ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 117

DOE Review Release 04/26/2013

Document Title:

Los Alamos National Laboratory –

Occupational Internal Dose

Document Number: ORAUT-TKBS-0010-5

Revision: 02

Effective Date: 04/18/2013

Type of Document: TBD

Supersedes: Revision 01

Subject Expert(s): Donald N. Stewart and Rowena S. Argall

Approval: Signature on File Approval Date: 04/16/2013

Donald N. Stewart, Document Owner

Concurrence: Signature on File Concurrence Date: 04/15/2013

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 04/15/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 04/15/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/18/2013

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 2 of 117

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

12/21/2004 00 New Technical Basis Document for the Los Alamos National 

Laboratory – Occupational Internal Dose. Incorporates formal 

internal and NIOSH review comments. Initiated by Jack E. 

Buddenbaum.

10/15/2009 01 Incorporates information from documents not available at the time 

the original document was published. The current information 

available for uranium has been incorporated into Section 5.2.4. 

Clarification of MDAs, UNAA/DNAA analysis results, guidance for

selection of enrichment, and a description of the new LANL Bioassay 

Repository (LABDR) content have been included. Appropriate 

references have been added. For claims involving uranium UNAA 

bioassay results: Because of the ambiguity in the interpretation of 

the uranium bioassay results when results were transferred into the 

LABDR, the guidance in the TBD might not have been interpreted 

consistently by DRs and PRs. UNAA/DNAA data provided by LANL 

is for DU and EU rather than U-238 and U-235. A table providing an 

aged “pure” 238Pu mixture has been provided for use when exposures 

to heat source technology plutonium is suspected. Changes to the 

mixture might affect the total dose for plutonium intakes evaluated as 

“pure” 238Pu. In addition, the Tiger Team Assessment (DOE 1991), 

the Human Experiments and the details of the RaLa SEC Report 

(NIOSH 2006) have been included in the revision. Incorporates 

formal internal and NIOSH review comments. Adds Attributions and 

Annotations section. A description of the class of employees added 

to the SEC for SEC Petition-00051 was added to the text describing 

the class of employees added under SEC Petition-00061 (employees 

associated with RaLa operations). A new Section 5.2.4.6 was added 

describing the limitations on internal dose reconstruction for 

employees in the non-excepted TAs (as listed in the petition). Added 

statement about only partial dose reconstructions are possible for 

most TAs. A new bullet was added in Section 5.7.2, General 

Guidance, indicating that only partial internal dose reconstructions 

are possible. Incorporates methods for estimating maximum doses 

in accordance with the evaluation report for Petition SEC-00109, 

changes to address comments by OCAS commenters, updated dose 

reconstruction methodologies and references, and addition of an 

algorithm to guide dose reconstructors for dose calculation and dose 

assignment. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by 

Donald N. Stewart.

04/18/2013 02 Revision initiated to incorporate SEC-00109. Removed provisions to 

estimate unmonitored intakes consistent with the wording in the 

Advisory Board recommendation (Melius 2012). Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by Donald N. Stewart.
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 3 of 117

TRADEMARK INFORMATION

Microsoft® and Excel® are registered trademarks of Microsoft Corporation in the United States and/or 

other countries.

All other trademarks are the property of their respective owners.
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TABLE OF CONTENTS

SECTION TITLE PAGE

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

5.1 Introduction ............................................................................................................................ 11

5.1.1 Purpose...................................................................................................................... 12

5.1.2 Scope......................................................................................................................... 12

5.1.3 Special Exposure Cohort............................................................................................ 13

5.1.4 Background of Internal Exposures.............................................................................. 14

5.1.4.1 Early Research with Humans...................................................................... 14

5.1.4.2 Early Safety Efforts..................................................................................... 14

5.1.4.3 Bioassay Program ...................................................................................... 17

5.1.4.3.1 Individual Bioassay Results....................................................... 17

5.1.4.3.2 UPPU Club................................................................................ 18

5.1.5 Unmonitored Workers and Unreported Bioassay Results ........................................... 18

5.1.5.1 Workers Before 1949.................................................................................. 18

5.1.5.2 Short-Term Workers ................................................................................... 19

5.1.5.3 Zia Company and Other Contract Service Workers .................................... 19

5.1.5.4 RaLa Project Operations, 1944 to 1963...................................................... 20

5.1.5.5 Responders to the Cerro Grande Wildlands Fire of 2000............................ 20

5.1.6 Overview of Internal Dose Reconstruction Process for Los Alamos National 

Laboratory .................................................................................................................. 21

5.2 In Vitro Bioassay Minimum Detectable Amounts, Counting Methods, and Reporting 

Protocols for Primary Radionuclides ...................................................................................... 21

5.2.1 Plutonium ................................................................................................................... 22

5.2.1.1 Special Considerations for 238Pu Dosimetry Related to a Wing 9 

Incident....................................................................................................... 22

5.2.1.2 Sample Collection Procedures.................................................................... 23

5.2.1.3 Missed Intakes............................................................................................ 24

5.2.1.4 Routine Sample Frequencies...................................................................... 24

5.2.1.5 Sample Analysis Procedures ...................................................................... 24

5.2.1.6 Plutonium Isotopic Ratios (Mixtures)........................................................... 27

5.2.1.7 Correcting for Urinalysis Volume................................................................. 30

5.2.1.8 Calculating Picocuries per 24 Hours for Urine Bioassay Results................. 30

5.2.1.9 Detection Sensitivities, Reporting Limits, and Tolerance Limits .................. 31

5.2.1.10 Validation of Samples................................................................................. 32

5.2.1.11 Sampling Protocol for Special Accidental Exposure.................................... 32

5.2.1.12 Nasal Swipes.............................................................................................. 33

5.2.1.13 Computer-Based Calculations and Sorting ................................................. 33

5.2.1.14 LANL Bioassay Data Repository................................................................. 33

5.2.2 Americium .................................................................................................................. 34

5.2.3 Tritium ........................................................................................................................ 35

5.2.3.1 Organically Bound Tritium........................................................................... 36

5.2.3.2 Metal Tritides.............................................................................................. 36

5.2.3.3 Analytical Techniques................................................................................. 36

5.2.3.4 Routine Sampling ....................................................................................... 37

5.2.4 Uranium...................................................................................................................... 37

5.2.4.1 Sampling Protocol ...................................................................................... 38

5.2.4.2 Uranium Analysis Techniques .................................................................... 39

5.2.4.3 Delayed Neutron Counting.......................................................................... 40
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5.2.4.4 Radiometric Alpha Spectroscopy ................................................................ 42

5.2.4.5 Excretion of Environmental Levels of Uranium............................................ 42

5.2.5 Polonium .................................................................................................................... 43

5.2.6 Other Limited-Exposure Radionuclides....................................................................... 44

5.3 In Vivo Bioassay Minimum Detectable Activities, Analytical Methods, and Reporting 

Protocols................................................................................................................................ 44

5.3.1 Whole-Body Counters ................................................................................................ 44

5.3.1.1 Minimum Detectable Activities and Decision Levels.................................... 45

5.3.1.2 Cesium-137 and Other Intakes from Fallout ............................................... 48

5.3.2 Lung Burdens ............................................................................................................. 49

5.3.3 Wound Monitoring ...................................................................................................... 51

5.4 Interferences and Uncertainties.............................................................................................. 51

5.4.1 Discrepancies in Results ............................................................................................ 51

5.4.2 Contamination of Samples.......................................................................................... 52

5.4.3 Uncertainties .............................................................................................................. 52

5.4.3.1 In Vivo Counting ......................................................................................... 52

5.4.3.2 In Vitro Measurements................................................................................ 52

5.5 Unmonitored Intakes.............................................................................................................. 53

5.5.1.1 Energy Employee or Survivor Interview ...................................................... 53

5.5.1.2 Initial Case File........................................................................................... 53

5.5.2 Unmonitored Intakes of Primary Radionuclides .......................................................... 53

5.5.2.1 Plutonium ................................................................................................... 54

5.5.2.2 Polonium .................................................................................................... 54

5.5.2.3 Uranium...................................................................................................... 54

5.5.2.4 Tritium ........................................................................................................ 55

5.5.2.5 Americium-241. .......................................................................................... 55

5.5.3 Unmonitored Intakes of LANL Radionuclides other than Primary Radionuclides ........ 55

5.5.3.1 Actinium-227............................................................................................... 56

5.5.3.2 Curium-244................................................................................................. 56

5.5.3.3 Neptunium-237........................................................................................... 56

5.5.3.4 Protactinium-231 ........................................................................................ 57

5.5.3.5 Thorium ...................................................................................................... 57

5.5.3.6 Thorium-230 ............................................................................................... 57

5.5.3.7 Thorium-232 ............................................................................................... 57

5.5.3.8 Mixed Fission and Activation Products........................................................ 58

5.6 Attributions and Annotations .................................................................................................. 58

References ........................................................................................................................................ 59

Glossary ............................................................................................................................................ 71

ATTACHMENT A, OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS ................. 73

ATTACHMENT B, BIOASSAY FOR RADIONUCLIDES OTHER THAN PRIMARY 

RADIONUCLIDES..............................................................................................103

ATTACHMENT C, RESPIRATORY PROTECTION PROGRAM...................................................... 108
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ATTACHMENT D, DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL 

DOSES ..............................................................................................................109

ATTACHMENT E, SCALING OF COWORKER DOSE INTAKES CONSISTENT WITH 

TYPE AND DURATION OF EXPOSURES......................................................... 113

ATTACHMENT F, BOUNDING ESTIMATES OF ORGAN DOSES FROM POTENTIAL 

INHALATIONS OF RADIOACTIVE MATERIALS DURING THE 

CERRO GRANDE FIRE AT LOS ALAMOS ....................................................... 114

LIST OF TABLES

TABLE TITLE PAGE

5-1 Historical in vitro bioassay, primary radionuclides .................................................................. 21

5-2 Historical in vitro bioassay not available in claim records ....................................................... 22

5-3 IMBA input parameters for types M, S, and nonmonotonic type J .......................................... 23

5-4 Activity composition of nominal 3% plutonium mixture ........................................................... 26

5-5 Activity composition of reference weapons-grade 6% plutonium mixture................................ 26

5-6 Activity composition of reference fuel-grade (12%) plutonium mixture.................................... 27

5-7 Plutonium bioassay sensitivity as listed in procedures and reports......................................... 28

5-8 “Pure” 238Pu mixture ............................................................................................................... 29

5-9 Fresh plutonium mixture......................................................................................................... 30

5-10 Americium-241 bioassay techniques and sensitivities ............................................................ 35

5-11 Tritium urine bioassay sensitivity levels.................................................................................. 37

5-12 Nominal weight composition and fractional activities for D-38 uranium................................... 38

5-13 Nominal weight composition and fractional activities for OR-93 uranium................................ 38

5-14 Routine uranium urinalysis detection levels............................................................................ 40

5-15 Factors to convert isotopic activities to total uranium.............................................................. 42

5-16 Characteristics of NU (Tuballoy)............................................................................................. 42

5-17 Routine 210Po urinalysis detection levels ................................................................................ 44

5-18 Routine whole-body counting detection levels........................................................................ 45

5-19 Mean body burdens of 137Cs from fallout in the United States ................................................ 49

5-20 LC and MDA values for lung counting ..................................................................................... 50

5-21 Hypothetical chronic intakes for NU, 1943 to 1946................................................................. 55

A-1 Identification match ................................................................................................................75

A-2 Assessment method codes for bioassay reports .................................................................... 75

A-3 In vitro collection requirements............................................................................................... 75

A-4 In vitro collection protocols ..................................................................................................... 75

A-5 Sample units ..........................................................................................................................75

A-6 In vivo detector systems......................................................................................................... 76

A-7 In vivo collection protocols ..................................................................................................... 76

A-8 Current and historical locations .............................................................................................. 77

A-9 Zia employee access to plutonium areas ............................................................................... 88

A-10 Historical in vitro bioassay...................................................................................................... 89

A-11 Current in vitro bioassay performed, 1997.............................................................................. 89

A-12 Routine sampling procedure .................................................................................................. 90

A-13 Plutonium urinalysis sensitivities and analytical techniques.................................................... 91

A-14 Example urine bioassay results.............................................................................................. 91

A-15 Routine plutonium sampling frequency................................................................................... 94

A-16 Routine sampling frequency 241Am exposures........................................................................ 96

A-17 Sampling protocol for tritium................................................................................................... 96
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A-18 Uranium routine sampling protocol......................................................................................... 96

A-19 Sampling protocol for accidental plutonium exposures, 1967................................................. 96

A-20 Sampling protocol for accidental plutonium exposures, 1973................................................. 97

A-21 Sampling protocol for accidental plutonium exposures, 1998................................................. 98

A-22 Sampling protocol for accidental 241Am exposures ................................................................. 98

A-23 Uranium nonroutine sampling protocol................................................................................... 99

A-24 In vitro bioassay sensitivity..................................................................................................... 99

A-25 In vivo bioassay routine frequency ....................................................................................... 100

A-26 Amount of plutonium administered to subjects by intravenous injection and the 

date of the administration..................................................................................................... 101

B-1 Specific activity of thorium isotopes...................................................................................... 106

C-1 Respiratory protection apparatus ......................................................................................... 108

D-1 Reported exposure incidents and results ............................................................................. 109

D-2 Summary of February 10, 1977, plutonium incident ............................................................. 112

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

F-1 Highest measured air concentrations of alpha emitters and consequent intakes.................. 116

F-2 Calculated doses by radionuclide and organ........................................................................ 117

F-3 Total organ doses for all evaluated radionuclides................................................................. 117
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ACRONYMS AND ABBREVIATIONS

25 235U or the standard ~93% enrichment (LANL designation)

28 238U or depleted uranium (LANL designation)

49 239Pu (LANL designation)

aCi attocurie, 10-18 Ci

AEC U.S. Atomic Energy Commission

AL action level

AMAD activity median aerodynamic diameter

Bq becquerel

CAM continuous air monitor

CFR Code of Federal Regulations

Ci curie

cm centimeter

CMR Chemistry and Metallurgy Research

cpm counts per minute

d day

D-38 depleted uranium mixture

DA delayed action

DAC derived air concentration

DL detection limit

DNAA delayed neutron activation analysis

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DR dose reconstructor

DTPA diethylene triamine pentaacetic acid

dpm disintegrations per minute

dps disintegrations per second

DU depleted uranium (less than 0.72% 235U)

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ESH Environmental, Safety, and Health

ET2 extrathoracic airways (posterior nasal passage, larynx, pharynx, and mouth)

EU enriched uranium (more than 0.72% 235U)

g gram

GeLi lithium-drifted germanium

GM Geiger-Müller

GSD geometric standard deviation

HPGe hyper-pure germanium

hr hour

HRL Health Research Laboratory

HT elemental tritium (tritiated gas)

HTO tritium oxide (water or water vapor)

IBF Ion Beam Facility

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch
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IVML In Vivo Measurements Laboratory (system)

keV kiloelectron-volt, 1,000 electron-volts

L liter

LABDR LANL Bioassay Data Repository

LAMPF Los Alamos Meson Physics Facility

LANL Los Alamos National Laboratory (1981 to present)

LASL Los Alamos Scientific Laboratory (until January 1981)

LC critical level

LN(TH) lymph nodes, thoracic

LSC liquid scintillation counting

L X-ray low-energy X-ray

MAC maximum allowable concentration

MAP mixed activation product

mBq millibecquerel

mCi millicurie

MDA minimum detectable activity

MDTA minimum detectable true activity

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission product

mg milligram

min minute

mL milliliter

mm millimeter

mo month

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPL maximum permissible level

mR milliroentgen

mrem millirem

mrep millirep

MSMA minimum significant measured activity

MT metal tritide

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NDA no detectable activity

NIOSH National Institute for Occupational Safety and Health

NRS nonroutinely sampled

NTA neutron track analysis

NU natural uranium

OBT organically bound tritium

Or-93 oralloy (Oak Ridge alloy); uranium enriched to 40% or 93%

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

OTLBD ORAU Team LANL Bioassay Database

OWR Omega West Reactor

PA prompt action

pCi picocurie
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PDF Portable Document Format

PF Plutonium Facility

PHA pulse height analysis

POC probability of causation

PR peer reviewer

PUQFUA plutonium body burden (Q) from urine analysis

RaLa radioactive lanthanum

RAS radiometric alpha spectroscopy

RBM red bone marrow

RIR Radiological Incident Report

RTG radioisotopic thermoelectric generator

s second

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

TA Technical Area

TBD technical basis document

TIMS thermal ionization mass spectroscopy

TRU transuranic

U.S.C. United States Code

UF uranium fluorophotometric

UNAA uranium analysis using DNAA

UPPU You Pee Plutonium (Club)

UR uranium radiometric (extraction chemistry with radiometric alpha proportional counting)

VFP volatile fission product

WGPu weapons-grade plutonium

wk week

X10 plutonium (LANL designation)

yr year

μBq microbecquerel

μCi microcurie

μg microgram

μm micrometer

§ section or sections
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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

] guidelines established under 

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

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

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

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 2010a):

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

conventional structures

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

reasons

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

The purpose of this technical basis document (TBD) is to provide information about occupational 

internal dose at the Los Alamos National Laboratory (LANL).2

 Occupational internal dose is the dose 

that was received by an individual from an intake of radioactive material while performing tasks in 

LANL buildings and structures or from activities outside buildings, such as burial of waste and 

monitoring of tests, where intakes of radioactive material could occur.

5.1.2 Scope

This document contains information for reconstruction of occupational internal doses at LANL facilities 

throughout its history. LANL has a long history of research and production missions, and there is 

potential for exposures to a diverse set of radioactive materials. Early controls on the use, handling, 

and storage of radioactive material were based on maintenance of intakes below an assumed safe 

exposure or tolerance level, so measurements and recordkeeping practices might not support 

accurate dose reconstructions in all cases. For these reasons and others (see section 5.1.3), three 

classes of LANL employees have been added to the Special Exposure Cohort (SEC) as described in 

Section 5.1.3. The addition of these classes of employees has limited the scope of potential internal 

dose reconstructions at LANL to the following.

• Internal dose from tritium from 1950 to the present;

• Internal dose from polonium from 1944 to 1956;

• Internal dose from plutonium from 1944 to the present;

• Internal dose from uranium from 1943 to the present;

• Internal dose from other radionuclides present at LANL may be reconstructed only in cases 

where internal dose monitoring data may be used for any individual claim, in accordance with 

the NIOSH dose reconstruction processes described in this TBD.

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 LANL 

workers. Sections 5.2 and 5.3 discuss in vitro and in vivo bioassay, respectively. Section 5.4 

describes interferences and uncertainties, and Section 5.5 discusses the treatment of unmonitored 

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

justification, or clarification of the associated information, are presented in Section 5.6.

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

workers. Attachment B discusses bioassay for radionuclides other than primary radionuclides, and 

Attachment C describes the respiratory protection program. Attachment D contains descriptions of 

some incidents that resulted in internal doses, and Attachment E describes scaling of coworker dose 

intakes consistent with type and duration of exposures. Attachment F discusses a bounding dose 

estimate for potential intakes of radioactive material for onsite responders to the Cerro Grande Fire of 

2000.

 2 The facility was known as the Los Alamos Scientific Laboratory (LASL) until January 1981, when the name was 

changed to Los Alamos National Laboratory. For convenience, this TBD uses LANL for all years of operation.
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5.1.3 Special Exposure Cohort

The Secretary of the U.S. Department of Health and Human Services has designated four classes of 

LANL employees as additions to the SEC based on the findings and recommendations of NIOSH and 

the Advisory Board on Radiation and Worker Health Board:

• Employees of DOE predecessor agencies and their contractors or subcontractors who were 

monitored or should have been monitored for exposure to ionizing radiation during radioactive 

lanthanum (RaLa) operations at Technical Area (TA)-10 (Bayo Canyon Site), TA-35 (Ten Site), 

and Buildings H, Sigma, and U (in TA-1) for a number of workdays aggregating at least 250 

workdays during the period from September 1, 1944, through July 18, 1963, or in combination 

with workdays within the parameters that have been established for one or more other classes 

of employees in the SEC (Leavitt 2006; NIOSH 2006).

Through the course of ongoing dose reconstruction and research, NIOSH determined that, due to 

undocumented worker movements across the site and limited claimant-specific information pertaining 

to work locations, it was unable to eliminate any specific worker from potential exposure scenarios 

based on assigned work location. NIOSH has found that a determination cannot always be made as 

to whether or not an employee worked in Technical Areas with a history of radioactive material use, or 

whether an employee should have been monitored for radiological exposures. Accordingly, NIOSH 

determined that it was necessary to remove the area-specific and monitoring criteria from the class 

description above, and to expand the SEC class definition to include all areas of LANL, and all

employees of the DOE, its predecessor agencies, and their contractors and subcontractors who 

worked at LANL during the specified time period, regardless of monitoring. For this reason, the class 

of employees listed below was added. 

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

and subcontractors who worked at the Los Alamos National Laboratory in Los Alamos, New 

Mexico from March 15, 1943 through December 31, 1975, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in 

combination with work days within the parameters established for one or more other classes of 

employees in the Special Exposure Cohort (Sebelius 2010).

• Employees of DOE, its predecessor agencies, and DOE contractors or subcontractors who 

were monitored or should have been monitored for radiological exposures while working in 

operational TAs with a history of radioactive material use at LANL for an aggregate of at least 

250 workdays during the period from March 15, 1943, through December 31, 1975, or in 

combination with workdays within the parameters that have been established for one or more 

other classes of employees in the SEC (Leavitt 2007; NIOSH 2007).

• Employees of the Department of Energy, its predecessor agencies, and their contractors and 

subcontractors who worked at LANL from January 1, 1976, through December 31, 1995, for a 

number of workdays aggregating at least 250 workdays occurring either solely under this 

employment or in combination with workdays within the parameters established for one or 

more other classes of employees included in the SEC (Sebelius 2012; NIOSH 2012).

NIOSH has determined that it does not have access to sufficient information – including internal 

personnel dosimetry, workplace monitoring data, or sufficient process and radiological source 

information – that would allow it to estimate with sufficient accuracy the potential internal doses 

workers might have received. NIOSH has determined that there is insufficient information either to 

estimate the maximum radiation dose for every type of cancer for which radiation doses are 

reconstructed that could have been incurred under plausible circumstances by any member of the 

classes or to estimate the radiation doses of members of the classes more precisely than a maximum 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 14 of 117

dose estimate. NIOSH has determined that it is possible to reconstruct or bound external doses, 

occupational medical doses, and the occupational internal doses listed in Section 5.1.2.

5.1.4 Background of Internal Exposures

When operations began at LANL in 1943, the only method of monitoring intake was through loose 

contamination swipes. Swipes with a lightly oiled filter paper were taken of surface areas likely to be 

contaminated. Any swiped area with an activity of more than 500 cpm (1,000 dpm) alpha required 

decontamination. (The efficiency of the stationary counters used to count the swipes approached 

50%.) In addition, nasal swipes (also called nose counts) were used to indicate potential intakes. 

Nasal swipes with alpha activity higher than 50 cpm indicated the need for follow-up bioassay. The 

first air samplers became available in the fall of 1944. However, the swipe technique continued as the 

primary method of detection in many areas until the early 1950s. Respiratory protection equipment 

(e.g., assault gas masks or respirators) was used as early as 1944 (Oppenheimer 1944).

In 1943, radiation hazards of the project were limited to external radiation from the cyclotron, the Van 

de Graaff accelerator, radium sources, and a few micrograms of plutonium that arrived in the summer 

of 1943. In the spring of 1944, the first milligram quantities of plutonium arrived at LANL. Until that 

time, the laboratory was in the construction phase. There were also internal and external radiation 

hazards from uranium (ENSR 2002).

In 1944, the radiological hazards of plutonium had been recognized, although it was not yet realized 

that plutonium was more hazardous than radium. Safety regulations based on experience with 

radium dial paint plants were established. Measures to control personnel exposures included multiple 

changes of clothing, showers before leaving the building, use of surgical gloves and respirators, and 

use of closed systems whenever possible. These measures were primitive by current standards. 

Most workers cooperated with safety rules to the best of their ability, but the potential for 

contamination and intakes was present. During the tension and feverish activity of the development 

of the first atomic bomb, it was difficult to avoid some shortcuts in the observation and enforcement of 

safety rules (Hempelmann, Richmond, and Voelz 1973). The Laboratory went from handling a few 

micrograms of plutonium in 1943 to kilogram quantities in 1945. This provided little time for the usual 

development of safe methods of handling and safety equipment design (Schulte and Meyer 1957). As 

research determined that plutonium was more hazardous than radium, tolerance levels and maximum 

permissible body burdens (MPBBs) were reduced significantly. The MPBB for plutonium began at 

5 μg (0.33 μCi) in 1943. This was reduced to 1 μg (0.07 μCi) in October 1944 and further to 0.6 μg 

(0.04 μCi) in 1951 (Langham et al. 1962).

5.1.4.1 Early Research with Humans

The Plutonium Experiment, which involved plutonium tracers, was undertaken between April 1945 

and July 1947. This was a joint project between LANL and the Atomic Energy Project of the 

University of Rochester School of Medicine and Dentistry. If there are references to involvement in 

human experimentation in the dosimetry records or the telephone interview, and they are determined 

to be relevant to the dose reconstruction, then details for this experimentation are available in 

Section A.15.

5.1.4.2 Early Safety Efforts

Early safety efforts were based on working in safe contamination levels (LASL 1944a). In 1952, the 

Health-Safety Rules for Building 52 included restriction of access only through locker rooms and 

release of clothing with levels less than 100,000 cpm (LASL 1952a). Smoking was permitted in 

various contaminated areas at least through 1952 (LASL 1952b). In 1954, contamination 

measurements were based on contact with a shield-open Geiger-Müller (GM) tube for beta/gamma 
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and a Pee Wee probe of a 55-cm2 area for alpha. The efficiency of these portable probes is 

approximately 10% to 15%. Swipes that were counted on a fixed proportional counter had an 

efficiency of approximately 40% to 50%, although the oil on the swipes that increased the collection 

efficiency might have decreased the counting efficiency slightly. A total alpha count rate of 500 cpm 

on a swipe corresponded to 0.007 μg or 0.0004 μCi of the plutonium isotopic mixture of the times 

(Hempelmann, Richmond, and Voelz 1973).

Tolerance limits were established as the level below which the risk of health effects was considered 

acceptable (safe) to continue work or to not take immediate action to correct the condition (LASL 

1945). The tolerance limit for wounds was 10 cpm for alpha and 0.15 mrep/hr for beta/gamma, 

except wounds contaminated with 90Sr, for which the tolerance limit was 0.05 mrep/hr. The skin 

contamination tolerance limit was 1 mrep/hr for beta/gamma, except for 90Sr, which was 0.05 mrep/hr. 

The skin contamination tolerance limit for alpha emitters was 1,000 cpm for polonium, 500 cpm for 

Tuballoy or oralloy, and 250 cpm for plutonium. Decontamination was required when swipe results 

were greater than 500 cpm. Floors in laboratories were mopped once or twice a day to maintain safe 

contamination levels (Hempelmann, Richmond, and Voelz 1973). Corrective actions appropriate to 

the situation (e.g., decontamination of an area, release of personnel with skin contamination) were 

initiated if tolerance limits were exceeded. Workers exceeding bioassay tolerance limits, as discussed 

in later sections, might have been restricted temporarily or permanently from working with radioactive 

materials (Kolodney 1946). Early Health Group reports indicate contamination inside many 

respirators, which indicates improper storage and handling and poor fit during use (LASL 1944b).

Kilogram quantities of plutonium began to arrive at LANL in April 1945. At that time, portable alpha 

counters, continuously operating air samplers, supplied air lines, and specially made positive pressure 

masks were available. Procedures were performed in open hoods and wooden dry boxes, which 

were the precursor to the modern glovebox. Research indicated that there was potential for work and 

casual encounters with plutonium and other radionuclides at various air concentrations or surface 

contamination levels, including levels that exceeded radiation exposure or control limits. Given that 

these operations were, in many cases, the first of their kind and that health physics practices were 

being developed and implemented at the same time plutonium processes were being brought on line, 

some level of chronic or episodic intake during this early period would be a reasonable assumption for 

any worker in the Laboratory (Hempelmann, Richmond, and Voelz 1973). The Health Safety Report, 

Chemical and Metallurgical Division, May, 1944 (LASL 1944c), discussed the contamination surveys 

from D-118 and D-119 as an example. “These data are further proof that both wet and dry material 

are airborne. They emphasize the necessity for assuming that every uncovered item in the laboratory 

is contaminated” (LASL 1944c). Therefore, exposure to some amount of airborne contamination is a 

reasonable assumption for all workers during this early era, regardless of job title, even construction 

trade workers and janitors. Consideration of this fact is a part of the basis for the additions of classes 

of LANL workers to the SEC as discussed above.

From 1943, the Health Group was responsible for the establishment of health standards, specifically 

for safe levels of exposure to radiation and to radioactive and chemical materials. The Health Group’s 

primary concern was to protect the health of Laboratory employees. Formalized Health Rules were 

established for various areas (Burke 1946; Tribby 1946a). Until mid-1951, for want of adequate staff, 

the Group accepted help for monitoring radiation-related activities from staff members in the 

Chemistry and Metallurgy Research (CMR) organization. Over the years, the Health Group evolved 

into the Health Division [and its successors the Environmental, Safety, and Health (ESH) Division and 

the Health, Safety, and Environment Division], with groups in each to address health physics, medical, 

(industrial) safety, biomedical research, industrial hygiene, industrial waste treatment, and 

environmental studies. While several division and group name changes have occurred since 1943, 

the generic Health Physics Group has existed since 1951. Throughout this period, that group has had 

the responsibility for assigning and scheduling bioassay analyses for intakes of all radioactive 

materials. Until the late 1990s, the Industrial Hygiene Group performed all bioassay analyses. Since 
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the late 1990s, bioassays have been performed by one of the chemistry groups. The Medical Group 

has treated individuals accidentally exposed to radiation and radioactive materials, performed physical 

examinations, and treated industrial accidents.

