. The 50th- and 84th-percentile intakes based on urinary excretion modeled as 241Am are 

summarized in Table B-16. Table B-17 lists the 50th- and 95th-percentile coworker americium intakes 

to be assigned for dose reconstruction. The coworker americium intakes should be assigned for 

cases where there was potential for internal exposure from americium or curium. Separated 

americium was used in Building 251 as early as 1955 for tracer studies. Curium was used primarily in 

Buildings 251 around the same time frame as americium, and 332. Americium coworker 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 coworker intakes, as part of a 

weapons grade mixture, should be assigned. 

Table B-15. Fiftieth- and 84th-percentiles for 241Am urine 

bioassay data for 1957 through 1996. 

Year

Effective

bioassay 

date

50th-percentile 

excretion rate

(dpm/d)

84th-percentile 

excretion rate

(dpm/d)

1957 7/1/1957 4.313E-02 2.785E-01

1958 7/1/1958 9.435E-02 1.002E+00

1959 7/1/1959 1.362E-01 1.361E+00

1960 7/1/1960 7.974E-02 3.566E-01

1961 7/1/1961 3.559E-02 4.136E-01

1962 7/1/1962 4.280E-02 1.113E+00

1963 7/1/1963 4.462E-02 4.147E-01

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Year

Effective

bioassay 

date

50th-percentile 

excretion rate

(dpm/d)

84th-percentile 

excretion rate

(dpm/d)

1964 7/1/1964 3.329E-02 1.106E-01

1965 7/1/1965 2.384E-02 1.910E-01

1966-67 12/31/1966 9.702E-03 1.660E-01

1968 7/1/1968 5.216E-03 2.327E-02

1969 7/1/1969 1.179E-03 1.462E-02

1970 7/1/1970 8.473E-05 2.928E-03

1971-72 12/31/1971 2.371E-03 2.889E-02

1973-75 7/1/1974 1.395E-02 1.283E-01

1976-78 7/1/1977 3.564E-02 7.259E-01

1979-81 7/1/1980 2.059E-02 1.807E-01

1982 7/1/1982 4.221E-02 2.041E-01

1983 7/1/1983 1.466E-02 9.062E-02

1984 7/1/1984 3.748E-02 2.872E-01

1985 7/1/1985 2.131E-02 1.345E-01

1986 7/1/1986 5.588E-02 2.278E-01

1987 7/1/1987 7.659E-02 5.604E-01

1988 7/1/1988 4.988E-02 3.161E-01

1989 7/1/1989 2.334E-02 2.421E-01

1990 7/1/1990 3.713E-02 2.369E-01

1991 7/1/1991 1.872E-02 1.024E-01

1992 7/1/1992 5.301E-02 2.260E-01

1993 7/1/1993 1.601E-02 1.010E-01

1994 7/1/1994 2.848E-02 2.101E-01

1995 7/1/1995 3.952E-02 1.900E-01

1996 7/1/1996 7.815E-02 3.557E-01

Table B-16. Fiftieth- and 84th-percentile Type M 241Am intake rates 

from coworker plutonium urine bioassay data for 1957 through 1996. 

From To

50th

percentile

(dpm/d)

84th

percentile

(dpm/d) GSD

1/1/1957 12/31/1960 10.1 43.8 4.36

1/1/1961 12/31/1965 3.48 43.8 12.6

1/1/1966 12/31/1970 0.49 5.72 11.8

1/1/1971 12/31/1996 1.93 12.7 6.54

Table B-17. 50th- and 95th-percentile Type M 241Am intake rates from 

coworker plutonium urine bioassay data for 1957 through 1996. 

From To

50th

percentile

(dpm/d) GSD

95th

percentile

(dpm/d)

1/1/1957 12/31/1960 10.1 4.36 113.84

1/1/1961 12/31/1965 3.48 12.6 224.74

1/1/1966 12/31/1970 0.49 11.8 28.41

1/1/1971 12/31/1996 1.93 6.54 42.38

There were no bioassay results for plutonium or gross alpha before 1957 in the MAPPER database 

and, according to the laboratory manager circa 1958, few or none might have been obtained (Bihl 

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2006b). Therefore, unmonitored plutonium doses will not be assessed for the period before 1958, 

which is included in the designated SEC period. 