Over the years, many improvements have been made in monitoring, bioassay techniques, safety 

equipment, and safety procedures (Schulte and Meyer 1957). Nevertheless, the potential for 

monitored and unmonitored intakes has existed throughout the history of the site.

Nuclides with the widest historical and current application throughout the LANL facilities are:

• Tritium (3

H),

• Uranium (238U, 234U, 235U),

• Plutonium (238Pu, 239Pu, 240Pu, 241Pu and, to a lesser exposure significance, 242Pu and 244Pu),

• Polonium (210Po) through the late 1950s, and

• Americium (241Am).

These radionuclides of primary internal dosimetric concern are listed in Laboratory reports from 1943 

to the present. (These are referred to as primary radionuclides throughout the remainder of this TBD.) 

Other radionuclides, such as the mixed fission products (MFPs) and activation products (mixed 

activation products, or MAPs) 14C, 11C, 13N, and 75Se, are associated with work areas and years of 

operation in Attachment A, Table A-8 (Inkret et al. 1998a). The fission products 140La, 140Ba, 90Sr, and 

89Sr that are associated with RaLa operations (September 1, 1944, through July 18, 1963) are also of 

significance. Dose reconstruction has been determined to be infeasible for radionuclides other than 

primary radionuclides in the period before January 1, 1976. This includes doses from RaLa 

operations, as the doses from these operations did not result from primary radionuclides. From this 

date through December 31, 1995, NIOSH has determined that it is feasible to reconstruct doses for 

which internal dose monitoring information is included in individual claim files, or in cases in which 

doses from primary radionuclides may be assigned using coworker intakes (NIOSH 2012).

Work areas include plutonium facilities (238Pu or 239Pu), uranium facilities, polonium facilities, tritium 

facilities, laboratory facilities, reactors, accelerators, and others. Case files might not have specific 

information about the assigned work areas of individuals, but detailed work histories, dose monitoring 

information, and incident reports might be found in the documents from DOL. When information about 

the work location is available, Table A-8 in Attachment A can be used to determine the probable 

nuclides. Section 5.2 discusses in vitro methods for specific radionuclides. Excreta bioassay 

methods for determining internal exposures were developed for plutonium in late 1944 (fully 

implemented in April 1945), for polonium in 1944, for uranium in 1949, and for tritium in 1950. Only 

workers with a significant potential for exposure were monitored. A survey in 1986 estimated that 

approximately 350 persons had known burdens of plutonium.

Starting in 1944, blood tests were performed after potential exposures. These were typically 

performed for blood count parameters in relation to external radiation exposure or the probability of 

poisoning rather than the concentration of radioactive material in the blood (Kolodney 1946); the 

results are not generally applicable to internal dose calculations.

Air samples, which might be identifiable in relation to an individual’s record, were collected and 

analyzed beginning in 1944 but are not routinely provided with the LANL records.

As the state of the art of radiation detection progressed, whole-body counting for fission products 

began in 1955, wound counting began before 1967, and chest counting began in 1970. Section 5.3 

discusses historical and current in vivo bioassay methods.
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5.1.4.3 Bioassay Program

Before the 1970s, individuals were assigned to a bioassay program as determined by the area health 

physics monitors. LANL deemed this program sufficient to ensure that all workers who might require 

monitoring were monitored. However, instances might have occurred, especially in the early history, 

in which a person not normally assigned to radiation work was asked to participate as a substitute in a 

task that involved radiation or radioactive materials. These persons were not likely to have regularly, 

or possibly ever, participated in the bioassay program. It is possible that their participation in these 

tasks was never recorded. Indications of this type of exposure might come from claimant interviews 

or work history statements. For this reason, potential unmonitored exposures should be considered in 

many cases for LANL workers. However, NIOSH has determined that it is infeasible to estimate 

internal dose other than those based on case-specific bioassay monitoring information or those that 

can be based on coworker intakes for primary radionuclides for all workers before January 1, 1996.

In the 1970s, LANL initiated the use of an Employee Health Physics Checklist. This checklist allowed 

the evaluation of each individual for potential internal and external exposure. Individuals were placed 

on a monitoring schedule based on this checklist. The checklist is still used and was computerized in 

1998 as the Dosimetry Enrollment System.

5.1.4.3.1 Individual Bioassay Results

The Oak Ridge Associated Universities (ORAU) Team has generated a database from the LANL 

Bioassay Data Repository (LABDR in this document) to consolidate dosimetry records and make 

these data available for use in the dose reconstructions. The repository consists of several 

applications including the Bioassay Enrollment Scheduling and Tracking, In Vivo Measurements 

Laboratory (IVML), and Radiological Incident Reports (RIR) systems, as applicable. Electronic data 

from many other sources were collected and uploaded as part of the project. The new database is 

referred to as the ORAU Team LANL Bioassay Database in this document (OTLBD). The OTLBD 

database consists of the in vitro records for only the five major analytes of dosimetric significance –

uranium, plutonium, americium, polonium, and tritium – and the results of in vivo bioassay. Other 

bioassay might have been performed for a worker but the results have not been captured in the 

database. The results will most likely not be supplied by LANL without special requests. Autopsy 

results are clearly labeled (ORAUT 2009a).

Bioassay results were supplied to the ORAU Team in both Portable Document Format (PDF) and 

Microsoft Excel formats; however, dose reconstructors are to use the PDF files in the case records 

system. In both types of file, the information is divided into five sections: the demographics of the 

worker, Current In Vitro Analysis Data (1990 to the present), Historical In Vitro Analysis Data (pre1990), In Vivo Analysis Data, and Incident Data. Specific details about each section of the report are 

discussed below in the sections on the relevant radionuclides and in Attachment A.

A description of the bioassay data that were reviewed and used to populate the OTLBD, along with 

the detailed validation and verification that were performed as part of that process, is being compiled 

in ORAUT-OTIB-0063, Los Alamos National Laboratory Bioassay Repository Database (ORAUT 

2009a). As a result of this project, LANL management and the NIOSH Project Manager have 

determined that the level of verification and validation is acceptable given the financial constraints of 

the Bioassay Data Repository Project. The validation and verification process is described in ORAUT 

(2009a).

Each PDF file has a page with the description of the database and a summary of the validation 

process of the data. A section on Data Use Instructions is included. The information in these 

sections is not repeated here because it might be modified if the format of the files is revised in the 

future. These sections in the PDF file should be reviewed for each energy employee in the event that 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 18 of 117

certain parameters might have been modified or updated in future files. Units and codes are clearly 

identified. Units might not be consistent within a sample result. For example, the sample result might 

be listed in picocuries per 24 hours, and the minimum detectable activity (MDA) might be listed in 

picocuries per sample or aliquot. Units of activity per day and activity per sample or aliquot are not

necessarily equivalent.

Descriptive codes and format descriptions are listed in each PDF file. Current codes are described in 

Attachment A. Information on documented incidents might be reported in the PDF file. This section 

might have captured recorded incidents that were listed in the original in vitro bioassay plutonium and 

americium database. Other incident information might not have been captured in the database.

Records associated with individual claims may indicate that chelation therapy was administered. 

Approximately 35 workers are known to have received chelation therapy for plutonium and americium 

intakes at LANL, and notes about chelation therapy might be found in the comments section of the 

PDF files. The comment “URINEfix” indicates a bioassay sample that is associated with chelation. 

Chelation therapy was listed as a code in the LABDR, but only the comments (not the codes) were 

transferred to the current OTLBD. It is assumed that every chelation event in the original LABDR

included comments, but this is not ensured. Records of administration of chelating agents, associated 

with individual worker names, exist in the project database.

5.1.4.3.2 UPPU Club

One small but significant group of workers is the UPPU Club (intended to be pronounced as “You Pee 

Pu”). This group consists of individuals who accumulated a significant plutonium body burden and 

who agreed to be monitored periodically and continue to be monitored even past the end of their 

employment at LANL. Only two members were added to the group after the initial 1951 startup. 

Membership in this group can be noted in claimant interviews. Bioassay results can be found in the 

individual’s record many years past the end of employment or past the time of potential exposure. 

This group has typically been monitored at 5-year intervals (Hempelmann, Richmond, and Voelz 

1973).

5.1.5 Unmonitored Workers and Unreported Bioassay Results

The following sections describe several types of employees for which potential unmonitored internal 

doses must be considered when reconstructing internal doses. In each case, a review must be 

performed for potential unmonitored internal doses using case information. Programmatic problems 

identified in the Tiger Team assessment (DOE 1991) indicate that the potential exists for unmonitored 

intakes. Unmonitored intakes may be estimated using ORAUT-OTIB-0062, Internal Dosimetry 

Coworker Data for Los Alamos National Laboratory (ORAUT 2009b). However, doses from potential 

intakes of other radionuclides may only be based on dose monitoring information in claim files. 

Because of the potential for internal dose that is not reflected in internal dose monitoring information, 

classes of LANL employees have been added to the SEC that include all workers in the period before 

January 1, 1996.

5.1.5.1 Workers Before 1949

The potential effects of exposure to plutonium were recognized early by its discoverer, Glenn Seaborg 

(Hempelmann, Richmond, and Voelz 1973). Therefore, when the first few micrograms of plutonium 

arrived at LANL in 1943, there was awareness of some of the radiological and biological effects and 

hazards. Stringent safety measures were put in place immediately, including the use of homemade 

dry boxes when practical. “However, because of the urgency of the times, work with plutonium had to 

proceed, and improvised methods of monitoring and decontamination were unbelievably primitive by 

today’s standards” (Hempelmann, Richmond, and Voelz 1973). In 1944, bioassay techniques 
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consisted of swipes of both nostrils at the end of the working day. Monitoring results were described 

as whether maximum permissible concentrations (MPCs) were exceeded, not whether there was an 

intake. When kilogram quantities of plutonium arrived at LANL in 1945, the surface swipes and nasal 

swipe counting (as primary monitoring) had been replaced in certain areas by continuously operating 

air samplers and portable alpha counters. Then, by March 1945, a urine assay method for plutonium 

had been developed. Body burden estimates were made from urine results as early as 1953.

Working conditions were described as “deplorable by present-day standards” (Hempelmann, 

Richmond, and Voelz 1973) until September 1946, when the new facility at DP Site (TA-21) was 

constructed. Before the move to TA-21, suspension of work and decontamination of the area were 

required when alpha contamination exceeded 10,000 cpm (assume about 50% counting efficiency). 

Decontamination was recommended between 2,000 and 10,000 cpm but was not required. Even 

hallways and other unrestricted areas had contamination. In 1944, shoe covers that were worn by 

secretaries and others working throughout the buildings had count rates of 2,500 to 7,500 cpm (LASL 

1944b). Contamination above 500 cpm (0.007 μg or 0.0004 μCi of the plutonium isotopic mixture) 

was the reporting level for plutonium areas. In June and July 1945, over 50% of the laboratories had 

areas that routinely exceeded the maximum removable contamination level. The potential for 

unmonitored intakes was significant in the early years (1944 to 1946) for any site worker, but bioassay 

was provided only for the most exposed workers. Because the excretion of plutonium is continuous 

after an intake, significant intakes might have been identified on later routine bioassay samples with 

improved sensitivity for workers who remained at LANL.

As bioassay sensitivities and respiratory protection equipment improved, the potential for intakes 

decreased. Due to rigorous workplace monitoring, the probability that a worker could have received a 

large intake of radioactive material that was unmonitored and unnoticed was less after 1946, although 

the probability of unmonitored small intakes was larger. Periodic reports from H Division of air 

samples, contamination incidents, and hot spots continued to identify a significant number of overtolerance occurrences throughout the 1960s. Respiratory protection was available but, except in a 

few locations, was only donned when a continuous air monitor (CAM) alarmed or when airborne levels 

approached tolerance or action levels (ALs). Therefore, a potential existed for an intake before the 

alarm. Review of Summary of Radiological Incident Reports (RIR), January 1993 to June 1998

indicates that, while the number of RIRs has decreased significantly, the potential for unmonitored 

small intakes continues to exist (Bates 1998).

In addition, instances might have occurred, especially in the early history, in which a person who was 

not normally assigned to radiation work was asked to participate as a substitute in a task that involved 

radiation or radioactive materials. These workers were not likely to have regularly, or perhaps ever, 

participated in the bioassay program or potentially had their participation in the task recorded. 

Indications of this type of exposure might come from claimant interviews.

5.1.5.2 Short-Term Workers

Short-term workers, such as summer students, persons engaged in postdoctoral work, contractors, 

and teachers, might not have fully participated in routine bioassay programs. These workers were 

monitored for internal exposure only in unusual circumstances. Near their termination dates, workers 

might have received in vivo counts. Workers might have been required to submit an initial urine 

sample or have an initial in vivo count.

5.1.5.3 Zia Company and Other Contract Service Workers

The Zia Company was the service worker contractor. Zia employees participated in a separate 

monitoring program from that for Laboratory employees. As early as 1946, directives restricted the 

assignment of maintenance workers to either DP East or West (not both) because of the difficulty of 
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separating the radionuclides in urine bioassay samples and monitoring DP East workers for plutonium 

when they might have been exposed in both areas (Burke 1946; Tribby 1946a). In 1954, that 

restriction was removed because of the improvement in the bioassay techniques that would permit the 

specific radionuclides to be identified. However, although listed as a monthly requirement for 

hazardous areas as early as 1946, submission of urine samples was dependent on determination of a 

“high nose count” or an “extremely hazardous job” or at the request of the DP East Section Leader 

(Meyer 1954).

In 1975, it was determined that the H-1 and H-5 efforts to schedule plutonium urine samples for Zia 

employees were inadequate. Only Zia employees who were permanently assigned to DP West and 

the CMR Building were being sampled; other Zia employees in plutonium areas were not. To ensure 

adequate coverage of Zia employees, the scheduling of plutonium samples for Zia was delegated to 

the Zia Safety Office. The requirement was annual samples for those employees “required to work in 

Pu areas” (Lawrence 1976). A quarterly report of the H-1 Division stated that Zia management had

interpreted this to mean employees “required to enter Pu areas for any reason,” (University of 

California 1978a p. 208), whereas a number of individuals in these areas would not have had a 

potential for intake of plutonium. “By July 1976 it became apparent that the number of Zia employees 

being sampled for Pu was growing without bounds”. A program review was conducted in 1978, in 

which it was shown that the number of bioassay samples had “exceeded the level of effort agreed 

upon by Zia management,” (University of California 1978a). In 1978, in another attempt to reduce the 

number of sampled Zia employees to 500 per year; supervisors who only performed inspections were 

eliminated from the schedule. In later years, some jobs were exempt from monitoring. Attachment A, 

Table A-9, contains a list of criteria and exempt job categories (University of California 1978a). In 

more recent years, other service contractors have participated in site activities.

5.1.5.4 RaLa Project Operations, 1944 to 1963

Workers on the RaLa Project from September 1, 1944, through July 18, 1963, might or might not have 

been monitored for potential intakes of 90Sr, 89Sr, 140La, and 140Ba; although it appears that RaLa 

workers at TA-10 (Bayo Canyon Site), TA-35 (Ten Site), and Buildings H, Sigma, and U (in TA-1) 

were monitored for external radiation. A total of about 250,000 Ci of RaLa was released to the 

atmosphere from these experiments. The highest exposures at LANL from RaLa operations were to 

the chemists who prepared the sources.

In vitro bioassay results related to RaLa operations have not been found in claim files to date. The 

whole-body counting program was not well established until 1970, although there are reports as early 

as 1957 that identified specific nuclides for workers at the cyclotron (Buckland 1959). Workplace air 

monitoring data for the RaLa areas also have not been found. Therefore, estimates of intakes and 

assessment of internal dose are not possible for RaLa operations workers (NIOSH 2006).

5.1.5.5 Responders to the Cerro Grande Wildlands Fire of 2000

In May of 2000, areas of Los Alamos were overrun by the 48,000-acre wildlands fire that came to be 

known as the Cerro Grande fire. Some LANL structures were destroyed and damaged, but there was 

no damage or destruction of stored special nuclear material. According to meetings conducted with 

personnel employed by LANL, responders to that incident included LANL staff and members of 

outside emergency response organizations who were not monitored for potential intakes of 

radioactive material.

During the fire, sampling of airborne particulate was conducted using LANL’s AIRNET system. The 

most abundant detected radioactive materials resulted from resuspension of radon progeny that had 

accumulated on vegetation and on the forest floor, but concentrations of plutonium, americium, and 

uranium were also measured. These results were, in general, consistent with measurements that 
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were made outside the fire period (Eberhart 2010). For estimation of doses from potential intakes of 

anthropogenic radionuclides, the results of the monitoring were evaluated as described in 

Attachment F. Doses to a responder in the area of the highest measured concentration, assuming a 

60-hour exposure, are listed in Tables F-2 and F-3, in Attachment F.

Because no organ dose was greater than 0.001 rem, no additional unmonitored dose, beyond doses 

from environmental concentrations of radioactive material, are necessary for potential exposure 

during the Cerro Grande fire.

5.1.6 Overview of Internal Dose Reconstruction Process for Los Alamos National 

Laboratory

As described in the preceding portions of this introduction, doses inferred from internal dose 

monitoring records might not completely represent the total dose that was received by a LANL 

employee. NIOSH has determined that it is not feasible to reconstruct internal doses when bioassay 

information is not present in claim files. Two types of internal exposures must be addressed in dose 

reconstruction for LANL workers: doses based on available internal dose monitoring records 

(measured and missed doses) and unmonitored intakes, which may be estimated using coworker 

intakes based on bioassay data for primary radionuclides for monitored workers.

The dose reconstructor must review the records in the case with special care to identify all potential 

internal doses, including potential unmonitored intakes. This evaluation will result in two possible 

methods for estimating internal dose: (1) calculating the dose from bioassay results (missed and 

fitted internal doses); and (2) assigning the dose based on coworker dose intakes in ORAUT-OTIB0062, Internal Dosimetry Coworker Data for Los Alamos National Laboratory (ORAUT 2009b).

Calculation of internal doses is discussed in Sections 5.2 to 5.4, which address in vitro bioassay, in 

vivo bioassay, and interferences and uncertainties, respectively. In addition, a methodology is given 

from ORAUT-OTIB-0062 (ORAUT 2009b) to allow the assignment of a dose using coworker dose 

intakes.

5.2 IN VITRO BIOASSAY MINIMUM DETECTABLE AMOUNTS, COUNTING METHODS, 

AND REPORTING PROTOCOLS FOR PRIMARY RADIONUCLIDES

Historically, the in vitro bioassay program included the nuclides and techniques listed in Tables 5-1

and 5-2. This section contains a detailed discussion of the analysis techniques that were used for 

selected radionuclides. Table A-24, Attachment A, summarizes in vitro bioassay sensitivities by 

technique and year. However, without a special request, only the results for the primary radionuclides 

(tritium, uranium, plutonium, 210Po, and 241Am as listed in Table 5-1) will be supplied for the worker. If 

an exposure to primary radionuclides that is not monitored by in vitro bioassay is suspected, dose is 

assigned based on the coworker intakes in ORAUT (2009b). In some cases, special inquiries can be 

made to determine if additional bioassay results exist for the worker (e.g., MFPs or 90Sr).

Table 5-1. Historical in vitro bioassay, primary 

radionuclides.

Material Start year Comments

Tritium 1950

Uranium 1949 Significant 

numbers of results 

begin in 1950.

Plutonium 1944

Polonium-210 1944

Americium-241 1954
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Table 5-2. Historical in vitro bioassay not available in claim 

records.

Material Start year Comments

Gross beta 1952 (maybe 1947) Not done regularly

Protactinium-231 1958 Not done regularly

Radium-226 1958 Not done regularly

Thorium-230 1958 Not done regularly

Actinium-227 1954 Not done regularly

Strontium-90 Unknown Not done regularly

5.2.1 Plutonium

The most serious intakes at LANL have involved isotopes of plutonium and 241Am. Operations with 

small quantities of plutonium and curie quantities of 210Po were conducted primarily in D Building in 

1944. The first urinalysis for evidence of plutonium uptake was performed on July 18, 1944, using a 

chemical procedure that was developed at the University of Chicago Metallurgical Laboratory in June 

1944. However, Health Group personnel soon discovered that the procedure, which used a 50-cm3

aliquot of a 24-hour urine sample, did not have the sensitivity to meet the tolerance level of a 5-μg 

(0.33-μCi) burden of plutonium. One of the first recorded accidents in which a human was subjected 

to a possible intake of plutonium occurred in May 1944 (Hempelmann, Richmond, and Voelz 1973). A 

minor chemical exposure on August 19, 1944, was the first recorded accident in which a human was 

subject to a possible intake of more than the 5-μg plutonium tolerance limit with the release of 10 g of 

plutonium into the face of a chemist. This accident resulted in LANL being authorized to proceed to 

develop a more sensitive procedure. The resultant cupferron extraction procedure and subsequent 

procedures are discussed later in this section (Moss 1990).

References to products called X10 and 49 indicate plutonium or isotopic mixtures of plutonium that 

are predominantly 239Pu (Lawrence 2004). (The “49” is a shorthand reference that is a combination of 

the last digit of the atomic mass number, 94 for plutonium, and the last digit of the atomic weight, 239 

for 239Pu.)

The chemical forms of plutonium that have been currently and historically encountered at LANL 

include oxide, nitrate, fluoride, and metal. Plutonium oxide is the most common form. However, in 

lieu of detailed information specific to the case, dose reconstructors typically assign the absorption 

type resulting in the highest dose. When such information is not present, types M and S are 

evaluated, and, in accordance with ORAUT-OTIB-0049, Estimating Doses for Plutonium Strongly 

Retained in the Lung (ORAUT 2010), correction factors for certain organs could result in additional 

dose.

5.2.1.1 Special Considerations for 238Pu Dosimetry Related to a Wing 9 Incident

A special case of excretion kinetics has been observed at LANL. The 1971 Wing 9 incident involved 

intakes of high-fired 238Pu oxide that exhibited a very low excretion rate for the first 100 days after 

intake. A gradual rise in excretion rate was observed after that period, with no additional suspected 

intake (Miller et al. 1999). This has been characterized as a “nonmonotonic” excretion rate, and this 

phenomenon observed at Los Alamos, along with a similar event observed at the Mound site, are 

described by ORAUT (2013). 

In 1971, an accidental release of 238Pu occurred in Wing 9 during examination of a radioisotope 

thermoelectric generator (RTG) powered by a ceramic 238PuO2. Intakes from this accident were 

characterized by nonmonotonic excretion rates by the exposed individuals (Guilmette, Griffith, and 

Hickman 1994). Parameters for the Integrated Modules for Bioassay Analysis (IMBA) program were 

developed that mimic these excretion rates, as documented in ORAUT (2013). The parameters are 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 23 of 117

reproduced here for applications to dose estimates that could be associated with this accident. These 

parameters, listed as “type J,” are compared with types M and S in Table 5-3.

Table 5-3. IMBA input parameters for types M, S, and 

nonmonotonic type J.

Rate constant Type M Type S Type J

kp,t 90 100 0.00189

kt,blood 0.005 0.0001 0.000257

kp,blood 10 0.1 1.00 E-06

kr 100 100.1 0.001891

ks 0.005 0.0001 0.000257

fr 0.09995 0.001 -0.156671

Dose reconstructors are cautioned that 238PuO2 typically exhibits excretion rates somewhere between 

types M and S (Cheng et al. 2004), and either M or S should normally be assumed. However, the 

dose from type J should be calculated for individuals who were involved in the 1971 Wing 9 incident.

It is not intended that Type J should be applied as an overestimating assumption; only that it is an 

available assumption for fitting bioassay data. 

5.2.1.2 Sample Collection Procedures

At the onset of the LANL bioassay program, samples were collected in a clean area after a 

decontamination shower. However, these samples had a potential for contamination. Occasional 

high values were assumed to be an artifact of the sample and not evidence of internal exposure 

(Hempelmann, Richmond, and Voelz 1973). By the spring of 1945, a Health Pass Ward was 

established at the hospital to ensure collection of contamination-free samples (Nickson 1945). This 

procedure allowed for 2 days off the job, 1 day to clean up, and one 24-hour period spent at Los 

Alamos Hospital where the urine sample was collected. The procedure was modified in 1948 to allow 

only 1 day off the job during the sample collection period. While a contamination-free sample was 

ensured, this procedure was later deemed extremely expensive and was eliminated in 1952 (Clark 

2005). A procedure that allowed collection of an equivalent 24-hour sample while the employee was 

off the site was initiated. Between 1952 and 1958, the samples were collected in three disposable 

bottles that were carried in a kit. The employee was to collect four voidings in the three bottles, which 

some found objectionable and led them to collect only three voidings. The kits were rarely washed 

and therefore were considered a source of potential contamination. In January 1958 the kit provided 

four disposable bottles (Clark 2005). A modified procedure to collect a 24-hour equivalent sample 

(two morning and two evening voidings) with a four-bottle disposable kit continues to be used.

Studies of special, timed spot samples have been performed to provide data that are used to apply 

correction by volume and specific gravity to provide realistic results in disintegrations per minute per 

day. All sample results, except thermal ionization mass spectroscopy (TIMS) results, are expected to 

be reported in units (typically picocuries) per 24-hour day in the database. TIMS results are reported 

in both activity per 24 hours and activity per sample. No sample volume is usually provided. The 

current database provides a clear description of units for each result. If MDAs are listed for an 

analysis, these values might be listed in units of activity per sample with a sample volume. Note that 

activity per sample and activity per 24 hours are not equivalent. It appears that the 24-hour sample 

concentrations have been normalized to a 1,400-mL volume.

LANL instituted a sample validation procedure (Lawrence 1978). The Plutonium Body Burden (Q) 

from Urine Analysis (PUQFUA) programs were revised through PUQFUA4. Therefore, other 

validation protocols might have been in effect after 1978. The purpose of these programs was to 

perform a statistical test of high values to determine if these results should be invalidated.
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 24 of 117

Attachment A, Section A.4, lists the validation procedure for plutonium urine bioassay samples. This 

procedure is more rigorous than can be implemented without computer software. However, certain 

samples in the original database were marked invalid based on these criteria. Because the codes 

were not carried over to the current database, it is expected that these designations have been 

removed. However, this practice is mentioned in the event that comparison of the results in the 

current database is made to the original database or logbooks. The dose reconstructor should not 

attempt to implement this validation procedure but should use best judgment in the evaluation of 

samples for inclusion in the intake calculations. At present, Bayesian statistics are used by LANL to 

determine the statistical validity of results. Notes on these determinations are included in the 

comments in the database. Indication that LANL considered the sample as invalid and the reason for 

the status might be included in the notes that follow the results in the PDF file. These comments 

might or might not be relevant to the dose reconstructors determination of the validity of the result.

5.2.1.3 Missed Intakes

Intakes of plutonium can occur from both acute and chronic exposures. Chronic exposures might not 

be identified as incidents but can still result in a measurable burden of plutonium. A body burden can 

result from chronic inhalation exposure to a low-level plutonium-contaminated atmosphere. A study of 

autopsy tissues from LANL workers with high and low potential for exposure to plutonium has shown 

that measurable amounts of plutonium, above that expected from global fallout, could occur in 

individuals with low exposure potential (Foreman, Moss, and Langham 1960). As noted above, 

smoking was permitted in various contaminated areas at least through 1952 (LASL 1952b).

The long-term excretion pattern of plutonium isotopes permits plutonium intakes that produced 

bioassay results below the detection threshold in the early years to become detectable as the 

sensitivity of the analysis technique improved. The date of the intake might not relate directly to the 

last bioassay result below the detection level. The intake might have occurred many years earlier 

(Hempelmann, Richmond, and Voelz 1973).

The first urine tests for plutonium were developed in 1944 and 1945. Urinalysis was difficult and time 

consuming. Therefore, personnel with the most positive nasal swipes had the most urine bioassays 

(Hempelmann 1946; ENSR 2002).

5.2.1.4 Routine Sample Frequencies

Table A-15 in Attachment A lists the routine sampling frequencies by period. In 1978, sampling 

frequencies were reevaluated for LANL and Zia employees. To attempt to reduce the number of 

sampled Zia employees to 500 per year, supervisors who only performed inspections were eliminated 

from the schedule (University of California 1978a). Special and emergency sample frequencies are 

listed in Tables A-19 to A-21 of Attachment A.

5.2.1.5 Sample Analysis Procedures

At first, the urine bioassay analysis procedure could not distinguish adequately between plutonium 

and polonium. Plutonium samples in 1944 were occasionally contaminated with polonium (Clark 

2005). During this period, the total alpha results were assigned to either plutonium or polonium based 

on the individual’s work history. The procedure was modified in the fall of 1944 to extract the 

plutonium. The count time was 30 to 60 minutes (Moss 1990).

The cupferron procedure was in use through late 1949. The bismuth phosphate-lanthanum fluoride 

serial coprecipitation procedure was in use between October 1949 and January 1957. No corrections 

were made for chemical blanks or counting geometry; average chemical recovery factors (82.3% 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 25 of 117

±19.4%) might have been used to interpret the data until 1957 (Clark 2005). The alpha proportional 

counters were in use until 1957.

Improvements were made in the coprecipitation procedure and counting techniques until the 

procedure was replaced by an aluminum nitrate extraction with neutron track analysis (NTA) (Clark 

2005) counting in 1957. Urine samples were radiochemically processed and electroplated, and 

activities were determined by exposure to NTA emulsions. The exposure time for this method was 

10,000 minutes with a background of 0.005 dpm. An ion-exchange technique replaced the aluminum 

nitrate in 1963 (Clark 2005). Use of a ZnS counter began in 1966. All of these methods measured 

total alpha activity from plutonium isotopes.

Radiometric alpha spectroscopy (RAS) was available for use beginning in 1967 (Clark 2005). The 

use of RAS in the analysis is indicated when both 238Pu and 239Pu are reported for the same sample. 