B4.2.3

Interpreting gross beta urinalyses requires 

Gross Beta/MFP 

• At least approximate knowledge of the ratio of the various possible fission/activation products 

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.

Reactor Mixtures (1974–1980) 

As shown in Figures B-55 and B-56, the gross beta urine data given in Table B-18 were modeled as 

100% Type F 90Sr. The resulting intakes were then multiplied by 0.024, which is the average fraction 

of 90Sr in urine following intakes of 10-day old reactor fuel in OTIB-0054 (ORAUT 2007b, Table 7-1). 

Table B-19 lists the fiftieth- and 84th-percentile intake rates of Type F 90Sr for 1974 through 1980. 

The 50th- and 95th-percentile 90Sr intake rates, which are given in Table B-20, should be used in 

conjunction with associated radionuclides’ ratios for estimating intakes for relevant radionuclides

based on intake activity relative to 90Sr for 10-day old reactor fuel from Table 7-3 in OTIB-0054. As 

noted previously, the LPTR operated from December of 1957 to March 31, 1980.

Table B-18. Fiftieth- and 84th-percentiles for gross beta 

urine bioassay data for 1974 through 1995. 

Year

Effective

bioassay date

50th-percentile

excretion rate

(dpm/d)

84th-percentile

excretion rate

(dpm/d)

1974–75 12/31/1974 19.22 199.1

1976 7/1/1976 17.74 136.8

1977 7/1/1977 18.13 103.8

1978–79 12/31/1978 1.474 55.76

1980–81 12/31/1980 13.04 165.8

1982 7/1/1982 22.56 236.8

1983 7/1/1983 96.75 636.4

1984 7/1/1984 9.806 99.19

1985 7/1/1985 20.17 145.1

1986 7/1/1986 17.59 203.3

1987 7/1/1987 64.88 395.0

1988 7/1/1988 125.1 777.1

1989 7/1/1989 62.62 544.2

1990 7/1/1990 62.73 455.0

1991 7/1/1991 14.79 145.5

1992 7/1/1992 7.883 76.81

1993 7/1/1993 1.128 19.69

1994 7/1/1994 0.1914 4.956

1995 7/1/1995 0.2789 4.694

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Table B-19. Fiftieth- and 84th-percentile intake rates of Type F 90Sr

for 1974 through 1980. 

From To

50th

percentile

(dpm/d)

84th

percentile

(dpm/d) GSD

1/1/1974 3/31/1980 1.51 14.71 9.75

Table B-20. 50th- and 95th-percentile Type F 90Sr intake rates for 

1974 through 1980. (These intake rates are used in conjunction with 

OTIB-0054 (ORAUT 2007b) to assign intakes of fission product 

mixtures from reactors based on the associated radionuclides ratios 

relative to 90Sr for 10-day old reactor fuel in Table 7-3 in OTIB-0054.) 

From To

50th

percentile

(dpm/d) GSD

95th

percentile

(dpm/d)

1/1/1974 3/31/1980 1.51 9.75 63.96

Weapons Residue (1974–1992)

As shown in Figures B-57 and B-58, the gross beta urinary excretion for 1974 through 1992 was 

modeled 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/dose from 181W. Table B-21 lists the 50th- and 84thpercentile 103Ru intakes. Table B-22 lists the 50th- and 84th-percentile 141Ce, and 89Sr intakes. Table

B-23 lists the 50th- and 95th-percentile 103Ru intakes to be assigned for dose reconstruction. Table B24 lists the 50th- and 95th-percentile 141Ce, and 89Sr intakes to be assigned for dose reconstruction. 

Table B-21. Fiftieth- and 84th-percentile intake rates of Type F 103Ru 

for 1974 through 1992. 

From To

50th

percentile

(dpm/d)

84th

percentile

(dpm/d) GSD

1/1/1974 12/31/1982 62.72 605.52 9.65

1/1/1983 12/31/1983 411.48 2,706.57 6.58

1/1/1984 12/31/1986 62.72 605.52 9.65

1/1/1987 12/31/1987 269.51 1,975.44 7.33

1/1/1988 12/31/1988 531.74 3,304.41 6.21

1/1/1989 12/31/1990 269.51 1,975.44 7.33

1/1/1991 12/31/1992 62.72 605.52 9.65

Aged Fission Products and Research-Related Radionuclides (After 1992) 

As shown in Figures B-59 and B-60, 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-25 lists the 50th- and 84th-percentile 90Sr, 90Y, 32P, 137Cs, and 14C intakes. Table

B-22 lists the 50th and 84th percentile 141Ce, and 89Sr intakes. Tables B-26 lists the 50th- and 95thpercentile 90Sr, 90Y, 32P, 137Cs, and 14C intakes to be assigned for dose reconstruction.