Plutonium-241 and 241Am were not measured by total plutonium alpha procedures. The dose 

contribution from these nuclides should be accounted for using the mixtures listed in Tables 5-4, 5-5, 

and 5-6, unless the plutonium in the intake was known to be a “pure” isotope.

From 1967 to present-day operations, counting has been performed by RAS for specific plutonium 

isotopes. Alpha spectroscopy cannot distinguish between 239Pu and 240Pu because of the similarity of 

the alpha energies. Therefore, the results for RAS that are labeled as 239Pu are actually 239+240Pu.

In 1997, TIMS and application of the class 100 clean room were added to the analytical technique. 

Samples are now routinely analyzed for 239Pu with TIMS and for 238Pu with RAS. The sensitivity of the 

analysis for 239Pu has been reduced 40-fold by the addition of TIMS (Inkret et al. 1999). TIMS does 

not measure 240Pu unless the activity of 239Pu is above an unspecified threshold, in which case the 

results for 240Pu would be reported separately (Lewis 2006a). Therefore, the amount of 240Pu in the 

mixture should be accounted for with the use of Table 5-4, 5-5, or 5-6 when 239Pu results are listed for 

TIMS.

Dose calculations performed in dose reconstructions, including those with IMBA, typically assume a 

combination of 239+240Pu that is treated as 239Pu. Therefore, if the results of TIMS analysis are used for 

calculation of intake, the contribution of 240Pu must be accounted for in the calculation.

In the current LABDR, if TIMS results are listed for a sample the 239Pu RAS analysis for that sample 

could be footnoted with an asterisk (*). The footnote states that 240Pu activity is not measured in the 

result. The comment is an artifact of the logic in the database and could be misleading to the dose 

reconstructor. The flag is applied to all results listed for the sample when TIMS analysis is included. 

RAS cannot distinguish between the energies of 239Pu and 240Pu; therefore the activity of 240Pu, if any, 

is included in the RAS activity listed as 239Pu (Lewis 2006a).

MDAs from procedures and reports for the various radiochemical and counting methods are listed in 

Table 5-7. The MDAs that are reported with the results in the LABDR are frequently in units of 

picocuries per sample. The results of the analysis are listed as actual values (positive, negative, or 

zero) but in units of picocuries per 24-hour sample. These are not equivalent units. The sample size 

is usually listed. The dose reconstructor should normalize the units before determining if the reported 

results are above or below the MDA. TIMS results are listed in picocuries per 24-hour sample and 

picocuries per sample, but no sample volume is typically listed and the listed results are not identical. 

Therefore, the results in picocuries per 24 hours should be used for the dose reconstruction. No MDA 

value is typically listed with the TIMS results, so the MDA in Table 5-7 should be used to determine if

results are above the MDA.
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 26 of 117

Table 5-4. Activity composition of nominal 3% plutonium mixture (Kinderman et al. 1953).

Mixture designation: Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aging:a 0 5 10 15 20 25 30

Component Specific activity in mixture (Ci/g)

Pu-238 9.85E-04 8.72E-04 8.39E-04 8.06E-04 7.75E-04 7.45E-04 7.16E-04

Pu-239b 6.01E-02 6.01E-02 6.01E-02 6.01E-02 6.00E-02 6.00E-02 6.00E-02

Pu-240 6.74E-03 6.73E-03 6.73E-03 6.73E-03 6.72E-03 6.72E-03 6.72E-03

Pu-241 1.44E-01 1.13E-01 8.92E-02 7.01E-02 5.51E-02 4.33E-02 3.40E-02

Pu-242 <1E-06 <1E-06 <1E-06 <1E-06 <1E-06 <1E-06 <1E-06

Am-241 0 1.02E-03 1.82E-03 2.44E-03 2.92E-03 3.28E-03 3.56E-03

Pu-239+240 6.7E-02 6.7E-02 6.7E-02 6.7E-02 6.7E-02 6.7E-02 6.7E-02

Pu-alpha 6.8E-02 6.8E-02 6.8E-02 6.8E-02 6.8E-02 6.7E-02 6.7E-02

Total alpha 6.8E-02 6.9E-02 6.9E-02 7.0E-02 7.0E-02 7.1E-02 7.1E-02

Component Activity ratios

Pu-239+240:Am-241 N/Ac 66 37 27 23 20 19

Pu-239+240:Pu-238 68 77 80 83 86 90 93

Pu-239b:Pu-240 8.92 8.92 8.92 8.92 8.92 8.92 8.92

Pu-241:Pu-239+240 2.2 1.7 1.3 1.0 0.83 0.65 0.51

Pu alpha:Pu-238 69 78 81 84 87 91 94

Pu alpha:Am-241 N/Ac 66 37 28 23 21 19

Pu-alpha:Pu-239+240 1.01 1.01 1.01 1.01 1.01 1.00 1.00

Pu-241:Pu alpha 2.1 1.7 1.3 1.0 0.82 0.64 0.50

a. Time since separation of Am-241 from the plutonium mix.

b. Use Pu-239 only for TIMS results; otherwise use Pu-239+240 for RAS or total plutonium alpha.

c. N/A = not applicable.

Table 5-5. Activity composition of reference weapons-grade 6% plutonium mixture (Battelle 2003).

Mixture designation: Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aging:a 0 5 10 15 20 25 30

Component Specific activity in mixture (Ci/g)

Pu-238 8.56E-03 8.23E-03 7.91E-03 7.60E-03 7.31E-03 7.03E-03 6.75E-03

Pu-239 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02

Pu-240 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02

Pu-241 8.24E-01 6.48E-01 5.09E-01 4.00E-01 3.15E-01 2.48E-01 1.95E-01

Pu-242 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06

Am-241 0 5.83E-03 1.04E-02 1.39E-02 1.66E-02 1.87E-02 2.03E-02

Pu-239+240 7.13E-02 7.13E-02 7.13E-02 7.13E-02 7.12E-02 7.12E-02 7.12E-02

Pu-alpha 7.99E-02 7.95E-02 7.92E-02 7.89E-02 7.85E-02 7.83E-02 7.80E-02

Total alpha 7.99E-02 8.53E-02 8.96E-02 9.28E-02 9.52E-02 9.70E-02 9.83E-02

Component Activity ratios

Pu-239+240:Am-241 N/Ab 12.2 6.87 5.13 4.28 3.80 3.50

Pu-239+240:Pu-238 8.33 8.67 9.01 9.38 9.74 10.1 10.5

Pu-239:Pu-240 4.24 4.24 4.24 4.24 4.24 4.24 4.24

Pu-241:Pu-239+240 11.6 9.09 7.15 5.62 4.42 3.48 2.73

Pu alpha:Pu-239+240 1.12 1.12 1.11 1.11 1.10 1.10 1.10

Pu alpha:Pu-238 9.33 9.66 10.0 10.4 10.7 11.1 11.6

Pu alpha:Am-241 N/Ab 13.6 7.62 5.68 4.73 4.19 3.84

Pu-241:Pu alpha 10.3 8.15 6.43 5.07 4.01 3.17 2.50

a. Time since separation of Am-241 from the plutonium mix.

b. N/A = not applicable.

Urine, fecal, and tissue samples have been analyzed for plutonium. Most in vitro bioassay samples 

have been urine. Fecal and tissue samples have been performed only if requested in special 

circumstances.
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 27 of 117

Table 5-6. Activity composition of reference fuel-grade (12%) plutonium mixture (Battelle 2003).

Mixture designation: Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aging:a 0 5 10 15 20 25 30

Component Specific activity in mixture (Ci/g)

Pu-238 1.71E-02 1.64E-02 1.58E-02 1.52E-02 1.46E-02 1.40E-02 1.35E-02

Pu-239 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.25E-02

Pu-240 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.71E-02 2.71E-02

Pu-241 3.09E+00 2.43E+00 1.91E+00 1.50E+00 1.18E+00 9.29E-01 7.30E-01

Pu-242 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06

Am-241 0 2.19E-02 3.89E-02 5.22E-02 6.24E-02 7.03E-02 7.63E-02

Pu-239+240 7.98E-02 7.98E-02 7.98E-02 7.97E-02 7.97E-02 7.97E-02 7.97E-02

Pu-alpha 9.69E-02 9.62E-02 9.56E-02 9.49E-02 9.43E-02 9.37E-02 9.32E-02

Total alpha 9.69E-02 1.18E-01 1.35E-01 1.47E-01 1.57E-01 1.64E-01 1.69E-01

Component Activity ratios

Pu-239+240:Am-241 N/Ab 3.64 2.05 1.53 1.28 1.13 1.04

Pu-239+240:Pu-238 4.67 4.86 5.05 5.24 5.46 5.69 5.90

Pu-239:Pu-240 1.93 1.93 1.93 1.93 1.93 1.93 1.93

Pu-241:Pu-239+240 3.87E+1 3.05E+1 2.40E+1 1.88E+1 1.48E+1 1.17E+1 9.16

Pu alpha:Pu-239+240 1.21 1.21 1.20 1.19 1.18 1.18 1.17

Pu alpha:Pu-238 5.67 5.87 6.05 6.24 6.46 6.69 6.90

Pu alpha:Am-241 N/Ab 4.39 2.46 1.82 1.51 1.33 1.22

Pu-241:Pu alpha 31.9 25.3 20.0 15.8 12.5 9.91 7.83

a. Time since separation of the Am-241 from the plutonium mix.

b. N/A = not applicable.

5.2.1.6 Plutonium Isotopic Ratios (Mixtures)

Dose assessments at LANL were historically based solely on the isotopes identified in the analysis. 

At present LANL dose assessments are based on a 6% plutonium mixture. The amount of 241Am 

present (i.e., the aging) in a sample with positive 239Pu or 238Pu is determined by analysis for 241Am or 

by process knowledge. If the determination of the mixture is based on process knowledge, the 

amount of 241Am could be noted in the comments. The dose reconstructor should evaluate 241Am

results for the intake. There could be a short interval between the plutonium and the americium 

samples (Lewis 2006b).

Most plutonium mixtures at LANL before 1957 should be assumed to be nominally 3% weaponsgrade plutonium (WGPu) and after 1956 to be nominally 6% WGPu (Lewis 2006b). Tables 5-4 and 

5-5 list activity and weight ratios as referenced for DOE sites for 3% and 6% WGPu, respectively.

In the early years, when total alpha measurements were made for plutonium, 239Pu was the isotope 

chosen based on process knowledge. Results of these total plutonium alpha measurements before 

1967 are found in the database and listed as 239Pu. Since 1967, plutonium analysis has been 

performed with RAS. RAS cannot distinguish between 239Pu and 240Pu because the alpha energies 

are similar and overlapping. However, 239+240Pu and 238Pu can be resolved using RAS. Therefore, 

when quantities of pure 238Pu arrived at LANL in 1968, the analysis was capable of distinguishing 

between the isotopes of plutonium. If the results of both isotopes are not reported as positive, then 

exposure to a mixture of predominantly pure 238Pu is usually indicated when the 238Pu results are 

positive and the 239Pu results are below the MDA.

Based on the isotopic composition of the material that was released to the atmosphere in 1970 during 

an accident in TA-21 (DP West) involving “pure” 238Pu (by weight, with a specific activity of 

14.1 μCi/μg), a typical mixture consists of 80% 238Pu, 16.9% 239Pu, 2.8% 240Pu, 0.2% 241Pu, and 0.1% 

242Pu (Meyer 1970). “Pure” 238Pu was typically present in areas involving heat source technology. A 

typical pure 238Pu mixture is described in Table 5-8. It is assumed that pure 238Pu would be “fresh” 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 

04/18/2013 Page 28 of 117

Table 5-7. Plutonium bioassay sensitivity as listed in procedures and reports.

Nuclide Period

Sample 

typea Technique (era)

MDA level

Unit/24-hr sampleb Reporting limitUnit/24-hr sampleb Tolerance limitcUnit/24-hr sampleb

Pu-total 

alpha

1944 U Cupferrond 0.7 pCi (Clark 2005)e >0.8e pCi 7 cpm/24 hr or 6.3 pCi

1945–1949 U Cupferrond 0.16 to 0.05 pCif

(1 dpm [0.45 pCi] DL [Moss 1990])

>0.8e pCi 7 cpm/24 hr or 6.3 pCi

1949–01/1957 U Biphosphate/alpha counting 0.20 to 0.07 pCif 2 dpm or 0.9 pCi

0.4 pCig

7 dpm/24 hr or 3 pCi

01/1957–1965 U Aluminum nitrate/NTA 0.03 pCih

(0.05 dpm at 99% confidence)i 0.2 dpm or 0.09 pCi 7 dpm/24 hr or 3 pCi

1966 U ZnS 0.03 pCij(0.07 dpm)

Pu-239 1967–1997 U RAS (PHA) only starting in 1971 0.03 pCij(1 mBq)

1977–1981 F RAS (PHA) 1 nCi/sample (less if Am-241 ratio known)

1981–1983 F Phoswich detector, 4-cm sample 

thickness

0.4 nCi/sample or 400 pCi/sample

(17-keV X-rays)

1982–1986 U Alkaline earthi

Oxalatei

Rapidi

Alkaline earth - 0.015 pCi/800 cm3

Oxalate – 0.015 pCi/700 cm3

Rapid – 1.6 pCi/L

1997–2002 U TIMS/RAS (alpha speck) 1.03E-4 pCi/ 24-hr sensitivity

7.6 μBq/24-hrl

1997–present U RAS (alpha spec) 8.1E-3 pCi 

2003–present U TIMSk,m 3E-4 pCi 

Pu-238 1967–1971 U RAS (alpha PHA) 0.03 pCim 0.2 dpm/24 hr investigate

1971–1976 U RAS (alpha PHA) 0.03 pCim

1977–1997 U RAS (PHA) 0.03 pCim

1977–1981 F RAS (PHA) 0.4 nCi/sample

1981–1983 F Phoswich detector, 4-cm sample 

thickness

0.2 nCi/sample (17-keV X-rays)

1997–present U RAS (alpha spec) 8.1E-3 pCi or 300 μBq

a. U = urine; F = fecal.

b. Unless otherwise noted.

c. The tolerance limit is considered the level under which it was safe.

d. “A successful method of analyzing urine was developed in Jan. 1945 but could not be used as a routine test until a contamination free laboratory (ML Building) was ready for use in Feb. 

1945” (Hempelmann 1946).

e. Not adjusted for potential chemical recovery (Clark 2005). Assumes “0.05 dpm/sample” refers to the entire 24-hour sample.

f. Background count rate 1 cpm (changed to 0.1 cpm at some unknown time before 1957), 1,000-minute count time, 50% efficiency, average recovery 82.3% ±19.4% (1945 to 1949) and 

67% ±21% (November 1949 to January 1957).

g. Results above these values were considered high (i.e., positive) and subject to statistical investigation. Source: Lawrence (1978).

h. Campbell et al. (1972); McInroy et al. (1991).

i. Procedures on record, not considered the default analysis method. Verify method used from dosimetry monitoring record.

j. Moss et al. (1969).

k. TIMS results are for Pu-239 only; the results do not include Pu-240, which is indistinguishable from Pu-239 by other types of analysis.

l. 2x sensitivity. University of California (1978a).

m. Inkret et al. (1999). TIMS with ultra-trace chemistry and class-100 clean room and alpha spectroscopy methods. Use of alpha spectroscopy allows direct measure of chemical efficiency 

and detection of Pu-238. TIMS quantifies Pu-239 only, RAS quantifies Pu-239+240.
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 29 of 117

during processing in the heat source technologies. However, intakes of aged pure 238Pu could have 

been possible during decontamination or decommissioning activities in areas where pure 238Pu was 

processed. Table 5-8 includes the activity ratios necessary to calculate the appropriate mixture for 

aged “pure” 238Pu. Potential ingrowth of 241Am can be determined from the ratios in Table 5-8.

Table 5-8. “Pure” 238Pu mixture (Meyer 1970).

Mixture 

designation: Weight 

fraction 

(fresh)

Fresh 5 yr 10 yr 15 yr 20 yr 30 yr

Years of aging:a 0 5 10 15 20 30

Component Specific activity in mixture (Ci/g)

Pu-238 0.80 1.370E+07 1.317E+07 1.266E+07 1.217E+07 1.170E+07 1.081E+07

Pu-239 0.169 1.048E+04 1.048E+04 1.048E+04 1.048E+04 1.047E+04 1.047E+04

Pu-240 0.028 6.352E+03 6.349E+03 6.345E+03 6.342E+03 6.339E+03 6.332E+03

Pu-241 0.002 2.061E+05 1.620E+05 1.274E+05 1.001E+05 7.871E+04 4.864E+04

Pu-242 0.001 3.933E+00 3.933E+00 3.933E+00 3.932E+00 3.932E+00 3.932E+00

Am-241 N/Ab 0 1.463E+03 2.601E+03 3.484E+03 4.167E+03 5.094E+03

Component Activity ratiosc,d

Pu-238/Pu-239 N/A 1,307.3 1,256.7 1,208.0 1,161.3 1,117.5 1,032.5

Pu-238/Pu-240 N/A 2,156.8 2,074.3 1,995.3 1,919.0 1,845.7 1,707.2

Pu-238/Pu-241 N/A 66.5 81.3 99.4 121.6 148.6 222.2

Pu-238/Pu-242 N/A 3.483E+06 3.349E+06 3.219E+06 3.095E+06 2.976E+06 2.749E+06

Pu-238/Am-241 N/A 0 9002.1 4867.4 3493.1 2807.8 2122.1

a. Time since separation of Am-241 from the plutonium mix.

b. N/A = not applicable.

c. Calculate dose for Pu-239 and Pu-240 separately using these ratios.

d. “Pure” Pu-238 was not on site at LANL until after the introduction of RAS for analysis of bioassay samples. Therefore, 

the activity ratios for total alpha plutonium are not applicable.

At LANL, the predominantly pure 238Pu has always been processed in dry boxes and gloveboxes 

separate from those for processing WGPu to prevent cross-contamination of the two forms.

There is no definitive historical information on the 240Pu:239Pu atom ratios of LANL sources, and there 

is no information on how the ratios vary with time and location. However, before 1970, plutonium at 

LANL came from the Hanford Site (Gallaher and Efurd 2002). According to environmental impact 

studies, atom ratios from LANL plutonium that was released before 1970 would have ranged from 

0.01 to 0.03 240Pu:239Pu. LANL plutonium came from the Savannah River Site starting in 1970. The 

atom ratio of LANL plutonium that was released after 1970 was expected to range from 0.05 to 0.07, 

although ratios up to 0.13 indicated the sediment was affected by LANL plutonium (rather than 

deriving purely from global fallout). Ratios above 0.07 in environmental samples were indicative of a 

mixture of global fallout and LANL plutonium.

However, Table 5-9 describes the plutonium mixture that was found in 1960 in dirt that was used to 

sandblast plutonium parts during the demolition of a plutonium filter facility. Thus, although the 

240Pu:239Pu ratios in pre-1970 plutonium mixtures might have been nominally 0.01 to 0.03, use of a 6% 

fresh mixture for the pre-1970 years is not unreasonable. However, in general the pre-1970 mixtures 

of 3% listed in Table 5-4 are suggested when isotopic ratios are not known (Gallaher and Efurd 2002). 

Mixtures of less than 6% (i.e., 3%) are appropriate to assume during the Manhattan Project era (Lewis 

2006a). Intakes of plutonium in the early history of the site should be considered fresh. Aging of 

plutonium was addressed with continual reprocessing, and although the nature of the process work at 

LANL involved research-grade plutonium mixtures that are usually considered fresh, there is 

indication that some americium refining took place at LANL. For this reason, aged plutonium mixtures 

should be assumed when plutonium intakes resulted from decommissioning, other sources where 

long-term residual contamination might have been expected (Lewis 2006b), and when the source 

term is unknown. Assume 3% or 6% fresh plutonium mixture when specific information is available in 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 30 of 117

Table 5-9. Fresh plutonium mixture 

(Christensen, Garde, and Valentine 1975). 

Isotope Weight percent

Pu-239 93.5%

Pu-240 6.0%

Pu-241 0.5%

the case records (Lewis 2006a,b) to support this assumption. There was practically no americium 

present in the plutonium at LANL in 1944 or 1945 (Voelz, Grier, and Hempelmann 1985).

Any 241Am results that are observed in lung counts that were performed years after the intake can be 

assumed to be the result of (1) the ingrowth of 241Am from the 241Pu in the mixture over time or (2) the 

241Am in the initial plutonium mixture unless the incident report specifically indicates potential 

exposure to pure 241Am or positive urine bioassay for 241Am is found for the intake period.

Fuel-grade plutonium (12%) was not known to be routinely encountered at LANL. However, the 

Clementine reactor (which operated from 1949 to 1952) and the LAMPRE I reactor (1961 to 1963) 

used plutonium fuel. Plutonium doses for individuals who worked in these facilities should be based 

on the application of a 12% fuel-grade mixture in accordance with Table 5-6.

To summarize, dose reconstructors should make the following assumptions for plutonium isotopic 

mixtures.

 3% fresh plutonium mixture 1944-1956 for most plutonium operations. No aging should be 

assumed for weapons research;

 6% fresh plutonium mixture from 1957 onward for weapons-related activities. 6% 10 aged 

plutonium should be assumed for activities not related to weapons, including maintenance, 

decontamination, and exposures to plutonium source terms of an unknown nature; and

 Fuel grade (12% Pu-240) for reactor operations personnel with plutonium bioassay in the 

years 1949-1952 and 1961-1963. 

5.2.1.7 Correcting for Urinalysis Volume

Urine samples were corrected for estimated 24-hour excretion of plutonium based on sample volume 

and/or specific gravity (Gautier 1983). All results in the OTLBD have clearly listed units with 

associated uncertainties that are reported as 1 standard deviation unless otherwise specified in the 

record. Some samples that were taken in response to suspected acute intakes might have been spot 

samples of less than 24-hour excretion. These sample results are either noted in the database or, 

usually, the results are normalized to 24-hour excretion. In addition, the MDAs might not be in the 

same units as the activity results. As stated above, activity per 24-hr sample and activity per sample 

are not equivalent units. TIMS results in the OTLBD are typically listed in picocuries per 24 hours and 

picocuries per sample, with no listed sample size.

5.2.1.8 Calculating Picocuries per 24 Hours for Urine Bioassay Results

All plutonium bioassay results records have been processed through the current database before 

reporting. However, should the need arise to calculate plutonium urine bioassay analysis results that 

are not fully processed (for results outside those supplied with claim files), equations in Lawrence 

(1978) can be used to convert the listed units to picocuries per 24 hours. Method selection should be 

based on the era of the samples. These calculations could be necessary to resolve differences if 

claimant-supplied results differ from results in the database.
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5.2.1.9 Detection Sensitivities, Reporting Limits, and Tolerance Limits

Detection sensitivity was not a primary concern in the early years of operation of LANL. The concern 

was whether a tolerance limit was exceeded. Results below the MDA were originally reported as 

less-than values [listed as “LX.XX,” where X.XX is the MDA or detection limit (DL)] or “0” until the 

1980s; after which actual positive and negative results were listed in the database. 

In the current LABDR, all results from all periods have been listed as actual positive or negative 

values. When the MDA is not specified in the original result, it is considered to be 2 times the 

detection level (or uncertainty of the blank). Average chemical recoveries and matrix blanks were not 

routinely reported before 1957. Therefore, the values in the original dosimetry records for MDAs 

before 1957 were actually the results that would be reported if an analyzed urine sample had results 

that were in the upper 12% of the background data, assuming subtraction of the standard background 

count rate, application of average counting efficiencies, and chemical recoveries. These are the 

values that would have triggered evaluations of whether the value was positive; therefore, these are 

being treated as MDAs in this document (Lawrence 1978). However, values in Clark (2005) for 

average background count rates, recoveries, and count times, and the assumption that the entire 24-

hour sample was analyzed, would result in lower MDA values. Values were calculated from the 

referenced data using the MDA equation (HPS 1996). Both values are listed in Table 5-7. 

Bioassay results in the database might have accompanying MDAs if the sample volume is also listed. 

The MDAs might be listed as activity per sample rather than as activity per 24 hours; this might also 

be true for the sample results. If the listed MDA values are used, the units must be normalized to 

activity per 24 hours. In some cases in the historical data the Activity field value might be null or zero. 

In these cases the MDA value should be used. This MDA value is either a sample MDA, if known, or 

an a priori MDA for the counting system of that era. The Uncertainty field value is valid only when an 

Activity value is present. Zero values rather than MDA values can exist as real values, although in 

historical times zero values were often reported when no activity was counted. If the MDA column is 

null, an MDA value for that era is unknown. All MDA values in the LABDR are true MDAs and not 

critical levels (LCs).

MDA 3 4.66Sb

ETY

+ = (5-1)

where:

Sb = standard deviation of the background count

E = efficiency of the counter

T = count time

Y = chemical recovery

MDA = the MDA at 95% confidence

Before the start of RAS analysis in 1967, plutonium results were analyzed as total alpha plutonium. 

These results are listed as 239Pu in the LABDR, but should be treated as total alpha plutonium, 

representative of the combination of 238Pu and 239+240Pu. Results from TIMS after 1997 are only 239Pu; 

there is no contribution from 240Pu. Before use of the results in dose calculations, the 240Pu 

contribution can be added to the 239Pu and entered as 239Pu, or the dose reconstructor can make 

separate calculations for 239Pu and 240Pu.

Plutonium bioassay sensitivities as listed in procedures and reports are in Table 5-7. In addition, 

Table 5-7 lists reporting and tolerance limits for certain years. Tolerance limits were defined as the 

level that could be accepted as safe. Reporting limits are values above the MDA that are considered 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 32 of 117

significant for recording or follow-up. However, the database contains the actual bioassay results 

even if the results are below the reporting level. 

When MDA values are available in the database, these values are typically listed in units of activity 

per sample. Except for TIMS analysis, the sample mass is also listed. It is assumed that the persample results use the listed sample mass. While this mass is usually a mass that is approximately 

the 1,400-mL volume, which is the default excretion per day, these are not exactly equivalent. 

Appropriate normalization to 1,400 mL should be made.

To summarize comparison of sample results to MDAs,

 When results are in activity/24 hours and no MDA is listed, the dose reconstructor should use 

the applicable default MDA from Table 5-7;

 When MDAs are listed in activity/sample, the results are usually listed as activity/sample as 

well on a separate line, which makes the determination straightforward;

 When MDAs are listed in activity per sample, in some cases, the sample mass is listed as well, 

and instructions in the paragraph above should be followed to normalize to daily excretion;

 When MDAs are listed in activity per sample, and the result is not, the result may be neglected 

in the case of a non-detect, when the MDA is assumed instead; but

 If none of these options are applicable, the site lead for dose reconstruction should be 

contacted.

5.2.1.10 Validation of Samples

In the original database, sample results in a worker’s dose record might be marked as invalid because 

the results did not meet the statistical criteria for a valid sample. If results in the LABDR are being 

compared to original data (or to bioassay results in files in the original database format before June 

2005, which might still be in claim folders) and comments on the invalidity of the sample are noted, 

the dose reconstructor should review the sample information carefully and make a decision on the 

relevance of sample results. Section A.4 in Attachment A lists the protocol followed by the historical 

LANL computer software to validate a sample result. This section is provided for information only to 

enable the dose reconstructor to understand the criteria that were used to cause a sample result to be 

marked as invalid. This method should not be used by the dose reconstructor in the validation 

process. The current LABDR might list comments taken from the original database. These 

comments frequently relate to the Bayesian statistical methods currently in use at LANL and are not 

necessarily relevant to the dose reconstruction.

The excretion rate after an intake of plutonium typically remains consistent or gradually rises for many 

years after intake. An example is the original group of 26 individuals who received intakes in 1944 

and 1945 and still had measurable urine bioassay results in 1991 (Voelz and Lawrence 1991). 

Attachment A, Section A.10, contains urine bioassay and nasal swipe results) from an individual with 

intakes confirmed by autopsy. These results are an example of the possible variability of results for 

an individual with an actual burden of plutonium.

5.2.1.11 Sampling Protocol for Special Accidental Exposure

Sampling protocols for special and accidental exposures are listed in Attachment A, Section A.12. 

However, review of the current LABDR does not completely support the bioassay frequencies that are 

listed in the internal dosimetry program. An example is a sample marked as involving chelation. The 
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previous sample, taken 7 months earlier, appeared to be part of the routine program. The next 

sample after the chelation was 7 months later. There was no sampling in between, no initial sample 

to confirm the intake, and no information on the date or details of the intake, except a notation of 

“URINEFIX” on the sample, which is assumed to be the code for chelation.

The incident section in the internal dose file provided to the dose reconstructor might list information 

on an incident that relates to the results. These records can provide more precise information on 

known or suspected intake dates. In addition, the record of nasal swipes, positive or negative, might 

be available in the incident section. Tables A-19 to A-21 in Attachment A list the historical and current 

protocols for sampling for suspected plutonium exposures.

5.2.1.12 Nasal Swipes

Until the first urine bioassay analysis was perfected and available in February 1945, nasal swipes 

(sometimes called nose counts at LANL) were relied on to be a qualitative indicator of plutonium 

intake (Hempelmann 1946). After the development of urine bioassay techniques, nasal swipes were 

used to indicate the need for follow-up bioassay, although it was not always performed immediately 

after a positive nasal swipe (McInroy et al. 1991). The maximum permissible level (MPL) was 50 dpm 

alpha per nostril. Positive nasal swipes can aid the dose reconstructor in determination of a plausible 

intake date. Dose reconstructors should not consider the absence of activity above 50 dpm a reason 

to invalidate bioassay results because many circumstances can contribute to a negative nasal swipe if 

an intake has occurred. A less-than-MPL nasal swipe, however, might have precluded a follow-up 

bioassay. In addition, in 1944 LANL began the practice that if the two nasal counts varied significantly 

the higher nasal count was considered spurious (Hempelmann 1944) and the lower of the two counts 

was considered the true count (LANL 1994). Information about a positive nasal swipe can often be 

found in the incident section of the PDF file for the individual.