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Table B-22. Fiftieth- and 84th-percentile intake rates of Type S 141Ce

and Type F 89Sr for 1974 through 1992. 

From To

50th

percentile

(dpm/d)

84th

percentile

(dpm/d) GSD

1/1/1974 12/31/1982 31.36 302.76 9.65

1/1/1983 12/31/1983 205.74 1,353.28 6.58

1/1/1984 12/31/1986 31.36 302.76 9.65

1/1/1987 12/31/1987 134.76 987.72 7.33

1/1/1988 12/31/1988 265.87 1,652.21 6.21

1/1/1989 12/31/1990 134.76 987.72 7.33

1/1/1991 12/31/1992 31.36 302.76 9.65

Table B-23. 50th- and 95th-percentile Type F 103Ru intake rates for

1974 through 1992. These intake rates are used to assign intakes of 

fission and activation product mixtures from weapons test residuals.

From To

50th

percentile

(dpm/d) GSD

95th

percentile

(dpm/d)

1/1/1974 12/31/1982 62.72 9.65 2611.88

1/1/1983 12/31/1983 411.48 6.58 9127.05

1/1/1984 12/31/1986 62.72 9.65 2611.88

1/1/1987 12/31/1987 269.51 7.33 7139.55

1/1/1988 12/31/1988 531.74 6.21 10723.46

1/1/1989 12/31/1990 269.51 7.33 7139.55

1/1/1991 12/31/1992 62.72 9.65 2611.88

Table B-24. 50th- and 95th-percentile Type S 141Ce and Type F 89Sr 

intake rates for 1974 through 1992. These intake rates are used to 

assign intakes of fission and activation product 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)

1/1/1974 12/31/1982 31.36 9.65 1305.94

1/1/1983 12/31/1983 205.74 6.58 4563.52

1/1/1984 12/31/1986 31.36 9.65 1305.94

1/1/1987 12/31/1987 134.76 7.33 3569.91

1/1/1988 12/31/1988 265.87 6.21 5361.73

1/1/1989 12/31/1990 134.76 7.33 3569.91

1/1/1991 12/31/1992 31.36 9.65 1305.94

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Table B-25. Fiftieth- and 84th-percentile intake rates of Type F 90Sr, 90Y, 32P, 137Cs, and 14C for 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

1/1/1993 12/31/1993 5.11 89.5 17.5

1/1/1994 12/31/1995 0.909 19.1 20.9

a. Refer to Section 5.3.2.3 for solubility types for Y-90, P-32, Cs-137, and C-14. 

Table B-26. 50th- and 95th-percentile Type F 90Sr, 90Y, 32P, 137Cs, 

and 14C intake rates for 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)

1/1/1993 12/31/1993 5.11 17.5 566.52

1/1/1994 12/31/1995 0.909 20.9 134.96

a. Refer to Section 5.3.2.3 for solubility types for Y-90, P-32, Cs-137, and C-14. 

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). 

Figure B-1. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type F natural uranium from January 1958 through 

December 1973 (blue dots). 

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

inhalation intake of Type F natural uranium from January 1974 through December 1987

(blue dots).

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

inhalation intake of Type F natural uranium from January 1988 through December 1996

(blue dots).

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

separate chronic inhalation intakes of Type F natural uranium.

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Figure B-5. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type F natural uranium from January 1958 through 

December 1973 (blue dots).

Figure B-6. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type F natural uranium from January 1974 through 

December 1987 (blue dots).

Figure B-7. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type F natural uranium from January 1988 through 

December 1996 (blue dots).

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

separate chronic inhalation intakes of Type F natural uranium.

Figure B-9. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M natural uranium from January 1958 through 

December 1973 (blue dots).

Figure B-10. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M natural uranium from January 1974 through 

December 1987 (blue dots).

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Figure B-11. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M natural uranium from January 1988 through 

December 1996 (blue dots).

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

separate chronic inhalation intakes of Type M natural uranium.