Intakes are not calculated from nasal swipe results, but the results of nasal swipes are used as 

indicators of possible intakes. Nasal swipes continue to be used in the present bioassay program as 

an indicator of possible intakes.

5.2.1.13 Computer-Based Calculations and Sorting

Beginning in 1959, LANL routinely used several computer-based systems to track plutonium bioassay 

and intakes. The first IBM-704 plutonium body burden report was made in September 1959. This 

program estimated the body burden of the employee and the amount of increase in body burden at 

6-month intervals (Lawrence 1978).

Of relevance to dose reconstruction, these computer programs allowed the tracking of individuals to 

ensure that those working with plutonium or frequenting plutonium areas were on the appropriate 

routine or special monitoring programs.

Details of these historical computer programs are in Section A.4 of Attachment A.

5.2.1.14 LANL Bioassay Data Repository

All plutonium data reside in the LABDR. Plutonium results, current and historical, have been 

converted to units of activity (typically picocuries) per 24 hours. All results are reported in actual 

values (i.e., positive, negative, or zero). Uncertainty values are listed for each measurement. Codes 

for the interpretation of the records are in the PDF file that is supplied for each individual’s results. 

Records of chelation therapy for plutonium might be noted in the comment field. However, the code in 

the original database that indicated that the sample was affected by chelation was not captured when 

the records were transferred from the original repository to the LABDR. Only comments with the 
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results were transferred. It is assumed that all chelation events would have comments, but that might 

not be the case. The standard comment notation is “URINEfix” with no information on intake date, 

etc. The dose reconstructor should carefully evaluate any grouping of high bioassay results. 

However, only approximately 35 workers have undergone chelation therapy in the history of the site. 

Therefore, if chelation therapy cannot be validated, chelation should not be assumed for positive 

bioassay results.

5.2.2 Americium

At LANL 241Am is usually encountered as a trace contaminant in plutonium mixtures. However, there 

is also potential for exposure to pure 241Am at LANL during actinide research activities and processing 

of plutonium mixtures. A procedure for determining 241Am in urine was in development in 1948 (LASL 

1949a). No exposure to pure 241Am was likely before the beginning of the americium bioassay 

program in 1954.

All americium data reside in the LABDR. Americium-241 results, current and historical, have been 

converted to units of activity (typically picocuries) per 24 hours. All results are reported in actual 

values (i.e., positive, negative, or zero). Uncertainty values might be listed for each measurement. 

Codes for the interpretation of the records are in the PDF file that is supplied for each individual’s 

results. Records of chelation therapy for americium might be noted in the comment field, usually with 

the notation “URINEfix.” However, as with the plutonium data, the code in the original database that 

indicated that the sample was affected by chelation was not captured when the records were 

transferred from the original repository to the LABDR. Only comments with the results were 

transferred. It is assumed that all chelation events would have comments, but that might not be the 

case.

There is no historical indication that workers participated in the americium bioassay program only if 

there was a potential for exposure to pure americium. Therefore, plutonium mixtures should not be 

inferred from americium bioassay results if no plutonium bioassay results are found for that period. At 

present, in vivo measurements might be performed to confirm the plutonium aging if plutonium 

bioassay results are above the MDA (Lewis 2006b). However, if the selected plutonium mixture 

(based on plutonium bioassay results) indicates the presence of americium, the absence of americium 

bioassay should not preclude the calculation of the dose from the americium contribution to the 

plutonium mixture.

Before the mid-1990s, very few individuals submitted urine samples that were analyzed for 241Am. 

Samples were submitted in response to incidents that involved exposures to sources of 241Am that are 

not contained in the plutonium mixture. Estimates of internal doses are based on the results of chest 

counts and urine bioassay (Inkret et al. 1998b). Urinalysis is the principal bioassay method for 

assessment of intakes. Publication 68 from the International Commission on Radiological Protection 

(ICRP) lists the respiratory absorption type for americium as type M (ICRP 1995). However, if 241Am 

is determined to be part of a plutonium mixture or ingrowth from the intake of a plutonium mixture, the 

absorption type for the plutonium mixture should be assumed for the 241Am (ORAUT 2007b).

The current routine and special sampling programs have protocols similar to those for the plutonium 

program (Inkret et al. 1998c). Table A-16 in Attachment A describes the routine sampling frequency. 

Table A-22 describes the current program of special sampling frequencies. The potential for intake 

can be indicated by a positive nasal swipe. A positive nasal swipe is indicated by 50 dpm alpha in 

either nostril. However, lack of a positive result on a nasal swipe does not eliminate the possibility of 

an intake.
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 35 of 117

Dose from 241Am intakes associated with plutonium mixtures can be reconstructed when associated 

plutonium data are available. Dose from pure sources may be estimated by bioassay data when they 

are present in claim records.

Minimum Detectable Activities

The historical practice emphasized being below the tolerance level rather than being below the MDA. 

All bioassay results are reported in picocuries per 24-hour sample; therefore, the listed urine bioassay 

MDAs in Table 5-10 have been normalized to picocuries per 24-hour sample. MDAs in the LABDR 

are listed in units of picocuries per aliquot. However, if no volume or mass for the aliquot is given, the 

sample volume should not be considered the aliquot volume. Results below the MDA were 

historically reported as less-than values (“LX.XX”, where X.XX is the MDA) or “0” until the 1980s. In 

the current LABDR all results from all periods have been listed as actual values, positive or negative.

When not specified, 2 times the detection level (or uncertainty of the blank) is considered the MDA. 

Table 5-10 lists MDAs. MDAs are based on MDAs listed in procedures and reports.

If no MDA value is provided with the sample or the MDA provided with the results cannot be 

normalized to a 24-hour volume, dose reconstructors should use the values in Table 5-10 to 

determine if a listed bioassay result is above the MDA or to calculate missed dose using detection 

levels.

Table 5-10. Americium-241 bioassay techniques and sensitivities.

Sample type Year Method MDA

Urine 1954–1957 Unknown 9.E-01 pCi/24 hra

Urine 1958–1982 Chemical extraction/ 

proportional counting

2.E-01 pCi/24 hrb

Fecal 1977 Phoswich 4.E-02 nCi/sample

Urine 1983c–2002d Coprecipitation/alpha spece 1.5E-02 pCi/24 hr

Fecal 1983c Am/Pu screening/phoswich 1.E-02 nCi/sample

Urine 2003–present RAS 1.0E-02 pCi/24 hr

a. No MDA available, use derived investigation level; tolerance level 3.1 pCi/sample (LASL 

1954).

b. McClelland (1958); method can carry over thorium, plutonium, curium, actinium, and 

neptunium. Exact start or end date of this MDA is not known.

c. Gautier (1983); exact start date of the MDA is not known.

d. Sent to an offsite laboratory from 2000 to 2002.

e. Inkret et al. (1998d).

5.2.3 Tritium

Tritium bioassay started in 1950. The program was formalized in 1952 (Clark 2005). 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. 

Each year approximately 100 individuals are monitored for tritium intakes at LANL. If the work 

location was in the Health Research Laboratory (HRL) and the Ion Beam Facility (IBF) Buildings, the 

predominant form of tritium was OBT. Stable MTs were handled in TA-35 (Building 1) and the TSF 

(TA-21-209, Room 179). Outside these facilities, in cases where exposures to OBT or MT are not 

implied by case-specific information, assume HTO. Guidance for estimating doses for OBT and SMT 

is available in ORAUT-OTIB-0066, Calculation of Dose from Intakes of Special Tritium Compounds

(ORAUT 2007a).

The potential for tritium intakes at the cyclotron and Van de Graaff accelerators might have existed 

before the beginning of the bioassay program in 1950. Site locations for potential exposure to tritium 

are in Table A-8 in Attachment A. Bioassay results for tritium are usually available in the LABDR. 

However, if only whole-body dose is listed in the dosimetry records and the bioassay results are not 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 36 of 117

available, note that all dose results from 1950 to 1988 were reevaluated by J. N. P. Lawrence in 1989 

(Clark 2005) using ICRP Publication 30 parameters (ICRP 1979).

5.2.3.1 Organically Bound Tritium

Information on potential exposures to OBT is not complete, but locations such as the HRL and IBF 

Buildings used labeled compounds as the principal type of radionuclides in these facilities. Guidance 

on the calculation of doses from OBT is given in ORAUT-OTIB-0066, Calculation of Doses from 

Intakes of Special Tritium Compounds (ORAUT 2007a).

5.2.3.2 Metal Tritides

Tritium exposures in the form of MT aerosols were possible. The compounds include the chemical 

hydrides and dihydrides of hafnium, erbium, titanium, zirconium, and other metals. Uranium beds 

were used to capture tritium, so uranium tritide was also present (Nasise 1988, p. 4). Information on 

building locations and years of operation when MTs might have been encountered is not complete, 

but two facilities, TA-35 (Building 1) and the TSF (TA-21-209, Room 179), were specially designed to 

handle highly reactive tritides. Glovebox lines were used to contain the tritides, and a recycling 

system was used to capture escaping tritium (Nasise 1988). Experimentation with MTs required dry, 

inert environments in order to prevent contamination of the material under experimentation with water 

from the environment. 

The High-Pressure Tritium Laboratory (HPTL), was an early tritium facility (TA-33-86) that used openfaced hoods, rather than gloveboxes (LANL 1988, Horak 1995). The sealed apparatus in the hoods

was the primary confinement, and the protection provided by the hoods was secondary. In contrast, 

the facility for handling MT built in 1953 (TA-35-2) did use a dry glovebox system (Harper and Garde 

1982, p. 3). No specific information on the form of the tritium processed in the Weapon Subsystem 

Laboratory, which used hoods that could be completely enclosed, has yet been located.

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, claimant telephone 

interviews can provide indications that a person was exposed to MTs.

Interpretation of codes and units of the bioassay data for tritium is discussed in Attachment A, 

Section A.2.

5.2.3.3 Analytical Techniques

From startup of monitoring in 1950 until 1969, HTO in urine was analyzed by hydrogen evolved by 

dropping the urine sample on calcium carbide, collecting the hydrogen in a vacuum, and drawing the 

hydrogen into a glass electrometer chamber. In 1952, a gross GM-type system replaced the glass 

electrometer. In 1954, the tolerance for tritium in urine was 250 μCi/L (McClelland and Milligan 1954). 

The analytical range was 1 to 250 μCi/L. In 1958, an internal GM counting technique was used. The 

reported MDA for this method was 1 μCi/L. In 1970, liquid scintillation counting (LSC) began and is 

still in use today. With the introduction of LSC, the DL was 1 μCi/L. The MDA consistently improved.

For tritium results, the denominator used for reporting purposes was per liter of urine. In 1992, all 

data from 1950 through 1992 were converted to microcuries per liter. Tritium bioassay results can be 

found as early as 1946.

Results of samples were initially listed as “0” or coded as “LX.XX” (where X.XX is the MDA) to indicate 

that the value was less than the sensitivity or reporting level of the analysis. In the late 1980s, the 

practice of recording the result (positive or negative) was begun, although as late as 1990 “0” was still 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 37 of 117

being entered into bioassay results when the value was below minimum detection levels (Lawrence 

1990a). MDA values are listed in Table 5-11 and summarized in Attachment A, Table A-24. All 

results in the LABDR are likely to be reported in actual values (positive, negative, or zero).

Table 5-11. Tritium urine bioassay sensitivity levels (μCi/L).

Time Detection level Reporting levela Counting method

1950–1951 1b Electroscope

1952–1953 1b GM counter

1954–1957 1

1958–1968 1 Internal GM

1969c–1987 0.02a,c 1 LSC

1988–1998 0.01a 0.1 LSC and 1 mL raw urine

1999–present 0.005 LSC

a. Use the reporting level as the MDA because results below these values were not 

reported (Clark 2005).

b. Expected to be the same as 1954.

c. Gautier (1983).

5.2.3.4 Routine Sampling

The routine sampling protocols for tritium bioassay throughout the history of the program are listed in 

Attachment A, Table A-17. Historically, persons working with at least 1 Ci of tritium were required to 

participate in a bioassay program with sampling every 2 weeks. In 1971, the program was changed to 

determine the frequency of sampling based on the level of tritium in the previous urine sample (Clark 

2005).

In 1998 the required turnaround times for tritium samples of all priorities was final completion within 

10 days. The preliminary report was due on priority 1 and 2 samples within 3 hours or by Monday 

morning (Inkret et al. 1998c). Inkret et al. did not specify the definitions of priorities 1 and 2.

5.2.4 Uranium

Bioassays were initiated in 1949 with analyses for uranium mass as an indicator of exposure (Clark 

2005). Historical uranium at LANL was primarily either depleted uranium (DU; often referred to at 

LANL as D-38 or 28) or enriched uranium (EU; referred to as Or-93 or 25) with a variety of isotopic 

ratios (Little, Miller, and Guilmette 2003a). Uranium exposures from 1950 to 1957 were expected to 

be primarily from DU because lathing and milling operations took place in large rooms with no exhaust 

systems. EU work was conducted in dry boxes, so the potential for intakes was lower (Clark 2005). 

Welding on DU was introduced in 1951. However, ventilation and other measures prevented serious 

overexposures (Shipman et al. 1952). Engineering controls, especially ventilation, were improved in 

the late 1960s and reduced the potential of intakes of uranium (Lewis 2006c).

Conversion factors differ among DOE sites because the fractional activity of the 234U differs in the 

enrichments. Therefore, 0.17% enriched DU at LANL might have a different amount of 234U than the 

same enrichment at a different DOE site, dependent on the source of the DU and the processing. 

Tables 5-12 and 5-13 describe the nominal weight composition and fractional activity for typical DU 

and EU mixtures, respectively (Lawrence 1990b). These are the typical enrichments said to be used 

at LANL. These tables should be used unless there is specific information that states that the 

enrichments differ significantly from the enrichments listed in Tables 5-12 and 5-13. It should also be 

noted that the PDF files that were provided with worker bioassay data list a factor of 3.33 × 10-7 Ci/g 

for 238U. This is the conversion factor for pure 238U only; it does not apply to the total alpha uranium 

results that are reported, especially for the uranium analysis using delayed neutron activation analysis 

(DNAA; UNAA for uranium analysis) as discussed below. The 3.33 × 10-7 Ci/g factor should not be 

used for conversion of the results to DU or EU.
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Table 5-12. Nominal weight composition and fractional activities for D-38 uranium

(Lawrence 1990b).

Measure U-234 U-235 U-236 U-238 Total

Weight fraction 0.00002 0.003 0.000003 0.997

Curies in 1 g D-38 1.251E-07 6.489E-09 1.942E-10 3.355E-07 4.673E-07

Isotopic specific activity (pCi/μg) 6.25E+03 2.16E+00 6.47E+01 3.37E-01 4.673E-01

Fraction of total activity 0.2677 0.0139 0.0004 0.7180

Table 5-13. Nominal weight composition and fractional activities for OR-93 uranium

(Lawrence 1990b).

Measure U-234 U-235 U-236 U-238 Total

Weight fraction 0.011 0.933 0.002 0.054

Curies in 1 g Or-93 6.879E-05 2.018E-06 1.295E-07 1.817E-08 7.096E-05

Isotopic specific activity (pCi/μg) 6.25E+03 2.16E+00 6.47E+01 3.37E-01 7.096E+01

Fraction of total activity 0.9695 0.0284 0.0018 0.0003

DU (D-38) is the most common form of uranium that has been encountered at LANL (Inkret et al. 

1998a), although natural uranium (NU, also called Tuballoy) was used in conventional weapons 

testing from 1949 to 1970. The most common chemical forms are oxides and metal. (T3O8 can refer

to an oxide of NU.) However, LANL has always treated uranium as either solubility class D or W. 

Urine assay data suggested that “all known LANL exposures to uranium were to a relatively soluble 

form (not class Y)” (Lawrence 1984, 1990b). The partition between classes D and W could not be 

determined. Class W was historically used for reporting results (Lawrence 1992a) because it 

produced larger doses.

In addition, intake was calculated by LANL based on class D mass of uranium, for assaying in relation 

to the nephrotoxic limit for uranium. The nephrotoxic limit is not relevant to dose reconstruction. 

However, the dose reconstructor should evaluate the data to determine the absorption type that is 

most appropriate and favorable to the claimant. Statements about welding DU and eventually EU, 

which would have been a potential for more insoluble forms of uranium, are in Shipman et al. (1952).

The LANL program calculated intakes for individuals with at least one positive urine analysis result. 

The potential for intake was indicated by a positive nasal swipe (50 dpm in either nostril). Lack of a 

positive result on a nasal swipe should not eliminate the possibility of an intake, but follow-up 

bioassay was not likely to have been performed as a result of a nasal count of less than 50 dpm.

Otherwise, dose reconstructors can use the nominal compositions in Tables 5-12 and 5-13 as default 

compositions when no other information is available on the enrichment of the uranium (listed as either 

mass or weight fraction of 235U or as a weight percent when multiplied by 100). The values in Tables 

5-12 and 5-13 are the compositions at LANL at least from 1970 through 1990 (Lawrence 1990b). In 

addition, in Lawrence (1990b) the presence of 236U in the isotopic mixture is indicated. The 236U is 

indicative of recycled uranium (RU). However, LANL did not process uranium except to extrude or 

machine it. Oak Ridge National Laboratory (ORNL) was the source of the uranium that was 

processed at LANL.

5.2.4.1 Sampling Protocol

Historical sampling protocols are listed in Attachment A, Table A-18. The program was initiated in 

1949 with analysis for uranium mass as an indicator of exposure. Samples were collected on Fridays 

just before workers left the site to maximize the sensitivity of bioassay detection (Lawrence 1992a). 

Since 1983, spot samples have been collected as far removed from the potential exposure time as 

possible, usually Monday mornings before workers entered the work area. The sample might have 

contained the last voiding on a Sunday evening and/or the first voiding on a Monday morning if extra 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 39 of 117

volume was necessary. This protocol ensured that the large fraction (approximately 80%) of rapidly 

excreted uranium, experienced on the first day after an intake, was excluded from the sample 

(Lawrence 1984). Beginning in July 1993, the larger volume of sample that was required for the alpha 

spectroscopy method of analysis made the Sunday evening/Monday morning collections standard 

protocol (Lawrence 1992b).

In 1992, retroactive calculations were made by Lawrence of the data beginning in 1949. Data for 

1951 were not included in these calculations because the bound Los Alamos Notebooks could not be 

found. Clark (2005) suggests that these notebooks were never found and, therefore, 1951 data are 

missing from the records. However, review of the current database indicates a number of uranium 

bioassay records for 1951 similar to the number of records in adjacent years. Therefore, it can be 

assumed that no data are missing (Buddenbaum 2006). For enriched 235U, the conversion of data 

into the computer files and retroactive calculations started with the 1955 data. Assays that were 

identified as 238U were converted to micrograms per day with values less than the MDA or zero set to 

the minimis MDA. Assays that were identified as 235U were converted to picocuries per day with 

values less than the MDA or zero set to the minimis MDA. In the LABDR, the micrograms-per-day 

and picocuries-per-day results have been converted to micrograms per liter and picocuries per liter, 

respectively. These results should be normalized to a 24-hour day using an assumed volume of 

1,400 mL/d of urine.

5.2.4.2 Uranium Analysis Techniques

Several analysis techniques were employed over the history of LANL. Techniques have included 

uranium fluorophotometric (UF), ion exchange, or extraction chemistry with radiometric alpha 

proportional counting (UR), DNAA (or UNAA), and RAS. UF and UR were total uranium techniques 

with the enrichment selected because of workplace knowledge. The uranium enrichment for 

DNAA/UNAA results was also determined by workplace knowledge.

Table 5-14 summarizes the routine urinalysis detection levels for various periods. The results of urine 

bioassay for uranium have been entered in the database. The results in 1991 or before are likely to 

be reported in actual values (positive, zero, or negative) (Lawrence 1992a). Results below the MDA 

might be listed as zero or blank.
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Table 5-14. Routine uranium urinalysis detection levels.

Period Method MDA DL

Reporting 

levela

1949–1967 UF (DU or NU) None listedb 50 μg/Lc >100 μg/L

1968–02/1976 UF (DU or NU) 4 μg/Lb,d U

03/1976–1978 UF (DU or NU) 1 μg/Ld U

1949–1954 Anion exchange/gross alpha counting 

(possibly used)

25 dpm/L >100 dpm/L

1955–1971 Extraction/alpha proportional counting 

(U-234 alphas measured)

Unknownb,e 50 dpm/Lc >100 dpm/L

03/1971–1976 Extraction/alpha proportional counting 15 dpm/Lb >100 dpm/L

06/1976–01/1977 Extraction/alpha proportional counting 10 dpm/Lb

02/1977–12/1977 Extraction/alpha proportional counting 4 pCi/L

1978–07/1982 UNAA/delayed neutron countingf

Only U-235 counted

1 μg/L DU 

0.17% enriched

07/1982–06/1991 UNAA/delayed neutron countingf

Only U-235 counted

4 pCi/L EU

93% enriched

1982–06/1991 UNAA/delayed neutron countingf

Only U-235 counted

4 μg/L DU 

0.17% enriched

1982–1992 Anion exchange/colorimetric (used to 

confirm UNAA >30 μg/L (not a 

standard method of analysis)

1 μg/50 cm3

07/1991g–1998 RASh for each isotope 0.1 pCi/L 0.05 pCi/L

1998–present RASh for each isotope 0.008 pCi/L 0.004 pCi/L

(1 standard 

deviation)

a. Exceeding reporting levels required investigation and evaluation (Lawrence 1984).

b. Lawrence (1984).

c. Use as detection level (MDA = 2 × DL) because this level would be used to determine positive results.

d. 50 μg/L was the level at which LANL considered the result to be a positive indication of NU material in the body; at 100 

μg/L, LANL identified the need for a workplace evaluation (Dummer 1958).

e. Specific for U-235 and U-233, 50 dpm/24-hour sample considered positive indication of EU in the body (Dummer 1958).

f. See Section 5.2.4.3 for full explanation of the interpretation of UNAA results including minimum reporting levels.

g. Results for July 1991 through December 1991 might be listed as UNAA results in the LABDR. RAS values were used 

directly in 1992 (Lawrence 1992b).

h. Can measure U-234, U-235, and U-238. Alpha spectrometry cannot differentiate between U-233 and U-234.

5.2.4.3 Delayed Neutron Counting

DNAA (also known as UNAA in the LABDR) was the analytical technique LANL used between 1978 

and 1992. The technique might have been implemented as early as 1962 (ORAUT 2009a), but UNAA 

was not in general use until 1978. Interpretation of the results during this time was dependent on 

identification of the radioisotopic mixture to which the person might have been exposed (Lawrence 

1992b). Because the person’s work history might put them with D-38 one day and Or-93 the next, 

uncertainty can be introduced into the results.

The DNAA method depends on counting the number of delayed neutrons produced by fission of 235U 

after thermal neutron activation. This method also fissions 239Pu and 233U but, according to Gautier 

(1983), 239Pu is not an interference unless the concentration in the sample is greater than 300 pCi/L. 

Uranium-233 was not a significant intake potential at LANL.

The reported results of the DNAA method for EU analyses are labeled 235U in the records, but have 

always included the alpha activity of 234U that is about 34 times that of 235U. Results for DU analyses 

are similarly labeled 238U, but have always been the total uranium mass based on the assumed 

isotopic mass fractions of 0.00002 234U, 0.0017 235U, and 0.9983 238U (Lawrence 1992a).
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 41 of 117

The DNAA/UNAA was designated as a screening procedure and not a quantitative measurement. 

(Although, except for the confirmatory procedure for results above 30 μg/L, there were no other listed 

quantitative procedures for that period). At the time when DNAA/UNAA was initiated, the uranium 

exposures had been reduced due to enhanced workplace ventilation units. The ventilation units drew 

off most of the airborne particulates from machining and grinding operations so the higher exposures 

in the 1950s and early 1960s were eliminated (Lewis 2006c).

In the DNAA measurement, calibration curves were made up of standards of EU (93% 235U) and DU 

(0.17% 235U) (Gautier 1983). These calibration curves were plotted (labeled) as delayed neutron

counts versus 235U pCi/L or 238U μg/L. For the EU plot, concentration was in terms of total activity of 

EU. For the DU plot, concentration was in terms of total mass of DU. The calibration curve was 

selected based on the type of work, and then the concentration of uranium was interpolated.

When a worker was first entered in the system, it was noted on the health physics checklist whether 

the worker worked with EU, DU, or both. This information was transferred to the sample kit label as 

“235,” “238,” or “235/238.” Absolutely no indication of the fraction of time or work location would have 

been recorded.

If the worker had a “235” on the sample kit, the EU calibration curve was used and an activity result 

that was labeled as 235U was calculated. If the worker had “238” on the sample kit, the DU calibration 

curve was used, and the mass result that was labeled 238U was calculated. If the worker had 

“235/238” on the sample kit, both curves were used to determine results and both a 235U by activity 

result and a 238U by mass result were calculated. No fractionation of the results is assumed when 

both isotopes are listed. The dose reconstructor must either use the result or the MDA most favorable 

to the claimant or make a judgment on the appropriateness of either DU or EU dependent on 

telephone interviews or DOE records.

The DNAA method actually measures the amount of 235U in the sample. Only if the ratios of 238U/235U 

and 234U/235U are well known can the uranium mass and activity in the sample be accurately 

determined (Lawrence 1992a). By the DNAA method, the minimum reporting value was 4 μg/L for DU 

(0.17% enrichment) and 4 pCi/L for EU (93% enriched). For D-38 (DU), a 4-μg/L sample would 

contain 0.0147 pCi/L of 235U [(4 μg/L × 0.0017 × 2.163 × 10-6 Ci/g × 1 × 1012 pCi/Ci)/(1 × 106 μg/g)]. 

The enrichment of 235U assumed for DU at this period was 0.17%. A 4-pCi/L sample of EU (93%) 

would contain approximately 0.116 pCi/L of 235U [0.116 pCi/L ÷ 0.0284 (from Table 5-13) = 4 pCi/L]. If 

the person’s actual urine sample contained 0.116 pCi/L of 235U, the reported value for EU would be 

4 pCi/L. If the analysis was designated for both EU and DU, the D-38 value would be reported as 

approximately 32 μg/L [(0.116 pCi/L ÷ 0.0147 pCi/L) × 4 μg/L] and was labeled U-238 (Lawrence 

1992b). Uncertainty about the material of exposure, whether enriched or D-38, could introduce as 

much as a factor-of-4 error in the evaluation of a worker’s dose (Lawrence 1992b).

Uncertainty for this method was considered to be 96% ±6% at 5 to 26 pCi/L for EU and 93% ±8% at 

4 to 17 μg/L for DU.

This method was eventually deemed operationally unacceptable because of long turnaround times 

and the need to identify the mixture. LANL changed to RAS techniques between July 1991 and 

January 1992. Extreme care should be used in interpreting the results of bioassays between these 

dates. Normalization might have been done to RAS results to maintain consistency with the DNAA 

results, as discussed in footnote g to Table 5-14.
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5.2.4.4 Radiometric Alpha Spectroscopy

This method consisted of a full chemical extraction process for the uranium from the urine. A tracer of 

232U was added to determine chemical recovery. The extracted and plated uranium was counted by 

RAS and the activities of 232U, 234U, 235U, and 238U were determined (Lawrence 1992a).

Several problems were initially encountered with this method. No chemical or synthetic urine blanks 

were used to establish the MDA. The counting blank was a blank planchet rather than a reagent 

blank, which would produce a lower background. Chemical recoveries ranged from 10% to 60%. 

Some analyses indicated isotopic ratios outside the realm of probability. The computer program 

converted all negative values to zero. All 1991 data yielded negative values for 235U rather than the 

expected negative and positive values. In early 1992, the computer program was still not modified to 

correct for chemical blanks.

Results that are listed as RAS in the current database, which are isotopic, should be evaluated 

according to the type of enrichment that is determined for the intake. Table 5-15 provides the factors 

to be applied to the isotopic bioassay results to estimate the total uranium as either DU, NU, or EU.

Table 5-15. Factors to convert isotopic activities 

to total uranium.

Radionuclide

Activity factora

DU NU EU

U-235 N/A N/A 35.211

U-234 3.736 2.058 1.031

U-238 1.393 2.033 N/A

a. N/A = not applicable and indicates that the fraction of 

the isotope is too small to be used as an indicator for 

the given enrichment.

5.2.4.5 Excretion of Environmental Levels of Uranium

NU from nonoccupational intakes (primarily food and water) can be excreted in urine at levels above 

the analytical MDAs for either the elemental uranium analysis or the alpha spectrometry analysis. 

The 234U:238U ratio can be used to distinguish DU from NU. For the purposes of dose reconstruction, 

environmental uranium is not subtracted from monitoring results, and all uranium is assumed to result 

from occupational exposure. Therefore, this section is for information only. Table 5-16 lists activity 

ratios for NU.

A 1992 study by Little, Miller, and Guilmette (2003a) listed the average drinking water concentrations 

for the Los Alamos/White Rock/Santa Fe area as 0.015 μg/L (0.01 pCi/L or 0.00037 Bq/L). Therefore, 

with the use of 0.01 as the factor of drinking water concentration to excretion concentration (Little, 

Miller, and Guilmette 2003a,b), LANL used the urinary excretion values of 0.00015 μg/L for elemental 

Table 5-16. Characteristics of NU (Tuballoy).

Measure U-234 U-235 U-236 U-238 Total

Weight fraction 0.0000537 0.0072 0 0.99274

Specific activity (pCi/μg) 0.33367 0.01557 0.33367 0.68291a

Fraction of total activity 0.4886 0.0228 0 0.4886

a. As listed in IMBA Version 1.0.42.

analyses, 0.0001 pCi/L for 234U and 238U, and essentially anything that was detected for 235U to 

distinguish between natural background and potential occupational exposure for uranium. In some 

cases, specific information in the worker’s file indicates that the excretion was from natural sources.
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 43 of 117

New Mexico is known for high levels of NU in the soil and groundwater. However, some LANL 

workers lived in areas of particularly high NU concentrations that ranged from 0 to 4 Bq/L (108 pCi/L) 

in 1992 and up to 6 Bq/L (162 pCi/L) in 2001 (Little, Miller, and Guilmette 2003b). These areas of 

high concentration are primarily in the Española area.

Beginning in 1992, all workers who participated in the uranium bioassay program were requested to 

bring samples of their drinking water to LANL for analysis with each urine bioassay sample (at least 

once each year). The concentrations were found to vary widely even for individuals. The results of 

the analyses of these water samples are listed with the bioassay data for the individual. When water 

sample results are noted in the record, LANL multiplied the results by 0.01 to calculate the 

concentrations of uranium isotopes to be subtracted from the individual’s bioassay results before 

calculation of an occupational intake. The factor is based on the consumption of 1.1 L of water per 

day (Little, Miller, and Guilmette 2003b). When LANL assesses dose or potential occupational 

intakes, the concentration of the most closely associated water sample might be subtracted from the 

bioassay sample results rather than an average value for all the drinking water samples for the 

individual. (There is no evidence that the individual bioassay results for the individuals in the 

database have had the drinking water concentrations subtracted. Therefore, these results should be 

used as found.) Other variables, such as drinking habits, dietary components, and individual 

physiological differences, can influence the individual’s baseline excretion rate of environmental 

uranium.

Background excretion of uranium in feces probably varied over an even larger range than urinary 

excretion. Fecal samples were rarely obtained for potential uranium intakes; when they were, the 

investigation report should discuss how the results were interpreted.

5.2.5 Polonium

Polonium for use in initiators was handled in quantities of 100 Ci/mo or more starting in February 1945 

and increasing to 500 Ci/mo by December 1945. Before that time, smaller quantities were handled at 

in TA-1 at H and Gamma Buildings (ENSR 2002). DP East, Buildings 151, 152, 153, and later (1949) 

155 were also used to process polonium for initiators.

Work at LANL with 210Po ceased in 1959. After that time, polonium was encountered only in the form 

of sealed Po-Be sources. After 1959, there were two incidents in which sealed Po-Be sources broke 

open; personnel involved in the incidents submitted urine samples for bioassay. Unless a worker was 

involved in the broken source incidents, no intakes should be calculated after 1959.

Work with 210Po was of a limited scope at LANL. Therefore, missed dose for polonium should be 

assessed only if actual polonium bioassay was performed for the individual or if there is indication, 

through claimant interview, that the employee was actually exposed to polonium. Potential missed 

dose should be assessed only during the interval covered by the bioassay or work period. In 1945, 

only two persons exceeded the 1,500-cpm/24-hr sample tolerance for polonium (ENSR 2002). 

However, University of California (1977a) stated, “During the 1940s and early 1950s, 210Po exposures 

at LASL occurred with some regularity. There was no method in routine usage to determine the 

actual exposure. (Exposure was controlled from raw urine assay data.)” Dummer (1958) stated that 

the possibility of exposure should be investigated at greater than 10 dpm/L in a urine bioassay 

sample.

Hempelmann (1946) stated that, because of the ease of the analysis, polonium urine bioassay was 

performed routinely and frequently on all individuals who worked with polonium. These records are 

expected to be in the current database.
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The initial urine bioassay analysis procedure could not adequately distinguish between plutonium and 

polonium. During 1944, the total alpha results were assigned to either plutonium or polonium based 

on the individual’s work history. The procedure was modified in the fall of 1944 to extract the 

plutonium. Hempelmann (1944) indicates that this modified procedure was originally developed at 

Monsanto. LASL (1945) lists the procedure and states that the recovery of polonium on the copperplated disk was expected to be greater than 80%. Gautier (1983) contains a procedure for 210Po in 

urine with an effective date of May 1, 1955, that states that the average recovery of standards was 

90% ±3% at the 15-pCi concentration. Discussions with a retiree indicate that no correction for 

recovery was included in the calculations [1]. Despite the assumption of 100% recovery by LASL 

personnel, based on the recovery factor that has been determined to apply to samples analyzed by 

the Monsanto method (ORAUT 2013a), the 10% recovery correction is applicable to LANL results as 

well.

Detection limits for routine urinalysis are listed in Table 5-17. A procedure is listed for 210Po in urine in 

LASL (1954) and Gautier (1983). The MDA continues to later periods unless changed.

Table 5-17. Routine 210Po urinalysis detection levels.

Period MDA Recheck Tolerance limitb

1943–1952 440 dpm/24 hra

1953 50 dpm/24 hra

1954 10 dpm/L 100 dpm/L 500 dpm/L

1955 0.1 pCi/L

1958 100 dpm/L 500 dpm/L

a. LASL (1979).

b. Workers exceeding the tolerance limit were removed from exposure 

potential until the levels dropped below the tolerance limit.

5.2.6 Other Limited-Exposure Radionuclides

LANL 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 LANL. Although some documentation has been found on bioassay 

for these nuclides, the records are not provided to the dose reconstructor as part of the case-specific 

information. 

5.3 IN VIVO BIOASSAY MINIMUM DETECTABLE ACTIVITIES, ANALYTICAL METHODS, 

AND REPORTING PROTOCOLS

In vivo counting equipment and techniques were developed in the late 1950s and have been in 

routine use for measuring X- and gamma-ray-emitting radionuclides since 1970, and possibly as early 

as 1960. There is some indication that some of the counts between the beginning of the program in 

1955 (Van Dilla 1959) and the 1960s were performed for development of the program rather than 

actual suspected intakes. Counts during this period should be evaluated as closely as possible for 

validity in the dose reconstruction. No in vivo results are expected to be found in the LABDR before 

1969.

5.3.1 Whole-Body Counters

The first whole-body counter to be used at LANL was the HUMCO I (Ennis 2003), which became 

operational in 1955 (the name stood for human counter). The counter consisted of a large double 

cylinder with a liquid scintillation fluid (possibly trichloroethylene) that filled the annular space between 

the cylinders. The scintillation fluid was viewed with an array of 5-in. photomultiplier tubes on the 

outside wall of the cylinder. The individual was placed inside the count chamber. The count rate was 

compared to the background count rate. The system typically used two energy windows: one for 40K 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 45 of 117

(1 to 2 MeV) and one for 137Cs + 40K Compton counts between 0.5 and 0.8 MeV. The result was 

obtained by subtracting the contribution of 40K. The system was not used for photons below 100 keV. 

This system was used to screen individuals who might have been exposed to fission products at the 

reactors or in flyovers during weapons testing. It was also used to detect the bremsstrahlung from 

90Sr intakes. The energy resolution of these counters was poor. When an elevation of the 

background in a region of interest was observed, the individual was referred for screening with either 

the shadow-shield or full-shield 4- by 8–in. NaI(Tl) crystals (Healy 1970). The sensitivities of the 

NaI(Tl) crystals were approximately the same as those for the HUMCO, but the count time was 

significantly longer.

The HUMCO II became operational in 1958. The HUMCO II was housed in a count room (SB-16) 

made of 7-in. pre-World War II steel. The resolution was improved, but it remained a screening 

counter. Evaluations were made using the shadow shield or NaI(Tl) crystals. The 1958 annual 

progress report for the H-1 Group discusses “significant” measurable 65Zn in cyclotron workers 

(Buckland 1959).

In 1970, an in vivo counter capable of measuring four separate regions of the body began operation 

(Vasilik and Aikin 1983). Twin phoswich (CsI and NaI) detectors were placed over the lungs. The two 

layers of the detector were capable of simultaneously, yet separately, monitoring chest burdens for 

10- to 250-keV photons (NaI), for plutonium and uranium isotopes and 241Am, and for 200- to 

2,000-keV photons (CsI) for a qualitative assessment of a variety of fission and activation nuclides. A 

planar hyperpure germanium (HPGe) detector monitored the region between 10 and 250 keV with 

excellent energy resolution and could be positioned over the liver or thyroid as needed. Last, a 

lithium-drifted germanium (GeLi) detector (which was later replaced by an HPGe detector) was 

positioned under the prone subject. This detector was primarily for whole-body assessment. This 

system could both identify radionuclides and quantify the burdens. In 1990, the Tiger Team 

Assessment report indicated that the phoswich detectors were not state-of-the-art and could not 

accurately measure plutonium and americium in the presence of interfering gamma emitters, such as 

137Cs (DOE 1991). The phoswich and germanium systems were operated concurrently during the 

period from 1998 to 1999. The twin phoswich detectors were replaced by twin three-detector arrays 

of HPGe detectors in 1999. Two of the six HPGe detectors were used when a thyroid count was 

required.

5.3.1.1 Minimum Detectable Activities and Decision Levels

Operationally, an observed signal must exceed the LC, formerly referred to as the minimum significant 

measured activity (MSMA), to result in the decision that some activity has been detected in the lung, 

body, or other organ; this is also called the decision level (DL) (Vasilik et al. 1984). The MDA, 

formerly the minimum detectable true activity (MDTA), is the smallest amount of activity that can be in 

the lungs or organ such that a measurement of an individual can be expected to imply, correctly, the 

presence of activity with a predetermined degree of confidence. The MDA (or MDTA) and LC (or 

MSMA) values are listed in Table 5-18 for various years of operation. The MDA and LC values for 

Table 5-18. Routine whole-body counting detection levels (Vasilik et al. 1984; 

Ennis 2003).

a

Period Nuclide LC (nCi) MDA (nCi)b,c

1955–1958d Cs-137 8

Sr-90e 30

1959–1970d Cs-137 4

Sr-90e 30

1971–1984 Be-7 0.9 1.8

Cs-134 0.9 1.8

Cs-137 0.9 2.1
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Period Nuclide LC (nCi) MDA (nCi)b,c

Co-57 2.1 4.8

Co-60 0.78 1.8

1985–1998 Tl-202 0.5 0.9

C-11/N-13f 0.3 0.5

Eu-152 2.2 3.3

Co-58 0.5 0.9

Co-56 0.5 0.9

Hg-197 3.1 4.6

Hg-195 2.5 3.7

Hg-195m 1.8 3.2

Hg-197m 3.8 6

Hg-203 0.8 1.2

Hg-193m 0.7 1.5

Cs-134 0.5 1.1

Os-185 0.6 1.1

V-45 0.5 0.8

Be-7 3.4 8.7

Sc-46 0.5 0.9

Mn-54 0.5 0.9

Cs-137 0.6 1.1

Co-60 0.5 0.8

Br-77 1.7 3.4

Sb-124 0.4 0.8

Ce-141g 2.2 4.4

Ce-144g 12.1 24.2

Cr-51 6.4 12.8

Co-57 1.4 2.8

Cu-67 1.5 3

Fe-59 1.2 2.4

Se-75 1.1 2.2

Se-73 0.4 0.8

Na-22 0.4 0.8

Zn-65 0.8 1.6

1999–presentg 511 keV 1.15 2.3

Be-7 4.2 8.4

Ce-141g 1.35 2.7

Ce-144g 6.5 13

Co-56 0.55 1.1

Co-58 0.5 1

Co-60 0.45 0.9

Cr-51 4.45 8.9

Cs-134 0.5 1

Cs-137 0.6 1.2

Cu-67 1.5 3

Eu-152 1.3 2.6

Hg-203 0.6 1.2

Mn-54 0.4 0.8

Na-22 0.45 0.9

Os-185 0.45 0.9

Ra-226 20.5 41

Sb-124 0.45 0.9

Sc-46 0.7 1.4

Se-75 0.85 1.7

Tl-202 0.5 1

U-235 1.35 2.7
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Period Nuclide LC (nCi) MDA (nCi)b,c

V-48 0.4 0.8

Zn-65 1.1 2.2

Zr-95 0.8 1.6

a. Listing of an MDA for a radionuclide does not necessarily mean that that radionuclide was 

frequently encountered. The MDAs listed in the individual’s results for a given count should 

be used if available.

b. Based on 95% confidence of detection.

c. MDA = LC × 2 unless otherwise specified.

d. The HUMCO I and II systems were designed for screening subjects. Subjects found to have 

contamination levels above background were referred to the 8- by 4-in. NaI detector, which 

had the same sensitivities with an extended count time.

e. By bremsstrahlung.

f. Based on 511 keV. C-11 is a positron emitter with no photons. However, the 511-keV peak 

should always be present due to positron annihilation. The 511-keV peak can have 

interference contributions from other sources including pair production interactions from 

nuclides with photon energies greater than 1,022 keV.

g. Lower sensitivities might be available using the lung counter for certain nuclides if lung 

counting is appropriate to the dose reconstruction.

lung counting are summarized in Table 5-20. The values of MDA and LC are calculated in accordance 

with the theoretical developments of Currie and of Altshuler and Pasternack (Vasilik et al. 1984). No 

information is available on MDAs for the thyroid detector or for 131I or 125I in the whole-body counter. If 

available, the MDAs that are listed in the records of the worker should be used rather than the MDAs 

listed in the tables.

Whole-body and lung counting are the primary methods for determining intakes of fission and 

activation products. However, both in vitro and in vivo results might be available.

Results in the bioassay records are generally reported in nanocuries unless otherwise indicated. 

Results that are listed as NULL indicate no detectable activity (NDA) rather than the MDA. Results 

less than the LC are marked as NDA in the database. The actual DL values and counting errors might 

be available on the report for the individual, but they are not available in the IVML database until 

2003. However, MDAs are listed with in vivo count results in the LABDR for 241Am and 239Pu. Since 

2004 MDAs are also listed for 235U and 234Th. In addition, every count performed since 1969 when the 

program was formalized has been recorded in the database and has a “White Card” on file. Counts 

without positive results for any analyte are marked NDA. However, the results of counts from before 

1969 were only performed in conjunction with significant radiological events. Records of these 

measurements are in logbooks, but they might only indicate that measurements were made, without 

associated results. Early in vivo measurements were made in “go/no-go” fashion using instruments 

with little or no energy discrimination capability. Negative results might not have been recorded but 

only verbally communicated to the subject. Logbooks with early in vivo measurements results are in 

LANL archives on the site (Hoover 2007).

An in vivo count spectrum is not analyzed for a fission or activation product radionuclide unless a 

peak that is associated with that nuclide is visible in the spectrum. When that peak is visible, the 

suspected nuclide is added to the library and the spectrum is reanalyzed. Visual or non-library-driven 

software recognition of a peak can be subjective and not directly correlate to MDA or LC calculations, 

especially with the broad peaks that are associated with scintillation detectors. For whole-body 

counts, it is not reasonable to assume that a worker was exposed to or is being monitored for all 

radionuclides potentially reportable simply because an MDA was determined and listed on the report.

In general, no information is available in the reports on the assignment of respiratory absorption type 

for specific fission and activation product nuclides. Individual guidance might be available in the 

narrative on the White Card that accompanies a worker’s in vivo bioassay report. It is unlikely that the 
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 48 of 117

information on the White Card has been transferred to the remarks section of the LABDR. However, if 

the dose reconstructor deems it appropriate, copies of the White or Beige Cards can be requested.

RaLa operations occurred at LANL from September 1, 1944, through July 18, 1963. The internal 

dose from intakes or potential intakes of 140La and 140Ba from RaLa operations at TA-10 (Bayo 

Canyon Site), TA-35 (Ten Site), and Buildings H, Sigma, and U in TA-1 for chemists and other

maintenance and operations workers and security personnel cannot be assessed due to the lack of a 

well-established in vivo bioassay program during that period. Some whole-body counts might have 

been performed beginning in 1955. In vivo results might be available in the LABDR.

5.3.1.2 Cesium-137 and Other Intakes from Fallout

Many workers in the early days of whole-body counting had detectable activities of 137Cs. Most of this 

was attributed to fallout. Some workers had even higher levels of 137Cs from consumption of wild 

game. A DL used to establish the difference between occupational and nonoccupational sources of 

137Cs and other fallout radionuclide intakes has not been discovered in the records. In lieu of other 

information, the following guidance can be used:

• The 137Cs intake should be considered occupational if the same whole-body count detected 

other fission or activation products. It should also be considered occupational if a fission or 

activation product or radiostrontium urinalysis showed detectable activity and the sample was 

obtained in a reasonable time before or after the whole-body count or within the period 

between the previous and next whole-body counts. The reasonable time is based on the 

biological retention pattern of the radionuclide in the body.

• All other fission or activation products that are identified in the whole-body or lung count 

should be considered occupational unless specifically stated on the White Card and the 

reasons for invalidating the results are acceptable to dose reconstruction practices.

• If an investigation occurred and the record clearly shows that the intake was due to a 

nonoccupational source, the 137Cs can be disregarded. The results of the investigation would 

be noted on the White Card.

• National Council on Radiation Protection and Measurements (NCRP) Report 94 provides 

mean body burdens of 137Cs for the United States for the years most likely to produce 

interference with occupational whole-body count results (NCRP 1988). Those values are 

listed in Table 5-19. If no other fission or activation products are linked to the intake and the 

137Cs result is less than the values in Table 5-19, the 137Cs result can be assumed to be due to 

fallout.
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Table 5-19. Mean body burdens of 137Cs from 

fallout in the United States (nCi).

a

Year Body burden Year Body burden

1953 0.27 1966 9.7

1954 1.1 1967 5.6

1955 2.2 1968 3.5

1956 4.3 1969 2.7

1957 5.1 1970 2.7

1958 6.5 1971 2.7

1959 8.1 1972 2.7

1960 6.8 1973 2.7

1961 4.6 1974 1.6

1962 6 1975 1.1

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

a. NCRP (1988).

5.3.2 Lung Burdens

Lung burdens of 239Pu, 238Pu, and 241Am were monitored using the phoswich lung detectors beginning 

in 1970. The 59.5-keV gamma line of 241Am is used to determine the 241Am burden (50- to 70-keV 

region). If the isotopic ratio for a given intake is known, the 239Pu and 238Pu can be determined from 

the 241Am. Otherwise, the plutonium is determined from the uranium low-energy X-ray (L X-ray) 

region. When 241Am, 239Pu, and 238Pu are present, corrections for the contribution of the neptunium 

L X-rays, from the decay of 241Am, to the 14- to 25-keV 239Pu and 238Pu regions must be considered. 

The phoswich detectors were eventually replaced by arrays of HPGe detectors, which greatly 

improved the energy resolution. Improved energy resolution permitted the system to distinguish 

between gamma and X-ray lines that are closer together. However, because the uranium L X-ray 

energies for the decay of 239Pu and 238Pu are the same, there is no way to differentiate between these 

two isotopes in an actual measurement. Isotopic information about the exposure is used to determine 

the appropriate calibration factor.

Efficiency, and therefore sensitivity level, varies for every count due to the effects of chest wall 

thickness on the attenuation of the 17-keV X-ray and the 59.5-keV gamma ray. Therefore, the MDA 

that is listed with the count should be used when available. The MDA and LC values in Table 5-20 are 

nominal and based on the calibration chest wall thickness of 2.5 cm. These can be used for 

correlation with projected bioassay results. Chest wall thickness for the individual can typically be 

found on the White Card that is associated with the in vivo counting record. Chest wall thickness is 

estimated by weight:height ratios for routine counting and by ultrasound for special or positive counts. 

For lung (chest) counts, increases in chest wall thickness can increase the MDA for the individual 

count. University of California (1977b) suggests that MDAs should be increased 50% for large 

individuals. The dose reconstructor should use best judgment in determining the applicability of the 

listed MDAs for bounding missed dose projections. The MDA and LC values for lung counting are 

summarized in Table 5-20.

MDAs are typically reported for 239Pu and 241Am for lung counts. However, 241Am is not always 

detectable even if a plutonium lung burden exists. Much of the plutonium at LANL, especially in the 

early years, was fresh [2]. Plutonium burdens from the 26 individuals from the original Manhattan 

Project have been calculated by LANL to be 6 to 80 nCi since before 1946. These individuals have

been followed for more than 50 years. Twenty-one of them received their first lung counts in 1971. At 

that time the 239Pu MDA was 7 nCi for a 2,000-second count. Positive counts were obtained for 14 of 

the 21 measured persons. The lung burdens ranged from 3 to 10 μCi 239Pu (Hempelmann, 

Richmond, and Voelz 1973). None of these individuals has had 241Am activity above background 
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Table 5-20. LC and MDA values for lung counting (nCi).

Period Radionuclide LC MDAa

1970–1973 (2,000-s count time) (Hempelmann, 

Richmond, and Voelz 1973)

Pu-239 3.5 7

Extended count time

1977b

Am-241 0.3

Pu-238 10

Pu-239 21

1980–1984c

(Ennis 2003)d

(1970–1979)e

Am-241 0.155 0.31

Pu-238 11 22

Pu-239 24 48

1984 

(Vasilik et al. 1984)f

Am-241 0.16 0.32

Pu-238 14 28

Pu-239 30 60

1998g–presenth

(Ennis 2003)

Am-241 0.1 0.2

Am-243 0.1 0.2

Pu-238 10 20

Pu-239 31 62

Th-234 0.85 1.7

U-235 0.1 0.2

Np-237 0.2 0.4

Np-239 0.1 0.2

1998–present

Fission/activation products (Ennis 2003)

511 keV 0.1 0.2

Be-7 0.35 0.7

Ce-141 0.1 0.2

Ce-144 0.25 0.5

Co-56 0.1 0.2

Co-58 0.05 0.1

Co-60 0.1 0.2

Cr-51 0.35 0.7

Cs-134 0.05 0.1

Cs-137 0.1 0.2

Cu-67 0.1 0.2

Eu-152 0.1 0.2

Hg-203 0.05 0.1

Mn-54 0.1 0.2

Na-22 0.1 0.2

Nd-147 0.1 0.2

Os-185 0.05 0.1

Ra-226 0.9 1.8

Sb-124 0.05 0.1

Sc-46 0.1 0.2

Se-75 0.05 0.1

Tl-202 0.05 0.1

a. Assume chest wall thickness of 2.3 cm.

b. As listed in 1977 quarterly progress report based on a 60-minute count time and a person of average build 

for a UPPU Club member (University of California 1977b).

c. Might be applicable to the startup of the phoswich system in 1970; no other information available. For a 

2,000-s count time.

d. Assume chest wall thickness of 2.5 cm.

e. There is a reasonable correlation between the expected MDA for a 15- to 20-minute count time, typical of 

standard in vivo counting beginning in 1970, and the 60-minute count MDAs listed in 1977.

f. MDA = LC × 2; recounts were performed if the results were greater than the LC.

g. Lung counter has 10- to 300-keV and 80- to 3,000-keV ranges, so a lower sensitivity for certain fission and 

activation products can be obtained.

h. The phoswich detectors were replaced by an array of six planar HPGe detectors around 1998 (Ennis 2003).
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using in vivo techniques (Voelz et al. 1979; Voelz, Grier, and Hempelmann 1985) except one 

individual with a positive lung count for 241Am 37 years later. This individual was suspected of an 

additional intake of plutonium that contained 241Am in [year redacted].

Based on the above discussion, ingrowth of 241Am from early intakes (1944 to 1945) should be 

expected to be negligible. Intakes in later years might result in 241Am ingrowth or might have 

contained 241Am in the original intake.

Follow-up lung counts for on members of the UPPU Club were typically longer (60 minutes rather than 

15 or 20) than those for routine or investigational lung counting. Therefore, the MDAs for these 

counts should be expected to be between 50% and 70% of the standard MDAs listed. (The rule of 

thumb is to quadruple the count time and halve the MDA.)

Results in bioassay records should be assumed to be in nanocuries unless otherwise stated. Results 

listed as NULL indicate NDA (not MDA). Results less than the LC are marked as NDA in the 

database. The actual DL values and counting errors might be available on the report of the individual 

but are not available in the database until 2003.

All individuals who receive lung counts are monitored for 239Pu and 241Am. In 2004, 235U and 234Th (as 

238U) were added to the routine in vivo analysis library.

5.3.3 Wound Monitoring

In August 1959, the H-6 Group acquired a probe to be used to monitor wounds that were 

contaminated with plutonium. This probe was capable of detecting soft plutonium X-rays. The 

sensitivity of this probe was 1 nCi of plutonium when it was unshielded. This was equivalent to 

detection of one-tenth of a permissible body burden of embedded plutonium in tissue to a depth of 

1 cm (LASL 1959). In 1977, a new NaI detector (12 by 2 mm) was being evaluated. This produced 

an MDA of 0.07 nCi based on WGPu (University of California 1977c).

Wound counting was used primarily as a tool for surgeons to locate plutonium in the wound, not as 

results for calculation of internal dose. Wound monitoring continues to be performed. In most cases, 

intake and dose are not assessed directly from the wound count but rather from urine bioassay data 

in accordance with the guidance in ORAUT-OTIB-0022, Guidance on Wound Modeling for Internal 

Dose Reconstruction (ORAUT 2005a). No other information on instrumentation or sensitivities is 

available.

5.4 INTERFERENCES AND UNCERTAINTIES

5.4.1 Discrepancies in Results

In the early years of the bioassay program, analysis techniques and protocols did not use chemical 

blanks and recoveries in the manner that became standard protocol in later years. When the 

databases were being constructed, beginning in the late 1960s, an attempt was made to incorporate 

parameters in relation to current analysis procedures (Glover 2006). Therefore, the results initially 

reported to the worker might not be the same as the results that finally appear in the database that 

was submitted to NIOSH. A project to validate these results is described in ORAUT-OTIB-0063, Los 

Alamos National Laboratory Bioassay Repository Database (ORAUT 2009a). If there are 

discrepancies between results in the current database and the results supplied by the worker, the 

dose reconstructor should refer to the results of the validation project to resolve any discrepancies.
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5.4.2 Contamination of Samples

Until the Health Pass Ward was established in 1945, the potential existed for sample contamination. 

It is likely that a contaminated sample will appear as an obvious outlier in the dataset for a worker. 

The use of a result from a contaminated sample could result in an overestimate of the worker’s dose, 

but the sample result should be considered real if other data do not demonstrate it to be a false 

positive result. Some variability is expected in any set of results. See Section A.10, Attachment A, for 

an example of a valid series of bioassay results for an individual whose autopsy results verified 

plutonium intakes. The Health Pass Ward was eliminated in 1952 because of expense. However, 

protocols to allow the samples to be collected in a contamination-free environment were established.

For in vivo measurements, contamination can occur as external to the body or, in the case of chest 

counting, as external to the lung. If a follow-up in vivo count shows a dramatic decrease in activity or 

NDA on the same day or within a few days, external contamination can be suspected. Radon 

progeny and medical diagnostic or therapeutic procedures that involve radionuclides can cause 

interference to in vivo measurements, especially for NaI detectors. However, unless the count was 

invalidated or noted as being influenced by such interferences, the results should be used as 

recorded.

5.4.3 Uncertainties

5.4.3.1 In Vivo Counting

Uncertainties for bioassay measurements might be included with the results for excreta or in vivo

measurements. For in vivo results, uncertainties are not reported in the database until 2003, but 

uncertainties might be listed on individual reports before 2003. The listed uncertainties are typically 

reported as 1 standard deviation.

Uncertainties that are associated with chest counting are reduced by the use of different calibrations 

for different chest wall thicknesses and the use of ultrasound to measure chest wall thickness. A 

1-sigma uncertainty of approximately 20% for americium and uranium values in chest counting, not 

including correction for interferences from bone and liver, is assumed. Uncertainties would be much 

higher for an individual with activity in the bone or liver. The uncertainty in lung activity estimates that 

are affected by contributions from activity in the liver and skeleton would probably range from 100% or 

more for levels near or below the MDA to 50% or more for activity above the MDA. The uncertainty in 

the estimate of chest thickness using the height:weight correction was at least 50% (University of 

California 1977b). The mathematical correction was made for routine counts. Special counts and 

counts with positive results were typically corrected using ultrasound chest wall thickness 

measurements.

5.4.3.2 In Vitro Measurements

Uncertainties for normal distributions are often listed in the LABDR for samples. The measurement 

uncertainty is typically listed at 1 standard deviation with the results (Lewis 2006a). When results are 

listed as negative, positive, or zero, use of a normal distribution is recommended because a lognormal 

distribution would not be applicable to negative results. However, a study of the variations in 

plutonium urinary data (Moss et al. 1969) determined that the LANL employee results exhibited as a 

lognormal distribution with a geometric standard deviation (GSD) of 1.9 (from the ratio of the 

84th percentile to the 50th percentile).
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5.5 UNMONITORED INTAKES

Exposure circumstances discussed throughout this TBD indicate that for Los Alamos National 

Laboratory employees, there is a potential for unmonitored intakes of secondary radionuclides. 

Partially as a result of this potential, NIOSH has added classes of LANL employees to the SEC that 

include all workers through December 31, 1995. For workers for whom partial dose reconstructions 

are performed, internal doses that may be assigned are limited to those based on bioassay monitoring 

information in the claim records and to potential intakes (implied by case-specific information) that 

may be assigned using coworker intakes. For determining when unmonitored intakes may have 

occurred, the most useful sources of information for this review (beyond internal dose monitoring 

records) are listed below.

5.5.1.1 Energy Employee or Survivor Interview

The interview record offers the energy employee the opportunity to choose from a list of radionuclides 

to which he or she might have been exposed, and offers the opportunity to include others that are not 

listed. In addition, the employee is asked about potential exposures and to provide details of 

exposures in relation to job assignments or incidents. The quality of information about exposures is 

much lower when the interviewee is a survivor rather than the energy employee. Even many energy 

employees did not have a detailed knowledge of what they might have been exposed to, or if they 

had, they might not remember. Some employees check every radionuclide in the list; this is not 

typically a credible set of exposures because not all of the radionuclides in the interview list were 

present at LANL.

Often LANL interviewees have excellent recollections of their activities. In dose reconstructions that 

have been completed to date, this has been shown to be particularly true of researchers and 

scientists, who typically listed the radionuclides to which they were most exposed. What is more, the 

individuals most likely to be exposed to LANL’s less common radionuclides are researchers and 

scientists because routine operations for processing did not generally occur on these materials.

5.5.1.2 Initial Case File

Another excellent resource for identifying work assignments that could result in unmonitored potential 

exposures is the initial case file supplied by the DOL. This file often has records that might list 

detailed job assignments. This information can be used to define a potential unmonitored exposure. 