Figure B-13. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M natural uranium from January 1958 through 

December 1973 (blue dots).

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Figure B-14. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M natural uranium from January 1974 through 

December 1987 (blue dots).

Figure B-15. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M natural uranium from January 1988 through 

December 1996 (blue dots).

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

separate chronic inhalation intakes of Type M natural uranium (blue dots).

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Figure B-17. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S natural uranium from January 1959 through 

December 1973 (blue dots).

Figure B-18. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S natural uranium from January 1974 through 

December 1987 (blue dots).

Figure B-19. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S natural uranium from January 1988 through 

December 1996 (blue dots).

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

separate chronic inhalation intakes of Type S natural uranium.

Figure B-21. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S natural uranium from January 1958 through 

December 1973 (blue dots).

Figure B-22. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S natural uranium from January 1974 through 

December 1987 (blue dots).

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Figure B-23. Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type S natural uranium from January 1988 through December 1996 

(blue dots).

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

separate chronic inhalation intakes of Type S natural uranium.

Figure B-25. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1957 through December 

1960. 

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Figure B-26. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1961 through December 

1964. 

Figure B-27. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1965 through December 

1968. 

Figure B-28. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1969 through December 

1981. 

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Figure B-29. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1982 through December 

1990. 

Figure B-30. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1957 through December 

1996. 

Figure B-31. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1957 through December 

1963. 

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Figure B-32. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1964 through December 

1967. 

Figure B-33. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1968 through December 

1981. 

Figure B-34. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1982 through December 

1990. 

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Figure B-35. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M plutonium from January 1991 through December 

1996. 

Figure B-36. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1957 through December 

1960. 

Figure B-37. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1961 through December 

1965. 

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Figure B-38. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1966 through December 

1968. 

Figure B-39. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1969 through December 

1981.

Figure B-40. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1982 through December 

1990.

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Figure B-41. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1991 through December 

1996.

Figure B-42. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1957 through December 

1963.

Figure B-43. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1964 through December 

1966.

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Figure B-44. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1967 through December 

1969.

Figure B-45. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1970 through December 

1981.

Figure B-46. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1982 through December 

1990.

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Figure B-47. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type S plutonium from January 1991 through December 

1996.

Figure B-48. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M 241Am from January 1957 through December 

1960. 

Figure B-49. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M 241Am from January 1961 through December 

1965. 

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Figure B-50. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M 241Am from January 1966 through December 

1970. 

Figure B-51. Predicted and observed 50th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M 241Am from January 1973 through December 

1979. 

Figure B-52. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type M 241Am from January 1957 through December 

1965. 

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Figure B-53. Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type M 241Am from January 1966 through December 1970. 

Figure B-54. Predicted and observed 84th-percentile urinary excretion assuming a chronic 

inhalation intake of Type M 241Am from January 1971 through December 1996. 

Figure B-55. Predicted and observed 50th-percentile gross beta urinary excretion assuming a 

chronic inhalation intake of Type F 90Sr from January 1974 through December 1992.

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Figure B-56. Predicted and observed 84th-percentile gross beta urinary excretion 

assuming a chronic inhalation intake of Type F 90Sr from January 1974 through 

December 1992.

Figure B-57. Predicted and observed 50th-percentile gross beta urinary excretion 

assuming a chronic inhalation intake of Type F 103Ru from January 1974 through 

December 1992. 

Figure B-58. Predicted and observed 84th-percentile gross beta urinary excretion 

assuming a chronic inhalation intake of Type F 103Ru from January 1974 through 

December 1992. 

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Figure B-59. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type F 90Sr from January 1993 through December 1995. 

Figure B-60. Predicted and observed 84th-percentile urinary excretion assuming a 

chronic inhalation intake of Type F 90Sr from January 1993 through December 1995.

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Assigned Intakes Using Case-Specific Information

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 are 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 LLNL employees, 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 coworker dose intakes. For personnel with full contact with the material, 

the 95th percentile of the calculated coworker intakes should be assigned. For others, the 50th 

percentile is more appropriate, as listed in Table C-1. 

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

Contact level Description Example job titles

Percentile of 

CW 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.

Assume 

environmental 

internal dose

ATTACHMENT C

SCALING OF COWORKER DOSE INTAKES CONSISTENT 

WITH TYPE AND DURATION OF EXPOSURES