In some cases, occupational medical records not included in the DOE files are in the DOL file. These 

can include records that contain important detail on potential exposures including extremity dose 

information, nasal count data, airborne contamination reports, and incident reports. Dose 

reconstructors estimating doses for LANL workers should always, therefore, review the DOL files.

Once the potential for an unmonitored intake is identified, a variety of techniques can be applied to 

estimate doses from unmonitored intakes as described in this section.

Assignment of coworker dose is done in accordance with the sections below, which discuss specific 

radionuclides, and with ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2007b). 

Evaluation of the incidents in attachment E should be performed before assignment of dose from 

coworker intakes to see if any correspond with a potential intake.

5.5.2 Unmonitored Intakes of Primary Radionuclides

Monitoring for primary radionuclides has been performed in most years at LANL, although bioassay 

results for americium are sparse before the common use of in vivo measurements from the 1970s. As 

described in Section 5.1.5, some work groups or individuals might not have been monitored or might 
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have been incompletely monitored, even for primary radionuclides. For unmonitored intakes of 

primary radionuclides that are implied in case records, except americium, dose may be assigned for 

using the coworker dose intakes in ORAUT-OTIB-0062, Internal Dosimetry Coworker Data for Los 

Alamos National Laboratory (ORAUT 2009b). Reconstruction of internal dose from intakes of 

americium has been determined to be infeasible for periods before January 1, 1976 (NIOSH 2007); 

americium dose assignment from this date forward is discussed in the next section.

From 1943 through 1946 there was a significant potential for unmonitored intakes of plutonium 

(nominally 3% fresh mixture), 210Po in designated areas, and uranium (nominally either DU or EU 

even though NU was used extensively in conventional weapons testing).

5.5.2.1 Plutonium

In accordance with the designation of the class of employees added to the SEC, reconstruction of 

internal dose from intakes of plutonium has been determined to be infeasible for periods before 1944 

(NIOSH 2007). Bioassay for plutonium was not available before late 1944 and was not completely 

established until 1945. If a worker had a potential for plutonium exposure from 1943 through 1946, 

with no bioassay during that period but plutonium bioassay during later years, dose reconstructors 

may, in some cases, use the later plutonium bioassay with assumptions that are favorable to 

claimants to bound intakes (Schulte and Meyer 1957). However, if the worker did not participate in a 

plutonium bioassay program after this period, coworker dose intakes from ORAUT-OTIB-0062 

(ORAUT 2009b) should be assigned from January 1, 1944, onward for 239Pu mixtures.

5.5.2.2 Polonium

Polonium was used in nuclear weapons initiators, and the primary manufacture of the parts took place 

at the Monsanto Chemical Company. In February 1945, the schedule for polonium delivery from 

Monsanto to TA-1 was increased from a few curies to 100 Ci/mo by June and 500 Ci/mo by December. 

At TA-1, polonium and radium were handled in H and Gamma Buildings. H Building was used for 

preparation of neutron and alpha sources for initiators and isotopic experiments. Workers who were 

involved in these operations were part of Group CM-15. Operations in H Building were carried out 

between 1943 and July 1945 when operations were moved to the DP Site. Polonium-210 was 

processed through various operations that included (1) solution chemistry, (2) electrodeposition,

(3) high-vacuum distillation, (4) metal plating, and (5) counting and assay of polonium.

Intakes of polonium might be type F or M, and unless bioassay data are precisely fit, the favorable to 

claimant absorption type should be assumed. Intakes of polonium do not exhibit a long-term 

excretion in the same manner as plutonium intakes; therefore, dose assessment for early intakes 

using bioassay data from a later time is not possible.

For polonium, existing bioassay results were used to estimate coworker dose intakes for 1944 to 

1956. Before 1944, the reconstruction of dose from intakes of polonium was determined to be 

infeasible (NIOSH 2007).

5.5.2.3 Uranium

Coworker dose intakes for uranium are available beginning in 1947. For exposure before 1947, the 

uranium intakes assigned for atomic weapons employers conducting uranium machining activities 

should be assigned to those workers who might have been involved with uranium machining (Battelle 

2011) as given in Table 5-21. These intake rates are applicable to machining work involving NU. 

Potential internal dose from exposure due to machining activities involving highly enriched uranium 

cannot be reconstructed.
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Table 5-21. Hypothetical chronic intakes for NU, 

1943 to 1946.

Job category Intake (pCi/d)

Machinist 19,654

General laborer 9,827

Supervisor 4,914

Clerical 491

5.5.2.4 Tritium

Coworker dose intakes have been calculated for LANL as described in ORAUT (2009b) and cover 

1950 and onward. Reconstruction of tritium doses has been determined to be infeasible before the 

year 1950 (NIOSH 2007).

5.5.2.5 Americium-241. 

A large number of bioassay results exist for this radionuclide in the post-1975 period. Although 

reconstruction of dose from americium before 1976 has been determined to be infeasible (NIOSH 

2007), dose reconstruction for the period after 1975 is straightforward due to the availability of 

personnel monitoring results, which are primarily from in vivo monitoring 

5.5.3 Unmonitored Intakes of LANL Radionuclides other than Primary Radionuclides

Since the mid-1970s, LANL has maintained consistency in its approach to internal dose monitoring. 

As described by the LANL program (Hoover 2008):

...internal (and external) dose monitoring is prescribed based on the likelihood of 

exposure. Routine internal dosimetry programs ... are established and implemented 

based on worker locations and activities, tailored to the types, quantities, and potential 

for intake of radioactive materials. Special internal dosimetry measurements are made 

under unique circumstances – typically radiological events – where additional or unique 

measurements are warranted, including when unusual results are obtained.

As listed in the reference, LANL maintains the following bioassay capabilities:

• Actinium-227. LANL has the ability to detect 227Ac with target in vivo measurements. 

Targeted dosimetry was performed in the past, but no intakes were detected.

• Neptunium-237. A routine threshold for internal dose monitoring exists in LANL TBDs. A 

contract exists to process in vitro samples, and the ready ability to perform in vivo 

measurements exists.

• Protactinium-231. LANL can detect 231Pa with in vivo measurements if warranted.

• Curium-244. LANL maintains in vitro capability with targeted bioassay and analysis of 

samples at ORNL.

• Thorium-230 and -232. LANL maintains an accredited in vitro bioassay program under the 

DOE Laboratory Accreditation Program using a contract laboratory for sample analysis. In 

vivo bioassay might be conducted on the site, and 232Th has been detected using in vivo

analysis.

• Strontium-90. LANL is able to detect this radionuclide using in vitro analysis. Targeted 

bioassay has been conducted in the past.
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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.

5.5.3.1 Actinium-227

Several buildings in TA-21 were used for work with 227Ac that was used as a substitute for the 210Po 

used in weapons initiators. Weapons that used initiators are likely to have been retired by about 

1963, and it is not known whether actinium was used as part of a weapon that was actually deployed. 

The actinium production program was cancelled in 1955 (DOE 1993), and processing activities at 

LANL would have ended by 1955 or 1956.

Actinium-227 bioassay results exist for LANL workers only for 1954, but have not to date been 

available to the dose reconstructor. A memorandum from the U.S. Atomic Energy Commission (AEC) 

that described LANL releases indicates that 2.5 Ci of 227Ac had been disposed in covered absorption 

beds, with a decay-corrected inventory of 1.4 Ci as of November 11, 1973 (Wingfield 1974). The 

originally disposed amount had decayed to approximately 56% of its original value, which implies a 

residence time of approximately 18 years in the burial ground. This would mean that the material was 

disposed of in 1955 or 1956.

A filter building was constructed in the late 1940s to process exhaust air from Buildings 146 and 152 

in TA-21 (Harper and Garde 1981). The likely primary source of any intake of actinium in the period 

after 1975 was the decontamination and decommissioning of this facility. Health physics techniques 

that were applied to the decommissioning and decontamination operation likely controlled the 

potential intakes of workers, and potential intakes to individuals outside this group would be negligible.

5.5.3.2 Curium-244

The effort to isolate actinides, along with investigation into their chemistry, was conducted partly at 

LANL, and it appears that use of this radionuclide related primarily to actinide research (Penneman 

and Keenan 1960). Research was conducted in the Medical Laboratory in TA-1 and in the CMR-4 

laboratory at DP West in TA-21. A number of papers were published on the radiochemistry of 

americium and curium that span a period from 1958 to 1967 (these were identified in searches of the 

Office of Science and Technical Information electronic archive using keywords americium and 

curium). A 1974 AEC memorandum that identified airborne releases in the period before 1967 lists a 

one-time release of 244Cm from Building 42 in TA-1 (the medical laboratory or ML Building) over the 

period from 1944 to 1956 (Wingfield 1974). Curium bioassay results are available for 1955 in a 

bioassay logbook (LANL 1954–1957) but are not known to be in claimant records.

In 2003, a worker’s DOE badge was found to be contaminated with 244Cm. However, rather than 

suggesting a generalized exposure hazard, the material was traced back to legacy contamination of 

objects that had been disposed as part of a remediation activity for the hot cell area of TA-48-1 (the 

Radiochemistry Laboratory). The material had been stored for tens of years and a sample from the 

box was labeled in the 1970s (DOE 2008). In this case, in vivo and in vitro bioassay using the RAS 

technique were performed; the individual’s dose was estimated to be approximately 0.01 rem (Little, 

Miller, and Guillmette 2003c).

5.5.3.3 Neptunium-237

Limited information was found about neptunium at LANL, but it appears that 237Np activities have 

primarily been associated with the Nuclear Fuels Group. The 237Np metal for the bare criticality 

experiment in 2002 (Roark 2003) was prepared at the CMR Building (Bodenstein et al. 1999). At that 

time, LANL maintained a routine dosimetry threshold for 237Np in internal dosimetry TBDs (Hoover 

2008); however, no workers at that time were on routine bioassay programs. This is likely due to the 
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isolation that is afforded by the hot cells facilities in Wing 9 of the CMR Building. These facilities had 

been available at LANL since their dedication in 1961 (Bodenstein et al. 1999).

5.5.3.4 Protactinium-231

Papers on the chemistry of protactinium were published from 1959 to 1979 (Kirby 1959; Asprey and 

Penneman 1964; Smith, Spirlet, and Muller 1979). The foundation work is documented in 

Radiochemistry of Protactinium, which is based on work performed at the Mound site (Kirby 1959).

Another Mound document (Eppley and Valleé 1979) shows that LANL had requested 231Pa in the 

amounts of 0.1, 30.1, and 11.0 g in fiscal years 1979, 1980, and 1981, respectively, but the pilot plant 

producing the material was shut down in 1979 (DOE 1993). Total production for the plant was less 

than 0.9 g (DOE 1993), so the amounts requested for years 1980 and 1981 could not have been 

delivered. No production-scale operations were identified at LANL, and there is no mention of the 

material in airborne radioactivity records or in the waste inventories in the 1974 AEC memorandum 

(Wingfield 1974). In vitro bioassay methodologies are described for 231Pa in Attachment C of this 

TBD, and LANL has maintained the ability to detect this radionuclide since the mid-1970s 

(Hoover 2008).

5.5.3.5 Thorium

Two distinct source terms are associated with 230Th and 232Th at LANL.

5.5.3.6 Thorium-230

In the 1950s an effort was underway within the AEC complex to build a production capacity for 230Th

(which was also called ionium). A projected use of the material was for radioisotope heat sources. 

This work seems to be associated with a number of sites as part of a general effort sponsored by the 

AEC. No records were found among the available literature that indicated the presence of isolated 

material other than a number of bioassays for 230Th in 1958 found in LANL bioassay records. In 

December 1956, the Mound plant shipped 49 g of 230Th to the University of California Radiation 

Laboratory in Berkeley (DOE 1993). The material does not compose a part of the airborne release or 

waste disposal inventories that accumulated through 1973 in the 1974 AEC memorandum (Wingfield 

1974).

5.5.3.7 Thorium-232

It appears that there might have been two periods when 232Th work was conducted. The first period 

might be associated with a potential airborne exposure from the Sigma complex over the period from 

1944 to 1963, and 232Th was one of the identified radionuclides (Wingfield 1974). Casting, machining, 

and powder metallurgy were conducted at the Sigma complex, which housed the Material Science 

and Technology Division. This facility was built in the 1950s and 1960s (LANL 2007). The material 

was in the form of ingots and oxides in the building designated as a Thorium Storage Building 

(Building 159). Building 66 in TA-3 was a warehouse where thorium was stored, and air emissions 

results from several years indicate airborne releases of 232Th from this location (Wingfield 1974).

A 1958 reference describes safety measures for uranium and thorium (Stout 1958). The procedures 

in the reference apply to both elements due to the similarity of the hazards (radioactivity, toxicity, 

explosion, and fire). The comment is made in this publication that this manual was directing a 

changed policy toward the handling of thorium, and that it was to be considered in the future an active 

alpha emitter (Stout 1958). The implication is that 232Th, with its very long half-life, had been 

considered essentially stable. It is true that the activity is low for this reason, but the ingrowth of its 

alpha-emitting progeny occurs within a short period based on their short half-lives. Because the 
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material had not been assumed to be a hazard before this statement, the material is assumed to 

consist of 232Th rather than the more active 230Th, which with its higher specific activity would have 

required controls with other active alpha emitters.

5.5.3.8 Mixed Fission and Activation Products

MFPs are typically associated with reactors or with facilities where irradiated reactor fuel is processed. 

Mixed activation products (MAPs) are the result of neutron irradiation in a reactor core or an 

accelerator. Before the advent of whole-body counting, radiation protection policy for MFPs was often 

to collect bioassay samples for gross beta and assign the dose to the most limiting radionuclide. For 

this reason, bioassay results for a spectrum of radionuclides from this source term are not expected, 

and gross beta or isotopic samples for 90Sr and 137Cs are typical. The in vivo monitoring program is 

capable of detecting a large range of fission and activation products. Guidance for estimating dose 

from gross beta urine samples is given in ORAUT (2007c). 

5.6 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database (SRDB).

[1] William Moss. University of California (retired). Radiochemist. 2007.

Discussions with William Moss indicate that no correction for recovery was included in the 

calculation of the polonium activity. There is no indication that any further recalculations were 

performed on the bioassay results before loading these results in the current bioassay 

database.

[2] James Lawrence. University of California (retired). Health Physicist. 2006.

Discussions with James Lawrence indicate that most of the early work with plutonium was with 

what he described as fresh plutonium.
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GLOSSARY

activation

Creation of a different isotope by bombarding the parent isotope with neutrons, protons, or 

other types of radiation that results in the formation of a different isotope or element.

activity fraction

Proportion of the total activity due to a particular radionuclide.

age or aging

In relation to reactor fuel and mixtures of plutonium isotopes, time since the step in the 

refinement process that separates americium from the mixture.

BiPO4 era

Period at LANL from October 1949 to January 1957 when plutonium samples were analyzed 

with the BiPO4 technique.

concentration guide

Average concentration of a radionuclide in air or water to which a worker can be continuously 

exposed without exceeding acceptable radiation dose standards.

cupferron era

Period at LANL from March 1944 to October 1949 when plutonium samples were analyzed 

with the cupferron technique.

diethylene triamine pentaacetic acid (DTPA)

Chelating agents in the form of calcium salts (CaDTPA) or zinc salts (ZnDTPA).

dry box

Predecessor to the modern-day glovebox made of wood with ports for attached rubber gloves, 

often filled with inert gas to further contain dust. High-efficiency filters on the exhaust and 

pressure differentials controlled the spread of contamination to the outside. The operator 

could perform tasks while completely shielded from exposure to contaminated dust or other 

airborne material.

implosion

Sudden inward compression and reduction in volume.

ionium

Thorium-230, the decay product of 234U.

kiva

One of the remotely controlled critical assembly buildings associated with the Critical 

Experiment Facility at LANL. From Hopi, ki- means house, but the meaning of -va is unknown.

nanocurie-year (nCi-yr)

Product of the current incremental body burden times the number of years from the estimated 

date of the incremental uptake to the date of the calculation. For a deceased person, it is the 

date of death; for a living person, it is the first date of the month after the latest sample result. 

Incremental nanocurie-years can be summed to provide total nanocurie-years.

oralloy (OR-93)

Uranium enriched to 40% or 93% 235U. The name derives from Oak Ridge alloy.
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rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose.

simulated 24-hour urine sample

Collection of all urine samples beginning with the void before retiring for the evening and 

ending with the first void after rising the next morning, for two consecutive nights, to simulate a 

24-hour urine sample.

tolerance values

Value above which the concentration of a radionuclide in a bioassay sample indicates an 

unacceptable intake or an unacceptable body burden in the individual.

T3O8

Uranium oxide based on Tuballoy (natural uranium).

Tuballoy

Natural uranium.
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TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Scope .................................................................................................................................... 74

A.2 Codes Used in Bioassay and Internal Dose Records ............................................................. 74

A.2.1. Los Alamos Bioassay Data Repository Codes............................................................ 74

A.3 Locations and Typical Radionuclides ..................................................................................... 74

A.4 Computer Code for Validating Plutonium Bioassay Samples ................................................. 76

A.4.1. Validations Techniques PUQFUA2............................................................................. 84

A.4.2. PUQFUA Primary Validation Technique ..................................................................... 86

A.5 Summary of Computer Codes................................................................................................ 87

A.6 Current and Historical In Vitro Bioassay Programs................................................................. 89

A.7 Estimation of Date of Intake ................................................................................................... 89

A.8 Routine Sampling Procedures................................................................................................ 90

A.9 Summary of Analytical Techniques for Plutonium .................................................................. 90

A.10 Example Bioassay Results..................................................................................................... 90

A.11 Routine Sampling Frequency ................................................................................................. 93

A.12 Sampling Protocol for Accidental Exposures.......................................................................... 93

A.13 Summary of In Vitro and In Vivo Sensitivities ......................................................................... 93

A.14 In Vivo Bioassay Routine Frequency...................................................................................... 93

A.15 Human Experimentation......................................................................................................... 93

LIST OF TABLES

TABLE TITLE PAGE

A-1 Identification match ................................................................................................................75

A-2 Assessment method codes for bioassay reports .................................................................... 75

A-3 In vitro collection requirements............................................................................................... 75

A-4 In vitro collection protocols ..................................................................................................... 75

A-5 Sample units ..........................................................................................................................75

A-6 In vivo detector systems......................................................................................................... 76

A-7 In vivo collection protocols ..................................................................................................... 76

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A-8 Current and historical locations .............................................................................................. 77

A-9 Zia employee access to plutonium areas ............................................................................... 88

A-10 Historical in vitro bioassay...................................................................................................... 89

A-11 Current in vitro bioassay performed, 1997.............................................................................. 89

A-12 Routine sampling procedure .................................................................................................. 90

A-13 Plutonium urinalysis sensitivities and analytical techniques.................................................... 91

A-14 Example urine bioassay results.............................................................................................. 91

A-15 Routine plutonium sampling frequency................................................................................... 94

A-16 Routine sampling frequency 241Am exposures........................................................................ 96

A-17 Sampling protocol for tritium................................................................................................... 96

A-18 Uranium routine sampling protocol......................................................................................... 96

A-19 Sampling protocol for accidental plutonium exposures, 1967................................................. 96

A-20 Sampling protocol for accidental plutonium exposures, 1973................................................. 97

A-21 Sampling protocol for accidental plutonium exposures, 1998................................................. 98

A-22 Sampling protocol for accidental 241Am exposures ................................................................. 98

A-23 Uranium nonroutine sampling protocol................................................................................... 99

A-24 In vitro bioassay sensitivity..................................................................................................... 99

A-25 In vivo bioassay routine frequency ....................................................................................... 100

A-26 Amount of plutonium administered to subjects by intravenous injection and the 

date of the administration..................................................................................................... 101

A.1 SCOPE

The purpose of Attachment A is to summarize the data available in Section 5 and to provide 

supporting information to discussions in the text.

A.2 CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS

A.2.1 Los Alamos Bioassay Data Repository Codes

The information for the worker in the current LABDR is provided to the dose reconstructor in PDF 

format. Each of these PDF files contains pages that define the codes that were used in reporting the 

data. These include codes that relate to the verification of the identification match and codes that 

relate to the bioassay results. These codes, as defined at the date of this document revision, are 

listed in Tables A-1 through A-7. However, the dose reconstructor should review the codes in the file 

to ensure they are interpreted in accordance with current status.

A.3 LOCATIONS AND TYPICAL RADIONUCLIDES

The historical and current typical locations for radionuclides are listed in Table A-8 (Inkret et al. 1998c; 

ENSR 2002).

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Table A-1. Identification match.

Match 

type

Match 

level Description

s 1 Social Security Number was matched.

n1b 2 First name and entire birth date (month, day, and year) were matched.

nb 2 Name and birth date were matched.

n 3 Full name match.

n1 4 The first letter of the first name was matched (for example, J. J. Smith for John J. Smith) and 

the last name was matched.

nmd 4 Last name and birth month and birthday were matched.

nmy 4 Name and birth month and birth year were matched.

none 5 No match was found in the LANL personnel tables.

Table A-2. Assessment method codes for bioassay reports.

Code Description

POR Polonium by alpha spectroscopy (May 1, 1955, to present)

LS Liquid scintillation (1981 to present)

UNAA Uranium by DNAA (January 1, 1978, to January 1, 1992)

GB Gross beta

T3R Tritium by radiometric (1950 to 1981)

UF Uranium by fluorophotometric determination (1950 to 1986)

UR Uranium by radiometric (gross alpha) (January 1, 1982, to present)

TIMS Thermal ionization mass spectroscopy (January 1, 1998, to present)

Ref. LA-UR-05-1942 Analysis type not available. Refer to LA-UR-05-1942 (Clark 2005).

Table A-3. In vitro collection requirements.

Code Description

Routine Collected on a regular schedule

Special One-time-only collection for incident or follow-up

Prompt action samples Collected or count performed after incident

Termination sample Collected or count performed due to termination from program or LANL

Table A-4. In vitro collection protocols.

Type Description

Spot sample Single void collection

Simulated 24-hr Morning and evening collection for 2-d period

True 24-hr A 24-hr urine collection

500-mL sample Two single voids collected for analysis

Home drinking water A 500-mL sample of individual’s home drinking water

Spot fecal A single fecal voiding

Autopsy tissue sample Collected at autopsy – tissue type listed as sample type

Unknown Sample collection protocol not available. Refer to LA-UR-05-1942 (Clark 2005).

Blood An intravenous sample of blood; see sample size and units for additional details

Table A-5. Sample units.

Code Description

G grams

ML milliliters

L liters

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Table A-6. In vivo detector systems.

Code Description

GELI-WHOLE BODY Whole-body counting system used in SB-14 from (approximately) 08/31/1978, to 

09/06/1979, when Ge(Li) and HPGe were moved to SB-16 and new phoswich 

system was installed in SB-16.

HUMCO II-WHOLE 

BODY

These systems were in use before 1969 for contamination screening. There are no 

numerical results in the database for counts made with either of these systems.

HPGE-THYROID HPGe detector used for thyroid counting.

ORTEC-WHOLE 

BODY ORTEC WBC

Single ORTEC coaxial HPGe detector over abdomen used as whole-body counter.

PGT-5-DET-CHEST 

SML

(Sum Low; 300 channels) five-detector array, used 01/01/2004 to 07/22/2004.

PGT-CHEST SMH (Sum High; 1024 channels) six-detector array, amplification set to capture 0–3 MeV.

PGT-SKULL SML (Sum Low; 300 channels) six-detector PGT array, on or before 06/20/1996 to after 

01/01/2004

PHOS-BACK Phoswich NaI/CsI.

PHOS-CHEST Phoswich NaI/CsI.

PHOS-HAND Phoswich NaI/CsI.

PHOS-HAND WOUND Phoswich NaI/CsI.

PHOS-LIVER Phoswich NaI/CsI.

PHOS-SKULL Phoswich NaI/CsI.

SB14 9X5 NAITHYROID9 × 5 NaI(Tl) in SB-14 in SB-14 was the whole-body counting system from about 

01/01/1969, to (approximately) 08/31/1978. LC is estimated as 0.5 × MDA.

SB14 9X5 NAI-WHOLE 

BODY

The 9 × 5 NaI(Tl)) in SB-14 was the whole-body counting system from about 

01/011969, to (approximately) 08/31/1978. LC is estimated as 0.5 × MDA.

Table A-7. In vivo collection protocols.

Type Description

BASELINE Initial count before work

NEW HIRE Initial count

RECOUNT Prior count indicates need for recount

REHIRE Individual rehired and reenrolled in in vivo monitoring

REQUEST Special count requested

ROUTINE Count performed on a regular schedule

SPECIAL One-time-only count for incident or follow-up

STANDARD Count performed on a regular schedule

TERMINATION Collected or count performed due to termination from program or LANL

TRANSFER Final count performed due to individual transferring from radiological organization

A.4 COMPUTER CODE FOR VALIDATING PLUTONIUM BIOASSAY SAMPLES

This section is for information only. The dose reconstructor should not attempt to apply these criteria 

to datasets.

Plutonium samples can be marked as invalid in the database because the results did not fit statistical 

expectations. However, the dose reconstructor might find it useful to include these sample results. 

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Table A-8. Current and historical locations.

Location Start

Demolished or 

decommissioned Nuclidea Comment

TA-0 None Original town site

TA-1, Original Main TA 1945 1965 active

1975 

decommissioned

All

TA-1, Building D, plutonium 

chemistry and metallurgy

1943 1954 Pu-239, Pu-238

U-238, DU

Am-241 Absorption depends on matrix 

or pure

Po-210

Ba-140

La-140 Might indicate Sr-90

TA-1, Building D-2, contaminated 

laundry

1953 Pu-239; Pu-240; 

Pu-238; U-236; 

DU;

Po-210;

Ac-227;

Ra-226

Depends on the compound

TA-1, Building D-5, Sigma vault -

storage

1965 Pu-239,

U-238

TA-1, ML Building, medical 

laboratory

Cm

Am

Processing

TA-1, Building C, uranium 

machining

1943 1964 Uranium

TA-1, Building G, uranium and 

graphite sigma pile

1959 Uranium

Ra-226

TA-1, Building H and Gamma 

Building

1945 1957, 1959 Po-210

MFPb InitiatorsCs-137 contamination incident 

occurred

TA-1, Building HT, heat treatment 

and machining

1965 NU and EU

TA-1, Building HT Barrel House, 

storage

1964 Pu-239,

U-238

TA-1, Building M, processing and 

recovery EU

EU Processing, metallurgy, and 

recovery

TA-1, Building M-1, machining U-238

TA-1, Building O 1956 Radium

Radon

Radon cooked off sources on a 

hot plate, Ra/RaBe calibration 

sources

TA-1, Building Q 1959 Radium

Radon

A spill occurred

Ra calibration sources

TA-1, Sigma Building 1965 NU, EU,

thorium

Casting, machining, powder 

metallurgy

TA-1, Building TU, machining 

Tuballoy

1964 NU

TA-1, Building TU-1, recovery of 

EU

1964 EU Furnace for burning rags

TA-1, Building V, Machine Shop 1943 1959 Uranium Unusual assignments

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Location Start

Demolished or 

decommissioned Nuclidea Comment

TA-1, Building W, Van de Graaff 

accelerator

Uranium

Po-210

Tritium

Th-228 “Mesiothorium” Ra-228

TA-1, Building X, Cyclotron Beryllium, uranium, 

lithium, tritium, 

strontium targets.

Zn-65

Targets had induced beta 

activity.

TA-1, Building Y, Physics 

Laboratory

Tritium,

uranium

TA-1, Building Z, Cockcroft-Walton 

accelerator

None

TA-2, Omega West Site Housed critical experiments

TA-2, Water Boiler 1943 1974 U-235,

I-131, I-125,

Rb-88, Cs-137,

Xe-131, Ar-41

tritium

Pu-239

Enriched uranium fuel

TA-2, Clementine 1946 1952 NU, plutonium,

I-131, I-125,

Rb-88, Cs-137,

Xe-131, Ar-41

Ruptured plutonium fuel rod, 

uranium reflectors

TA-2, OWR 1956 1995 U-235

I-131, I-125,

Rb-88, Cs-137,

Xe-131, Ar-41

Cr-51,

Na-24

Tc-99m

Enriched uranium fuel

I-125 production loop schedule 

– at times operated “around 

the clock”

TA-3, South Mesa Site, technical 

facilities

1953 All Plutonium processing

TA-3, Building 29, CMR (SM-29) 1951 Pu-238 Wing 9 handled irradiated 

uranium and plutonium In 

hot cells. Small quantities 

of uranium and plutonium, 

MFPs including iodines, Pu238

TA-3, Building FE-19 Plutonium

TA-3, Building 34, Cryogenics Tritium 3,000 Ci HTO released in 1979

TA-3, Building 16, Van de Graaff 

accelerator

Tritium 800 Ci HTO released in 1977

TA-3, Building 35, press building NU, U-235 Normal metallic and oxides

TA-3, Building 39, Tech Shop DU

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Location Start

Demolished or 

decommissioned Nuclidea Comment

TA-3 SM-40, Physics Allc Incident contaminated large 

portion of building with Po210 through ventilation (late 

1950s or early 1960s)

TA-3, Building 66, Sigma Complex, 

metallurgy and fabrication

<1957 EU, DU powders, 

NU, plutonium, 

thorium

Plutonium processing

Normal metallic and oxide 

uranium (McKown 1958)

TA-3, Building 102, Tech Shops Uranium

Plutonium

TA-3, Building 141, Rolling Mill DU 

Plutonium

TA-3, CMR Building, Wings 3,5,7 H-3 HTO, HT

TA-3, CMR Building, Wing 9 1961 Cs-137 Potential for low-level chronic 

intake in hot cell work

MFPb including

I-131

Pu-238, -239, -240 0.1–10 μm activity median 

aerodynamic diameter 

(AMAD), oxide, nitrate, 

fluoride, and metal. Oxide 

is most common.

TA-3, IBFa, SM-16 I-125 Iodide, labeled organics

H-3 Labeled DNA precursors (OBT), 

water (HTO), HT

P-32 Labeled organics, phosphates

TA-3 Tritium Instrument Calibration 

Facility, SM-40

H-3 HTO, HT

TA-3-184, Occupational Health Pu

TA-3-216 Weapons Test Support Pu

TA-3-700,

SM-700

Acid Neutralization and Pump 

Building

Pu

TA-4, Alpha Site 1956 DU Firing site until 1956, materials 

disposal site C

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Location Start

Demolished or 

decommissioned Nuclidea Comment

TA-6, Two-Mile Mesa Site 1944 1950 DU Detonator manufacturing

TA-8, Nondestructive testing 1984 Pu-239; Pu-238

U-235; DU

Gun firing site

Co-60; Ir-192

Cs-137

TA-10, Building, CMR-10, Bayo 

Canyon Site

1944 1950 Sr-90, RaLa radiochemistry

DU, NU

La-140

Ba-140

TA-11, K Site 1947 Ra-226-Be 20-MeV betatron

TA-15, Electron accelerator 1962 Present Pu-239 PHERMEX

DU

H-3

TA-16, S Site, WETF 1989 Present Pu-239 Explosive casting and 

DU machining

H-3

TA-16, WETF, Bldg 205 Present H-3 Labeled DNA precursors (OBT), 

water (HTO), HT

TA-18 U-235; U-233

Pu-239; Pu-240;

MFPb

;

I-131;

polonium

0.1–10 μm AMAD, oxide, 

nitrate, fluoride, and metal. 

Oxide is most common.

Ruptured polonium source, 

1953

TA-18, Pajarito Laboratory, Rover 

reactor, criticality experiments

1946 1973 MFPsb Betatron used from 1951 to 

1954,

EU metal sphere 1952;

plutonium core added 1 year 

later.

1954 unreflected, delta phase 

plutonium.

TA-19, East Gate Laboratory 1962 None

TA-21, DP-West, plutonium facility 11/1945 WGPu

TA-21, CMR, heat sources Pu-238 Accident with glovebox breach, 

1971

TA-21 DP West Pu-238, -239, -240 0.1–10 μm AMAD, oxide, 

nitrate, fluoride, and metal. 

Oxide is most common.

TA-21, Buildings 2 and 3, wet 

chemistry

1982 Pu 1958 accident, separated 

phases in plutonium process 

tank, unshielded tank

TA-21, Buildings 4 and 5, dry 

chemistry

1981 Pu

TA-21, Building 12, filter building 1975 Pu Actinium-contaminated

TA-21, Building 3, oxalate 

precipitation operations

Pu-239

Pu-238

U-235

TA-21, Building 4 1945 1948 EU hydride

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Location Start

Demolished or 

decommissioned Nuclidea Comment

TA-21, Building 4 1960 Pu-239 Hot cell examine irradiated 

plutonium and EU fuel elements

TA-21 1965 Pu-238 and

Pu-239

TA-21, Building 5, plutonium 

fabrication

Limited use in 1975 Pu-239, -238 Fire contaminated exhaust filter, 

1959

TA-21, Building 150, plutonium 

fuels development, heat sources 

development

1963 Pu-238, -239 Sealed capillary broke, 2,800 ×

MPC

10/1970

TA-21, Building 210, plutonium 

research

Plutonium

TA-21, Building DP-East EU

TA-21, Building 155, TSTA, 

deuterium and tritium fuels.

1984 1990 HT and HTO >10 billion Ci. Equipment failure 

– H-3, 13.8 Ci released

TA-21, Buildings 151, 152, 

experimental program

1945 Po-210 Produced initiators

Ac-227 Produced initiators

TA-21, Building 155 1949 1984 Po-210 Produced initiators

Ac-227 Produced initiators

TA-21, Building 153 1945 In service until 

1970–1973

Po-210 Produced initiators

TA-21 DP-East, TSTA Tritium Test 

Assembly Facility, Building 155 

and the Salt Laboratory, Building 

209 

H-3 Labeled DNA precursors (OBT), 

water (HTO), HT

TA-21, Liquid Waste 

Reprocessing, Buildings 35 and 

257

Late 1940s 1986 Allc Plutonium and transuranic liquid 

wastes

TA-22, TD Site Ac-227 Produced initiators

TA-23 NU Site 1945 1950 Unknown Firing site

TA-24 T Site 1944 DU Facilities transfer to TA-16

TA-25, V Site 1944 1946 DU Taken over by TA-16

TA-26, D Site 1946 1948 U-235, U-238

H-3

Storage vault

TA-27, Gamma Site (Far Point) 1945 1947 Pu-239

DU

thorium

Plutonium gun assembly

TA-28, Magazine A 1979 Present DU Firing site

TA-29, Magazine B 1957 DU Explosives storage area

TA-30 through TA-32 Unknown

TA-33, HP Site, High Pressure 

Tritium Laboratory, Building 86

1950s 1980s late H-3 Tritium oxide (HTO), tritium gas 

(HT)

TA-35, LAMPRE 1955 1967 MFPsb Molten plutonium fuel 

Sr-90 

Co-60

VFPsd

MAP

TA-35, LAPRE I, LAPRE II test 

reactors

1955 1960 MFPsb Highly enriched U fuel

Sr-90

Co-60

VFPs

MAPse

TA-35, Target Fabrication Facility 

(TFF), TSL-213

H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

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Location Start

Demolished or 

decommissioned Nuclidea Comment

TA-35, Ten Site, CMR-10 1950 1963 La-140, Sr-90 contamination suspected 

Ba-140 (F,S)

TA-35 1981 U-235 General site

DU

Np-237

Plutonium

Polonium

TA-35, Laser Fusion Research 1974 Unknown

TA-36, Kappa Site 1950 Present DU

TA-37, Magazine Area C DU

TA-39, Ancho Canyon Site 1955 1960 NU, DU

thorium

Firing points

TA-40, Detonator Firing Site H-3

TA-41, W Site, Weapons Group 

WX

H-3

plutonium

uranium

americium

Engineering of nuclear 

components

Fabrication of test materials

TA-41, Ice House, Building 4 H-3 Tritium oxide (HTO), tritium gas 

(HT)

TA-42, Incinerator Site 1970 Allc Reduced low-level plutoniumcontaminated waste

TA-43 HRLa 1953 1970 I-125 Iodide, labeled organics

H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

C-14 Labeled DNA precursors,

P-32 Labeled organics, phosphates

TA-43 1953 1970 Allc

TA-45, Radioactive Liquid Waste 

Treatment Plant, WD Site

1951 1964–operations 

ended

Allc Removed plutonium before 

discharging effluents 

1960 1963 MFPsb

1967–decom.

TA-46, WA Site 1950 1974 Rover batteries

TA-46, WA Site 1976 1980s U-235, -238 Uranium isotope separation

Thorium

TA-48, Radiochemistry Site 1950s Present All Actinide chemistry and hot cell 

MAPs isotope production e

MFPsb

TA-48, Nuclear Chemistry 1950s Present Se-75 Spallation product, seen in hot 

chemistry on targets

H-3 Tritium oxide (HTO), tritium gas 

(HT)

Cd-109 CL-/NO3

- mixture loaded in 

SnPO4 resin. Cd phosphate is 

most probable material of 

intake, very soluble, 1 μm 

AMAD

I-131 Fission product chemistry

TA-49, Frijoles Mesa Site 1960 1961 H-3

plutonium

uranium

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Location Start

Demolished or 

decommissioned Nuclidea Comment

TA-50, Waste Management Site 1963 Pu-238, -239, -240

Allc

0.1–10 μm AMAD, oxide, 

nitrate, fluoride, and metal. 

Oxide is most common.

TA-51, Environmental Research 

Facility

1962 Present Co-60

Sr-90

Animal exposure facility.

Presently environmental impact 

research.

TA-52, Reactor Development Site Late 

1960s

1970 U-238,

Pu-238

H-3

VFPsd

krypton, xenon

High-temperature, gas-cooled, 

graphite reactor, UHTREX 

(Ultra-High Temperature 

Reactor Experiment)

TA-53, Los Alamos Neutron 

Science Center, largest 

accelerator facility, Los Alamos 

Meson Physics Facility (LAMPF)

1972 C-11, N-13, O-15, 

Ar-41

I-131

Short-lived air activation

Medical isotope production

Induced activity in 

uranium targets, 

corrosion products

TA-54, Waste Storage Facility Pu-238, -239 -240

Allc

0.1–10 μm AMAD, oxide, 

nitrate, fluoride, and metal. 

Oxide is most common.

TA-55, Plutonium Facility (PF-4) 1969 H-3 Labeled DNA precursors (OBT), 

water (HTO), HT

TA-55, Plutonium Facility (PF-4) 1969 Pu-238, -239, -240 0.1–10 μm AMAD, oxide, 

nitrate, fluoride, and metal. 

Oxide is most common.

LANSCE Be-7 2A metal, metalloid behavior, 

very reactive, occurs in virtually 

massless quantities, typically 

seen when target cells are 

opened for maintenance, 

usually in oxide form 

C-11 Byproduct at LANSCE, seen in 

workers during beam cycle

N-13 511 keV during beam cycle

a. Intakes of labeled compounds do not follow the default ICRP Publication 68 models (ICRP 1995). However, doses 

based on bioassay results should be estimated using standard Project assumptions for the analyte. 

b. Cs-137.

c. All = Pu-239, Pu-240, Pu-238 (type S-M); U-235, DU (type M); H-3 (SR-2,F); Po-210 (type F or M); Ac-227 (type S); Ra226 (type S).

d. VFPs = volatile fission products.

e. C-11, N-13, O-15, Ar-41, Be-7, Na-22, Na-24, Co-58, Co-57, Mn-54, Mn-52, V-48.

Review of studies listing the bioassay results of the original 26 workers with intakes of plutonium 

shows considerable variability between samples in both the long and short term (Voelz et al. 1979). 

Section A.10 contains a complete listing of bioassay results for one individual. These results display 

typical variability seen in other individuals involved in the study. If the dose reconstructor chooses to 

maintain the result as invalid, the logic below will provide an explanation of how the data were 

evaluated. The information in the following sections is from Lawrence (1978).

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A.4.1 Validations Techniques PUQFUA2

The PUQFUA programs were revised through PUQFUA4. Therefore, other validation protocols might 

have been in effect after 1978.

1. The purpose of the validation is to determine high results to invalidate the sample.

2. Only samples of a single era could be used to validate samples from that era (e.g., samples 

analyzed by BiPO4 or cupferron could be used to validate samples of the era).

3. Samples are considered positive if they exceed the value of LEAST. If the sample passes 

Test A, do not perform Test B. The tests are applied sequentially to the largest results first, 

then the second largest, etc., until all results exceeding LEAST are examined. If results are 

below LEAST, use Test C.LANL recognizes that unique situations occur when:

a. An individual urine result should be retained despite its being invalidated by the test 

procedures, and

b. An individual urine result should be invalidated despite its being validated by the test 

procedures.

4. Samples surviving Tests A, B, and C would be subjected to the PUQFUA1 primary validation 

technique.

Cupferron era Other era

LEAST

0.10 pCi

High

0.8 pCi

LEAST

0.075 pCi

High

0.4 pCi

Test A

1. If there are four or more positive samples (greater than LEAST) ±1 year – continue.

Select the four closest positive samples, ±1 year.

a. Find the average of the four samples.

b. Find the average of the standard deviation.

2. If the sample result is less than the average plus one-fourth the standard deviation average, 

retain the sample.

Test B – samples not passing the Test A criteria

1. If there are not three or more positive samples after the sample being tested, within the same 

sample era, the sample is validated by default.

2. Determine if one or more retained potential accident dates occurred between the sample being 

tested and the next earlier positive sample.

3. Perform the calculation of the expected excretion levels on the next three positive samples.

4. Perform the calculation using the retained potential accident date or 15 days before the 

sample date if no potential accident date is available.

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RF = [I ÷ (E + I)]0.74 (A-1)

where:

RF = reduction factor

I = number of days between assumed exposure and sample

E = number of days between the sample being tested and the date of the later sample

CL = RF × UR (A-2)

and

PS = RF × U2S (A-3)

where: 

CL = calculated urine level

UR = result

PS = pseudo-standard deviation

U2S = standard deviation of the result

1. If the actual measured urine excretion level of two of the three later samples is greater than 

the calculated urine excretion rate minus one-third of the pseudo-standard deviation on the 

appropriate dates, retain the sample.

2. If two or more retained potential accident dates result in conflicting decisions, retain the 

sample.

Test C

Test C tests all samples less than LEAST in any era. Test C eliminates those low or negative urine 

results that, if left valid, would cause subsequent invalidation of high results that preceded them.

All low and negative samples are retained until the first sample exceeding LEAST is encountered.

Cupferron Era

1. If there is one sample >0.1 pCi in the set and the standard deviation is <0, the sample is 

rejected as too small.

2. If there are three or more samples >0.1 pCi in the set and the standard deviation is <0.1 pCi, 

the sample is rejected as too small.

BiPO4 Era

1. If there is one sample >0.075 pCi in the set and the standard deviation is <0, the sample is 

rejected as too small.

2. If there are three or more samples >0.075 pCi in the set and the standard deviation is <0.075 

pCi, the sample is rejected as too small.

NTA-ZnS or PHA Era

1. If there are at least two samples >0.075 pCi in the set and the standard deviation is ≤0, the 

sample is rejected as too small.

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2. If there are at least five samples >0.075 pCi in the set and the standard deviation is ≤0.04, the 

sample is rejected as too small.

Surviving Samples

Samples surviving the above tests are examined in relation to the earliest date of a validated HIGH 

sample. The date of the earliest sample that exceeds the high criteria is used to test the surviving 

samples. All surviving samples dated before the first high samples are kept with the added notation of 

no highs.

Samples Occurring after the First High Sample

1. The average of the four retained samples greater than LEAST and closest in time to the 

sample being tested is calculated. There is no time limit within the analysis era on dates of 

samples greater than LEAST.

2. If the standard deviation exceeds this average, the sample being tested is kept; otherwise, it is 

rejected.

3. The sample can be rejected if the result is <0 or the standard deviation is <LEAST.

4. If there are not four retained samples greater than LEAST, the sample being tested is kept by 

default.

A.4.2 PUQFUA Primary Validation Technique

1. Starting with the latest sample and working toward earlier samples, successive pairs are 

examined. The later sample of each pair is used to test the validity of the earlier sample.

2. If the earlier sample is validated, it is used as the later of the next pair of samples to be tested.

3. If the earlier sample is invalidated (i.e., set equal to zero for the calculations), the later sample 

of that pair remains the later of the next pair to be tested. The sample next earlier than the 

invalidated sample becomes the one to be tested.

Uc = Ue[(De – E1)/(Dl – E1)]0.74 (A-4)

and

σc = σe[(De – E1)/(Dl – E1)]0.74 (A-5)

and

(Uc – Ul) > 1.282Z1-α [(σc

2

/nc) + (σl

2

/nl)]1/2 (A-6)

where:

De = integer date of Ue

Dl = integer date of Ul

E1 = estimated integer date of intake

Uc = calculated urine result expected from Ue

Ue = earlier dated urine result

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Ul = later dated urine result

σc = calculated standard deviation of Uc

σe = standard deviation of Ue

σl = standard deviation of Ul

1.282Z1-α = standard normal variable for (1 – α) one-sided confidence interval; α = 0.1

Hypothesis

4. Does the calculated urine result Uc exceed the measured result Ul?

5. The Ul is permitted to exceed Uc because the larger value of Ul might have resulted from an 

additional intake between De and Dl.

The basic assumption is that, in the event of no additional intake, the urine level at any later 

time can be calculated from Langham’s urinary elimination equation, provided the date of 

intake has been established. (PUQFUA4 did not use Langham’s urinary equation, but rather 

the equations developed in the late 1980s or early 1990s, which take into account very-longterm elimination 20 or more years after uptake. Again, these are discussed fully in the 

description of PUQFUA4.)

6. Where nc = nl = 1, Uc does exceed Ul at the chosen level of significance, and sample Ue on 

date De is rejected as invalid.

A.5 SUMMARY OF COMPUTER CODES

The PUQSRT code provided a summary report from the PUQFUA database of:

1. Individuals with a potential accident recorded but no urine sample and individuals with a 

potential accident recorded after their latest urine sample

2. A list of individuals whose last urine sample indicated a body burden of >2 nCi

3. A list of individuals whose next-to-the-last urine sample indicated a body burden increase of 

>2 nCi

4. A list of individuals whose total body burden is >10 nCi

Tracking is available for 238Pu, 239Pu, and 242Pu body burdens. These reports were performed 

routinely.

The PUQFUA code provided calculation of 238Pu, 239Pu, and 242Pu body burdens and validation tests 

of the analytical results for the samples. Various versions were used over time as the code was 

upgraded to accommodate additional models and other features. The original version became 

operational in 1959. This program utilizes a set of power function elimination equations for the 

excretion of plutonium. Lawrence (1978) discusses the history and development of the program 

including its use to track accidents and potential accidents and validate urine samples. PUQFUA1 

and PUQFUA2 are revisions to the original code that address recognized deficiencies. When 

possible, results of the calculations have been compared with autopsy data. PUQFUA1 tended to 

overestimate by a factor of 2 to 8; this was corrected in PUQFUA2. The latest version was 

PUQFUA4.

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Internal monitoring of individuals with a potential for exposure to plutonium was controlled using a 

computer code.

The Z1YRPUU code provided a method to mark Zia employees who were permitted access to 

plutonium areas based on the submission of their annual plutonium bioassay samples. This program 

was in use from 1976. Table A-9 lists the areas and exempt areas for access.

Table A-9. Zia employee access to plutonium areas (University of California 1978a).

Area

Urine sample within 

425 d of entry Exempt

Job requiring respiratory protection X

Modifications or repairs on dry boxes or other highly 

contaminated equipment

X

Replacement of plutonium-contaminated filters at all sites X

Janitorial (long-term) work in plutonium operation areas X

Long-term operations (weeks) in areas of low levels of plutonium 

contamination (>1,000 dpm-60 cm2 and <10,000 dpm-60 cm2)

X

Decontamination of plutonium spills with >10,000 dpm-60 cm2 X

Work in burial pits at TA-54 when personnel contamination 

potential is moderate to high

X

Short-term jobs (2–3 d) when sizable quantities of plutonium 

(grams of Pu-238 or kilograms of Pu-239) are present in dry 

boxes (even when work is being done outside dry box)

X

Supervisory personnel X

(base urine 

sample on record)

Short-term jobs (2-3 d) in areas of CMR Building, Ten Site, TA50, TA-55, TA-54, TA-18, TA-48, or TA-21, where there is little 

plutonium contamination (<1,000 dpm-60 cm2)

X

Jobs in other minimum exposure potential areas when respiratory 

protection is not required and possibility of plutonium 

contamination is minimal

X

PUANUD (Plutonium Urine, Accident, Lung, and Wound Data) was a code to permit entry of urine 

assay, potential accident, lung count, and wound counting data into a single file that was used as 

input for the PUQFUA calculations. The software was used beginning in 1974 for 238Pu and 239Pu. 

Flags were available for the following categories.

1. Unspecified type of accident;

2. Wound case with excision;

3. Wound count equal to or greater than 0.2 nCi;

4. High room air count if next year’s urine shows obvious increase;

5. High nasal count if next year’s urine shows obvious increase;

6. Nose count over 1,000 dpm; and

7. Other accident if next year’s urines show obvious increases.

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A.6 CURRENT AND HISTORICAL IN VITRO BIOASSAY PROGRAMS

Historically, the in vitro bioassay program included the nuclides listed in Table A-10. The current in 

vitro bioassay program includes the nuclides and techniques listed in Table A-11. The in vitro 

bioassay samples have been taken at LANL throughout the history of the bioassay program. 

Sensitivities vary with technique and period.

Table A-10. Historical in vitro bioassay.

Nuclide Nuclide

Tritium Iodine

Radium and daughters Fission products

Uranium Activation products

Plutonium Other alpha emitters

Polonium Other beta/gamma

Curium Americium

Strontium

Table A-11. Current in vitro bioassay performed, 1997 (Inkret et al. 1998a).

Material Analytical technique Number monitored

Tritium Liquid scintillation 139

Uranium Alpha spectroscopy 66

Plutonium Alpha spectroscopy 1,467

Plutonium TIMS 408

Americium Alpha spectroscopy 75

Strontium Liquid scintillation (gross beta counting–LSC sent to contract laboratory) 4

A.7 ESTIMATION OF DATE OF INTAKE

The following guidance from Lawrence (1978) might be relevant to an estimate of the date of 

plutonium intake, based on routine and special sampling protocol, when no intake date is recorded or 

to understand the LANL rationale for selection of an intake date listed with the bioassay results if no 

incident date is provided in the records. Otherwise, the dose reconstructor should use current models 

and best judgment in the determination of the dates for acute intakes.

1. Halfway between dates of consecutive pairs of samples if no potential accidents are recorded 

between samples.

2. One-half day before later sample of a pair if a potential accident is recorded on the date of the 

later sample.

3. The earliest potential accident date if several occur between the paired samples.

4. One-half day before the initial sample if the potential accident is recorded on the date of the 

initial sample.

5. The earliest potential accident date before the initial sample if any potential accidents occurred 

before the initial sample.

6. Fifteen days before the initial sample if no potential accidents occurred before the initial 

sample.

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7. Estimated date of intake of the later sample if a pair is reassigned to be the estimated date of 

intake of the earlier sample, if the earlier sample is invalidated and a potential accident date 

occurred on the same date as the earlier sample.

A.8 ROUTINE SAMPLING PROCEDURES

Contamination of bioassay samples was a concern throughout the history of the program. Various 

sampling protocols were followed to minimize the potential for contamination, to detect levels above 

the tolerance level, and to attempt to obtain an equivalent 24-hour collection. Table A-12 lists the 

sampling protocols.

Table A-12. Routine sampling procedure.

Years Nuclide Protocol

1944 Plutonium Collected on 24-hr urine sample in clean areas after decontamination. Working 

in plutonium areas operations – weekly, daily nasal swabs. Blood counts every 6 

wk (Hempelmann 1944).

1945–1952 Plutonium Collected 24-hr urine sample away from working environment to ensure 

contamination-free urine (Clark 2005). 

1953–1957 Plutonium Metal kit for collection of last voiding of day and first in morning for 2 consecutive 

days (Clark 2005).

1958–1967 Plutonium Four 1-pint bottle disposable kit. More closely approached volume of true 24-hr 

urine sample. Reduced potential contamination. 

1944 Polonium Collected on 24-hr urine sample in clean areas after decontamination. Working 

in polonium areas operations – weekly, daily nasal swabs. Blood counts every 6 

wk (Hempelmann 1944).

1945–1952 Polonium Collected 24-hr urine sample away from working environment to ensure 

contamination-free urine (Clark 2005).

1970–1975 Tritium >150 μCi/L – 1/d; 150-75 μCi/L – every 2 wk; 74 -24 μCi/L – 1/wk (Healy 1970).

1975–present Tritium Routine – every 2 wk; or 2 hr after expected exposure then if <1 - no more 

samples, 1-10 μCi/L – 1 wk in next month, 10-100 μCi/L – weekly samples, >100 

μCi/L – daily samples. 

1944–1967 Uranium Weekly urinalysis (Hempelmann 1944).

1967–present Uranium Annual, including drinking water sample since 1992 (Inkret et al. 1998c).

A.9 SUMMARY OF ANALYTICAL TECHNIQUES FOR PLUTONIUM

Various analysis techniques have been employed for plutonium over the history of the program. The 

techniques are summarized in Table A-13.

A.10 EXAMPLE BIOASSAY RESULTS

Table A-14 contains actual bioassay results from an individual whose 1959 autopsy results confirmed 

plutonium in the liver, lungs, bone, and lymph nodes. Plutonium-238/239 ratios indicative of early and 

late mixtures of plutonium were identified in the autopsy tissues. The late mixtures of plutonium were 

found in the lung. During his work at LANL, the individual had only 18 nasal swipes above 50 dpm 

(the level of significance). All the nasal swipes greater than 50 dpm were before 1948. Only one 

incident in 1955 showed a nasal swipe of approximately 28 dpm. Records of the work locations and 

results were carefully maintained (Foreman, Moss, and Langham 1960). No significant incidents or 

accidents were noted in the individual’s records. The probable intake scenario for this individual is 

listed as long-term chronic or intermittent low-level inhalation. Review of the bioassay results should 

assist the dose reconstructor when reviewing the variability of other data sets.

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Table A-13. Plutonium urinalysis sensitivities and analytical techniques (Clark 2005).

Period Urinalysis method

1944–1952 Plutonium extracted by cupferron chloroform, gas flow counter (1944), Simpson alpha counter 

efficiency = 50%, background = 1 cpm to 0.1 cpm (1945+). Could not differentiate polonium 

and plutonium. Selected based on potential exposure. 12% of blanks showed ~1.4 cpm or 

0.8 pCi (1945). Assay results reflect no chemical blanks, recovery factor, or counting 

geometry corrections.

1949–1957 Bi-phosphate-La fluoride coprecipitation, Simpson alpha counter, efficiency = 50%, 

background = 0.1 cpm.

1957–1963 Bi-phosphate precipitation – alkaline earth-phosphate precipitation, plutonium plated on 

stainless-steel disk. NTA counting method, 1,000-min exposure to emulsion, background = 

0.007 dpm.

1963–1965 Ion exchange/plutonium separated on an ion-exchange resin. Electroplated for NTA 

counting, 84% ±14% recovery.

1966 ZnS counter.

1967–1971 Either ZnS or alpha pulse height analysis (PHA) RAS permit measurement of Pu-238; 

background = 0.003 ±0.003 cpm.

1971–present All alpha PHA, computerized spectrometry.

09/18/1980 Began Pu-242 analysis.

1982–1986 Coprecipitation (alkaline earth PO4 or oxalate). Alpha spectrometry (60,000-s count time) or 

rapid Alpha Phosphor Scintillation. Counting (3,600 s).

1997 Alpha spectroscopy, based on class Y, 1 μm, Pu-239 (Inkret et al. 1999).

1997–present TIMS with ultra-trace chemistry and class-100 clean room and alpha spectroscopy methods. 

Based on class Y, 1 μm, Pu-239, use of alpha spectroscopy allows direct measure of 

chemical efficiency and detection of Pu-238 (Inkret et al. 1999).

Table A-14. Example urine bioassay results (Foreman, Moss, and Langham 1960). 

Date (left/right nasal count)

Average room air

concentration (dpm/m3) Activity (dpm/24 hr)

08/09/1946 (189/320 dpm 07/29/1946) 6–188 1.2

09/19/1946 (149/19 09/05/1946) 4.7

12/18/1946 (57/68 12/10/1946) 1.7

04/18/1947 (164/106 dpm 12/30/1946); (102/61 

dpm 01/21/1947); (91/135 dpm 04/01/1947)

11–98 0.7

05/23/1947 0.7

06/26/1947 0.7

07/30/1947 (144/40 dpm 07/07/1947) 0.0

08/27/1947 1.0

10/02/1947 1.5

11/07/1947 (120/78 dpm 10/03/1947) 0.8

12/08/1947 4.0

01/13/1948 24–69 1.0

02/13/1948 (0/59 dpm 02/10/1948) 0.0

03/19/1948 0.0

04/22/1948 (86/3 dpm 04/26/1948) 2

06/23/1948 (244/72 dpm 06/10/1948) 3.7

07/22/1948 (72/1 dpm 07/02/1948) 0.0

08/19/1948 (65/0 dpm 08/02/1948) 0.8

09/20/1948 3.0

10/26/1948 2.0

11/22/1948 2.5

12/21/1948 (50/38 dpm 12/01/1948) 0.0

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Date (left/right nasal count)

Average room air

concentration (dpm/m3) Activity (dpm/24 hr)

01/24/1949 19–72 (Moved to uranium work)

01/31/1949 2.0

07/14/1949 1.2

02/08/1950 0.8

09/01/1950 0.1

02/28/1951 0.8

09/04/1951 0.7

05/19/1952 0.3

12/14/1952 0.0

09/04/1953 0.0

06/04/1954 Off uranium

06/18/1954 0.0

06/08/1955 (Moved back to plutonium)

08/01/1955 3 1.2

08/12/1955 (22/28 dpm 08/09/1955) 0.6

08/19/1955 0.0

09/30/1955 0.7

11/14/1955 0.7

12/27/1955 0.6

02/09/1956 3 0.0

04/05/1956 0.7

04/30/1956 0.0

06/08/1956 0.4

07/20/1956 0.6

08/23/1956 0.5

09/25/1956 0.0

10/24/1956 0.0

11/23/1956 0.0

12/17/1956 0.1

01/31/1957 4 0.23

02/28/1957 0.68

04/12/1957 0.22

05/14/1957 0.12

06/14/1957 0.11

07/15/1957 0.03

08/19/1957 0.00

09/20/1957 0.20

10/22/1957 0.21

10/31/1957 0.39

11/14/1957 0.51

01/10/1958 4 0.00

02/21/1958 0.65

03/25/1958 0.51

05/07/1958 0.025

06/19/1958 0.55

07/30/1958 0.49

09/15/1958 0.79

11/28/1958 0.47

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A body burden of plutonium was potentially maintained by this individual since 1946. The bioassay 

results varied as listed.

Average room air activity concentrations for each year are listed. When more than one room was 

involved, the range of average concentrations is listed. The period from 1946 to 1949 involved 

plutonium nitrate, plutonium oxalate, and plutonium fluorination. The period from 1955 to 1958 

involved primarily liquid-liquid extraction of plutonium under greatly improved exposure conditions.

A.11 ROUTINE SAMPLING FREQUENCY

Routine sampling frequencies place upper bounds on the potential exposure for monitored individuals 

with all results less than detection levels. Table A-15 lists routine sampling frequencies for plutonium. 

Although sampling of individuals with the highest potential for intakes was performed from the 

beginning of the program in 1944, no specific information on the nonincident sampling program is 

available before Lawrence (1967). Table A-16 lists routine sampling frequencies for 241Am exposures; 

routine samples were not performed for 241Am before 1998. Routine and postexposure protocols are 

combined in Table A-17 for tritium. Table A-18 lists frequencies for uranium. Routine work or 

frequent entry in an area with beta/gamma emitting radionuclides currently requires annual wholebody count (LANL 2004).

A.12 SAMPLING PROTOCOL FOR ACCIDENTAL EXPOSURES

Awareness of the sampling protocol associated with exposures provides guidance for the association 

of results with intake regimes. Results corresponding to an incident can be expected to follow the 

appropriate protocols listed below. No information on the protocols is available before Lawrence 

(1967). It should be assumed that the protocol continued until the year of the next dated reference 

unless otherwise stated. Tables A-19 to A-23 list protocols for accidental exposures.

A.13 SUMMARY OF IN VITRO AND IN VIVO SENSITIVITIES

Table A-24 lists sensitivity data for in vitro bioassay.

Tables A-25 and A-26 list sensitivity data for in vivo testing.

Monitoring was available for the liver and thyroid using the HPGe detector. In 1977 the MDA for the 

liver scan was reported to be less than 1 nCi for 235U and 241Am (University of California 1977b).

A.14 IN VIVO BIOASSAY ROUTINE FREQUENCY

Table A-25 provides the standard protocol for scheduling in vivo counts. This is the current protocol. 

No information is available on when the protocol was first established; however, it has been in effect 

as long as anyone currently involved in the program can remember.

A.15 HUMAN EXPERIMENTATION

The first research into the behavior and deposition of plutonium in the human body began by accident 

in August 1944 when a vial containing 10 mg of plutonium exploded in the face of a [title redacted], 

who swallowed a portion of the material. The chemist then used his own urine to develop a bioassay 

analysis to detect the early analysis technique for plutonium. However, this information did not 

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Table A-15. Routine plutonium sampling frequency.

Frequencya 1944 (Kolodney 1946)

1967 

(Lawrence 1967)

1973 

(Lawrence 1973)

1998 

(Inkret et al. 1998 b,c; LANL 2004a)

Monthly Great exposure

Quarterly Moderate exposure 1. Persons working with ≥10 g Pu239 in chemical or metallurgical 

operations, inside or outside 

glovebox

2. Persons with 50% body burden 

(MPBB = 0.04 μCi Pu-239)

1. Persons working with ≥10 g Pu239 or ≥ 0.04 g Pu-238 (~0.6–

0.7 Ci of either) inside or 

outside glovebox, or

2. Persons with body burden 

>20,000 pCi

Semiannual 

(biannual)

Slight exposure 1. Persons working with <10 g of 

Pu-239 in chemical or 

metallurgical operations

2. Supervisors of quarterly sampled 

category

3. Persons with 25% burden (MPBB 

=0.04 μCi Pu-239)

1. Persons working with Pu but

≤10 g Pu-239 or ≤0.04 g Pu238 in chemical or metallurgical 

operations,

2. Supervisors of the quarterly

sampled category, or

3. Persons with >10,000 pCi body 

burden but <20,000 pCi

1. Working with ≥0.04 g (0.7 Ci) of 

Pu-238, analyzed by RAS

2. Performing chemical or 

metallurgical operations or 

maintenance on systems 

containing ≥10 g of Pu-239 or Pu240 and ≥0.04 g of Pu-238 (0.6–

0.8 Ci) analyzed once by TIMS 

and RAS and once by RAS

Annual 1. Other supervisory personnel

2. Persons working with sealed 

containers of plutonium

3. Other persons with casual 

encounters with plutonium who 

regularly work in areas where 

plutonium is handled

1. Supervisors, persons working 

with sealed containers of Pu,

2. Casual encounters with Pu, or

3. Working with prepared counting 

foils containing >20 mg of Pu239 or 0.08 mg of Pu-238 (~1.3 

mCi of either)

1. Routine work with operations of 

<10 g (0.6 Ci) of Pu-239 or Pu-242 

(0.6–0.8 Ci)

2. Performing maintenance on 

systems containing ≥10 g of Pu239 or Pu-240 (0.6–0.8 Ci) 

analyzed by TIMS and RAS

3. Working with operations of or 

performing maintenance on 

systems with <0.04 g of Pu-238

4. Line supervisors of personnel in 

semiannual categories

5. Transuranic (TRU) glovebox, bag 

outs, etc.

6. ESH-1 radiological control 

technicians (RCTs) who frequently 

enter work areas

7. All personnel with confirmed, 

measurable intakes of Pu-238 

and/or Pu-239

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Frequencya 1944 (Kolodney 1946) 1967 (Lawrence 1967) 1973 (Lawrence 1973)

1998 (Inkret et al. 1998b,c; 

LANL 2004a)

Initial 

(baseline)

All new hires or assigned to work 

with LANL where other persons are 

routinely sampled

1. If there has been risk of exposure 

to plutonium or if exposure history 

is missing or inconclusive

2. Entry requirement for unescorted 

access to some facilities

3. Occasional work in plutonium 

areas but do not handle plutonium

Termination 1. Persons terminating 

employment who have 

previously submitted urine 

samples, or

2. Have been working in major 

plutonium areas, but never 

sampled

Workers who submitted routine 

samples

UPPU Club

(see Section 

5.1)

Complete bioassay and physical 

examination at 5-yr intervals. 

Bioassay continues after termination 

of employment.

Complete bioassay and physical 

examination at 5-yr intervals. 

Bioassay continues after termination 

of employment.

Complete bioassay and physical 

examination at 5-yr intervals. Bioassay 

continues after termination of 

employment.

a. The frequencies above do not apply to working with sealed sources in TA-15 where bioassay is as needed.

b. Samples analyzed by TIMS.

c. Samples analyzed by RAS.

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Table A-16. Routine sampling frequency 241Am exposures.

Frequency 1998 (Inkret et al. 1998d; LANL 2004)a

Quarterly None

Semiannual 

(biannual)

1. Performing operations in gloveboxes with ≥0.2 g of Am-241 (0.6 Ci)

2. Performing maintenance on systems containing ≥0.2 g of Am-241 (0.6 Ci) 

Annual 1. Performing operations in gloveboxes with <0.2 g of Am-241 (0.6 Ci)

2. Frequent entry or performing maintenance on systems containing <0.2 g of Am-241 

(0.6 Ci)

3. Line supervisors of personnel in semiannual categories

4. ESH-1 RCTs who frequently enter work areas

5. All personnel with confirmed, measurable intakes of Am-241 

Initial (baseline) If there has been risk of exposure to Am or if exposure history is missing or inconclusive

Termination Workers who submitted routine samples

a. Samples analyzed by RAS.

Table A-17. Sampling protocol for tritium (Clark 2005).

Period Program Protocol

1950–1970 Biweekly for significant quantities or more often if exposure 

confirmed, removed from tritium work at 124 μCi/L.

1971–1975 >150 μCi/L – 1/d; 150–75 μCi/L – every 2 wk;

74–24 μCi/L – 1/wk (Healy 1970)

1975–1998 Routine – every 2 wk; or 2 hr after expected exposure then 

if <1 μCi/L - no more samples, 1–10 μCi/L – 1 wk in next 

month, 10.1–100 μCi/L – weekly samples, >100 μCi/L –

daily samples (Inkret et al. 1998a)

1998–present Work on regular or intermittent 

basis with or on systems that 

have contained 1 Ci in HT and 

any other form, or 0.1 Ci HTO, 

or MT, or 0.1 Ci of organic 

tritium

Every 2 wk 

Table A-18. Uranium routine sampling protocol (Clark 2005).

Period Protocol

Pre-1983 Biweekly samples collected

Collected on Fridays, minimum annual

08/1983–06/1993 Spot sample collected on Monday mornings before entering work area

07/1993 Last voiding on Sunday night and second on Monday morning, first voiding

1998 Spot samples every 2 wk/persons performing hands-on work/potential for 100 mrem; 

machining operations, polishing operations, foundry work, chemistry operations in which 

>10 g of U, work with oxidized metal, >100 g of bulk powder.a

a. Source: Inkret et al. (1998c).

Table A-19. Sampling protocol for accidental plutonium exposures, 1967 (Lawrence 1967).

Severity class Description Sampling protocol

PA (prompt action) Most serious accidents

Injection detectable by plutonium wound 

monitor (~0.01 μg)

Chemical burns from plutonium solutions

Facial contamination >20,000 cpm

Nose swipes >500 dpm

Evening after accident

10 d later

1 mo later

Each 14 mo thereafter.

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Severity class Description Sampling protocol

DA (delayed action) No known equipment failure, 10 × MPC for 1 

wk or 50 × MPC for 1 d

Skin contamination in excess of limits

Superficially contaminated cuts that are 

positive on surface monitoring only.

Equipment failure causes exposure for 

indeterminate period.

1 mo after accident

11 more samples at monthly 

intervals. No entry is made in “No. 

of days accident prior to sample” 

column of PUQFUA.

NRS (nonroutinely 

sampled)

Small wounds occurring in major plutonium 

areas (DP-West, Ten Site, CMR 

Building)

Exposures without respirators at 10 × MPC 

for period less than 1 wk

Nose counts >50 dpm

Skin contamination >500 cpm/60 cm2

1 urine sample collected 1 mo after 

intake. 

Automatic 

rescheduling

Detect unexpected high exposure and verify 

its existence

Detect contaminated urine sample

Previous sample is 1 dpm/sample

Request another sample

Persons routinely excreting plutonium 

are exempt from special sampling 

but are maintained on routine 

sampling

Table A-20. Sampling protocol for accidental plutonium exposures, 1973 (Lawrence 1973).

Severity class Description Sampling protocol

PA (prompt action) Most serious accidents

Injection detectable by plutonium wound 

monitor >2,000 pCi (DL 600 pCi Pu-239 

and 200 pCi Pu-238)

Chemical burns from plutonium solutions

Facial contamination >40,000 dpm

Nose swipes >500 dpm

Evening after accident

10 d later

1 mo later

Each 12 mo thereafter.

Fecal samples might be collected 

DA (delayed action) No respiratory protection, 10 × MPC for 1 wk 

or 50 × MPC for 1 d

Skin contamination after decontamination in 

excess of limits

Superficially contaminated cuts that are 

positive (500 dpm) on surface monitoring 

only. 

Fecal collected on day 2 for inhalation

1 mo after accident fecal and urine

11 more urine samples at monthly 

intervals. Possible fecal

Potential accident entry is made in 

PUQFUA

NRS (nonroutinely 

sampled)

Small wounds occurring in major plutonium 

areas (DP-West, Ten Site, CMR 

Building) no activity detected. <500 dpm

Exposures without respirators at 10 × MPC 

for a period less than 1 wk

Nose counts >50 dpm

Skin contamination >1,000 dpm/60 cm2

Potential accident date is noted in PUQFUA 

and no sample has ever been submitted 

or routine is not scheduled for at least 

3 mo.

1 urine sample collected 1 mo after 

intake.

Fecal samples for types 2 and 3.

Memorandum sent to H-1; potential 

accident entry in PUQFUA might 

be required

Automatic 

rescheduling

Detect unexpected high exposure and verify 

its existence

Detect contaminated urine sample

Previous sample is 1 dpm/sample

Request another sample

Persons routinely excreting plutonium 

are exempt from special sampling 

but are maintained on routine 

sampling

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Table A-21. Sampling protocol for accidental plutonium exposures, 1998 (Inkret et al. 1998c).

Severity class Description Sampling protocol

PA (prompt action) Most serious accidents

Injection detectable by plutonium wound 

monitor >0.2 nCi

Chemical burns from plutonium solutions, skin 

contamination >500 dpm alpha by 60-cm2

probe or >0.2 nCi in wound area

Facial contamination >10,000 dpm alpha

Nasal >100 dpm

Airborne >200 DAC-hr without respiratory 

protection.

Skin after decontamination is >1,000 dpm 

alpha by 60-cm2 probe

As requested 

Pu-239, Pu-240

Urine samples days 1, 3, 5 after intake

Fecal samples days 1, 3, 5 –optional

In vivo chest count days 3, 5 –optional

Pu-238

Urine samples on days, 1, 3, 5, 8, 16, 

30, 60, 120, 240 after intake

Fecal samples on days 1, 3, 5 –

optional

In vivo chest count on days 3, 5 –

optional

Urine analyzed by RAS and TIMS

DA (delayed 

action)

Effective dose equivalent >100 mrem; 

committed dose equivalent >1 rem any 

tissue or organ

Present in room when CAM alarms

Other individuals on PA

Positive wound count

Chemical burns; after decontamination skin 

>100 dpm

Facial contamination, before decontamination, 

>1,000 dpm

Nasal swipes; 1 nostril >15 dpm, both sum = 

35 dpm alpha

Average ≥40 DAC-hr, without respiratory

Skin contaminate >100 dpm after 

decontamination.

Wounds contaminated with >250 dpm (by 

alpha probe)

Request by manager, group leader, or 

medical

Typically 1 urine, by RAS and TIMS, if 

Pu-239, or

At direction of ESH-1 personnel.

Chelation therapy Accident scenario exceeds PA by 20–40 

times.

Medical procedure. Interpretation of early 

radiological results under advice of ESH12 dose assessment.

Collection for duration of chelation 

therapy, up to 60–100 d after therapy.

PA schedule followed by monthly 

urine samples.a

Might include blood and fecal.

Or as directed.

Follow-up Any positive routine urine As directed.

a. Samples taken for up to 100 d after termination of the therapy should not be used in the calculation of intake.

Table A-22. Sampling protocol for accidental 241Am exposures (Inkret et al. 1998d).

Severity class Description Sampling protocol

PA (prompt action) Most serious accidents

Injection detectable by americium wound 

monitor >0.2 nCi

Chemical burns from americium-bearing 

solutions, skin contamination >500 dpm 

alpha by 60 cm2 probe or >0.2 nCi in 

wound area

Facial contamination >10,000 dpm alpha

Urine samples days 1,3,5, 

8,16,30,60,120,240 after 

intake

Fecal samples days 1,3,5 –

optional

In vivo chest count days 3,5 –

optional

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Severity class Description Sampling protocol

Nasal >100 dpm

Airborne >200 DAC-hr without respiratory 

protection.

Skin after decontamination is >1,000 dpm alpha 

by 60-cm2 probe

As requested 

DA (delayed action) Effective dose equivalent >100 mrem; 

committed dose equivalent >1 rem any tissue 

or organ

Present in room when CAM alarms

Other individuals on PA

Positive wound count

Chemical burns; after decontamination skin 

>100 dpm

Facial contamination, before decontamination, 

>1,000 dpm

Nasal swipes; 1 nostril >15 dpm, both sum = 35 

dpm alpha

Average ≥40 DAC-hr without respiratory

Skin contaminated >100 dpm after 

decontamination

Wounds contaminated with >250 dpm (by alpha 

probe)

Request by manager, group leader, or medical

Typically 1 urine, by RAS

At direction of ESH-1 personnel.

Chelation therapy Accident scenario exceeds PA by 20–40 times.

Medical procedure. Interpretation of early 

radiological results under advice of ESH-12 

dose assessment.

Collection for duration of chelation 

therapy, up to 60–100 days 

after therapy.

PA schedule followed by monthly 

urine samples.

Might include blood and fecal or as 

directed.

Table A-23. Uranium nonroutine sampling protocol.

Years Nuclide Protocol

1998 Uranium Days 1, 4, 8 after possible incidents [high airborne, high alpha skin contamination with 

>10,000 dpm or nasal swipes over 100 dpm (summed)](Inkret et al. 1998b)

Table A-24. In vitro bioassay (except plutonium and americium) 

sensitivity. See corresponding sections in the document for references. 

Radionuclide Period Typea MDAb

Plutonium U/F See Table 5-7

Americium U/F See Table 5-10

H-3 19 U See Table 5-11

Uranium See Table 5-14

Fission product 1950–1970 U 50 to 100 dpm/L

Sr-90 No information availabled

Cs-137 1965–present U 100 pCi/L

Po-210 1954 U 10 dpm/L

1955–1960 U 0.1 pCi/L

P-32 1975 U 40 pCi/L

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Radionuclide Period Typea MDAb

Th-230 as thorium 1958–1963 U 0.01 μg/L

1963-?–present U 20 μg/L

Pa-231 1985 U 0.88 dpm/24-hr sample

a. U = urine; F = fecal.

b. Source: Gautier (1983).

c. Currently sent to an offsite laboratory.

d. Use the reporting level because results were not reported below this level.

Table A-25. In vivo bioassay routine frequency (Inkret et al. 1998c).

Description Frequency Comments

Baseline As requested Unescorted access into radiological controlled areas, radiological buffer areas, 

or radiological areas. Personnel who occasionally work or visit areas where 

plutonium is handled but do not handle plutonium.

Pu2 Annually Routine work or system maintenance with ≥0.2 g pure Am-241.

Routine work or system maintenance with ≤0.2 g pure americium or frequent 

entry where work is performed with any pure Am-241.

Routine work or frequent entry in any area with beta/gamma-emitting 

radionuclides.

Pu1 Biennially Works with plutonium.

Routine work or frequent entry in any area with Pu-239 or -240, including 

material processing, system maintenance, supervision, or other support.

Routine work or frequent entry in any area with Pu-238.

Personnel who routinely work with plutonium and perform chemical or 

metallurgical operations with <10 g of Pu-239 or -242 or 0.04 g of Pu-238.

TRU glovebox/fume hood operations.

TRU hot jobs.

TRU bag outs.

Glovebox operations.

Fume hood operations.

Hot jobs.

Bag outs.

Maintenance operations on process systems.

Radiological control technician duties.

Pu Semiannually Special request.

Uranium Annual Weekly access to areas where DU is machined or polished, casting and 

cleaning crucibles outside dry boxes, chemical operations, including 

purification and recovery. Handling ≥100 g metal or bulk powder outside 

gloveboxes, or operations with uranium-hexafluoride in uncontained 

systems.

MFPs Semiannually Special request.

MAPs/MFPs Annual Works in high contamination areas and/or in airborne radioactivity area.

Th Annual Works with thorium.

provide excretion rates and deposition patterns necessary to assess the health risks for other workers 

appropriately.

The Plutonium Experiment involving plutonium tracers was undertaken between April 1945 and July 

1947. This was a joint project between LANL and the Atomic Energy Project of the University of 

Rochester School of Medicine and Dentistry. None of the subjects was administered the injections of 

plutonium at Los Alamos. However, the data were collected and analyzed by Wright Langham and 

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others at Los Alamos. The subjects of the experiments were males over 45 years of age, an 18-yearold female, an 18-year-old male, and a 5-year-old Australian male child. The child was also injected 

Table A-26. Amount of plutonium administered to 

subjects by intravenous injectiona and the date of the 

administration (Langham et al. 1980).

Designation 

of the subject

Date of 

administration

Pu injected 

(μg)b

HP-1 October 16, 1945 4.6

HP-2 October 23, 1945 5.1

HP-3 November 27, 1945 4.9

HP-4 November 27, 1945 4.9

HP-5 November 30, 1945 5.1

HP-6 February 1, 1946 5.3

HP-7 February 8, 1946 6.3

HP-8 March 9, 1946 6.5

HP-9 April 3, 1946 6.3

HP-10 July 16, 1946 6.1

HP-11 February 20, 1946 6.5

HP-12 April 10, 1945 4.7

a. Plutonium was administered as Pu+4-citrate in 0.5 mL of 0.41% 

sodium citrate 2H2O.

b. Average standard deviation of determination of dose was 3%.

with radiostrontium and radiocerium. These individuals were selected because they suffered from 

terminal illness and were not likely to live beyond 10 years. Six of the individuals outlived the

predictions including one person who died 44 years later at the age of 80 (Hughes 2000). Urine and 

fecal samples were collected from the living subjects and autopsy samples were collected from three 

patients who died within 30 days (Clark 2005). Intake results are available for 12 subjects as reported 

in Langham et al. (1980). [The original paper was published in 1950 but not declassified until 1971.] 

These intakes are listed in Table A-26. However, it appears that an additional six subjects were also 

involved. These are designated as Chi-X (University of Chicago) or Cal-X (University of California, 

San Francisco, where “X” is the subject number for that location (Hughes 2000).

It was determined that deposition was primarily in the trabecular bone, bone marrow, and liver (Clark 

2005). The subject designation might be identified in the telephone interview.

Other human experimentation was conducted at Los Alamos in later years. Tritium experiments were 

conducted in the 1940s and 1950s. These experiments involved exposures of humans to HT, HTO, 

and HTO inhalation, ingestion, and submersion in vapor. Researchers used themselves and probably 

family members as subjects (Hughes 2000).

In the late 1950s, Los Alamos biomedical researchers orally administered 125I and 131I to a group of 19 

subjects. The cohort included eight children between 4 and 10 years old, as well as three 

adolescents aged 12, 13, and 14. The remainder of the cohort consisted of eight adults whose ages 

ranged from 26 to 46. All of the subjects lived in Los Alamos and were either employed by the 

Laboratory or relatives of employees (Hughes 2000). Each subject was administered 0.01 μCi of 

iodine orally.

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In 1961, research on 137Cs was conducted with a cohort of four men aged 27, 33, 35, and 37. In the 

1960s, four human subjects participated in a study involving 65Zn with a single oral administration. 

The subjects were the researchers, a coworker, and a family member of one of the scientists. There 

were 11 types of experiments involving approximately 130 subjects (Hughes 2000).

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Fission and Activation Product Analysis

The first gross beta urine count was devised in 1947 (LASL 1947). Urinalysis was used to monitor for 

intakes of fission products until whole-body counting was fully implemented in 1970. A 1958 

procedure lists gross beta urine analysis from 90Sr/Y, 140Ba/La, 144Ce/Pr, 89Sr, and gross fission 

products (McClelland 1958). This might be the same as the procedure referenced in 1947. No 

sensitivity is listed for this procedure. According to the procedure, background was counted before 

and after each sample. Healy (1970) lists nonspecific sensitivities of 50 to 100 dpm/L with an 

investigation level of greater than 200 dpm/L. A similar procedure of oxalate coprecipitation and beta 

counting, effective in 1974 (Gautier 1983), lists a sensitivity of 1 to 2 dps/L and an MDA of 25 pCi/L. 

This procedure might have been effective as early as 1950. No further mention of a gross fission 

product procedure has been found. However, a procedure for 137Cs in urine using gamma 

spectroscopy lists an effective date of May 1965 and an MDA of 100 pCi/L (Gautier 1983).

Reactors operated from 1944 to 1992. Records of stack releases exist for 131I and other fission 

products. Room air concentrations and urine bioassay results have been found in office memoranda 

(LASL 1950–1960). Fission product air contamination around the Omega Water Boiler was noted as 

a “special problem” because ordinary filter paper failed to pick up the gas in the air. This air 

contamination was thought to have occurred because of residual 235U from an old spill that was 

lodged in the interstices of the reactor (Shipman et al. 1952).

Some individuals might have been exposed to fission products while off the Laboratory site or during 

weapons testing. In addition, a potential for exposure to 140Ba/La and 90Sr/Y exists for those 

participating in the RaLa chemistry and testing, 1944 through 1963, primarily at TA-10.

Urine bioassay samples were submitted for gross beta and gamma counting. Samples were analyzed 

by gamma spectroscopy (MDAs for specific geometry and nuclides are unknown.) The sample was 

then processed through alkaline phosphate precipitation and counted on a gas flow counter with a 

background of 530 cpm (40% efficiency). This would be an MDA of approximately 120 dpm/sample. 

Samples were counted at least five times over a 3-week interval to determine the half-life of the 

nuclides. For example, if the sample exhibited a 12-day half-life, 140Ba was recorded; for an 8-day 

half-life, 131I was recorded. Results had a background of 40K subtracted (typically 20 cpm/1,500 mL) 

(LASL 1950–1960).

Gas leaks were detected with filter papers and charcoal, which were analyzed by gamma 

spectroscopy. Potentially exposed personnel were analyzed in the human spectrometer (HUMCO). 

Individuals showing elevated background in the HUMCO were counted in the NaI(Tl) counter [9.5- by 

6-in. NaI(Tl) crystal]. Iodine-135 (6.57-hour half-life)/135Xe (9-hour half-life), 131I, and 133I were 

observed in individuals exposed to gas leaks at reactors (LASL 1950–1960; Van Dilla 1959). Through 

the late 1940s and as late as 1961, before the use of charcoal cartridges, iodine was analyzed from 

the paper filter with an assumed collection efficiency of 0.1%. Radioiodine and noble gases are 

released from facilities that performed fission product chemistry [Wing 9, CMR (TA-3) and TA-48], 

medical isotope preparation (TA-53), and research reactors (TA-35). There is a very small 131I release 

(environmental release of less than 1 mCi/yr) from TA-48 operations, and the Omega West Reactor 

(OWR) Facility (TA-2) is a source of radioiodine.

Interpretation of the fission and activation product urinalysis in a way that is meaningful, as 

representative of all the possible fission and activation products to which a worker might theoretically 

have been exposed, is a challenge. The gross beta procedure separated and counted radionuclides 

of alkaline earths and rare earths, such as strontium, yttrium, barium, lanthanum, and cerium. The 

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procedure did not account for the radionuclides of ruthenium, cesium, zinc, cobalt, manganese, 

niobium, or zirconium. The abundances of all the fission products, in relation to each other, varied 

considerably. Certain reactors at LANL operated only briefly. Some exposure might have occurred 

during decommissioning operations and operations with weapons testing. No discussion is available 

for the interpretation of fission product mixtures. However, interviews with current and past LANL 

personnel involved with bioassay indicate that fission products were not considered a significant 

source term for intake among LANL workers.

During the late 1940s and early 1950s, atmospheric testing of nuclear weapons was still being done. 

After such tests in Nevada in the 1950s, these radionuclides were found in urine bioassays for gross 

fission products. The background levels, which were variable, provide a complicating factor for the 

use of gross fission product analyses for dose reconstruction. No specific guidance is available from 

LANL on nuclides or amounts to subtract from bioassay samples. Therefore, any activity detected in 

gross beta urine counts will have to be considered occupational.

The HUMCO whole-body counter was used for screening beginning in the 1950s with follow-up by a 

4- by 8–in. NaI(Tl) detector. Body counting and thyroid counting in the current form were not available 

until 1970. Once whole-body counting was established, fission and activation products were more 

often evaluated by whole-body counting. If fission and activation product potential exposure is 

indicated, refer to Section 5.3 for in vivo bioassay sensitivities. The MDAs are listed for many of the 

fission and activation product nuclides (Vasilik and Aikin 1983).

At present, the LABDR does not contain MFP urine bioassay results, and there is no plan to supply 

results for workers unless specific requests are made for the results. Unless the telephone interview 

indicates that the worker might have received bioassay for MFPs, there is no way of knowing if these 

results exist for the worker. After the mid-1950s, when the HUMCO and the NaI(Tl) detectors were 

first placed in service, in vivo bioassay might have been used to assess the potential of intakes. The 

results of any in vivo bioassay performed before 1969 are not readily available (Hoover 2007).

In addition to workers at the reactor areas, assessment of internal dose or potential internal dose for 

workers involved in exposure to ionizing radiation associated with RaLa operations at TA-10 (Bayo 

Canyon Site), TA-35 (Ten Site), and Buildings H, Sigma, and U (in TA-1) is not possible due to the 

lack of urine bioassay monitoring between September 1, 1944, and July 18, 1963, and the 

unavailability of in vivo monitoring until the mid-1950s (NIOSH 2006). These workers, primarily 

chemists but including other operations and maintenance workers and security personnel, had the 

potential for exposure to 90Sr, 89Sr, 140Ba, and 140La.

Similarly, sufficient records for radionuclides other than tritium, polonium, plutonium, and uranium 

have not been found that would allow the reconstruction of internal doses for most TAs with a history 

of radioactive material use. For this reason, only partial dose reconstructions are possible for LANL 

workers with employment before January 1, 1976 (see Section 5.1.3).

Strontium

Records of 90Sr urinalyses, routine or special, are very sparse. Inkret et al. (1998c) stated that four 

bioassays were performed for 90Sr in 1997. It is estimated that approximately 200 bioassays were 

performed for 90Sr during the history of the site. The historical compilations of procedures do not list a 

specific 90Sr urinalysis procedure. It appears that any record of 90Sr analysis indicates that LANL 

performed a gross beta analysis (which was determined to be representative of 90Sr) or sent a sample 

to an outside laboratory. Strontium-90 dose currently can be reconstructed only when 90Sr results are 

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listed for an individual. No information is available on an MDA for 90Sr analysis. Section 5.3.1

discusses the early HUMCO whole-body counters as being capable of counting the bremsstrahlung 

from 90Sr.

Exposure to 90Sr can be expected for persons working in the RaLa Program because 90Sr 

contamination was present in the material used for the program. Persons frequenting the area of the 

shots and those involved with the extraction chemistry would have had a potential for exposure to and 

intake of 90Sr. Persons from TA-1, Sigma, H, and U Buildings, where RaLa operations were 

conducted, would also have the potential for intakes of 140Ba and 140La. The RaLa sources were 

prepared by Group G-7 or G-6 workers at the TA-10 Chemical Process Building from 1944 to 1950. 

This function moved to TA-35 (Ten Site) from 1951 to 1963. The name of the site is likely tied to the 

operating group, CMR-10. The CMR-10 group relocated to Ten Site sometime between April and 

December 1950. The Idaho Chemical Processing Plant became the source of purified 140Ba in 1956, 

and a typical shipment was about 40,000 Ci of 140Ba. The 140La sources prepared at Ten Site were 

usually in the range of 2,000 to 4,000 Ci. Almost 2 million Ci of 140Ba had been handled at Ten Site by 

1963 when the RaLa Program was terminated. The TA-35 RaLa cell and control room have been

completely dismantled. Barium-140 and 140La would have been detectable in the HUMCO screening 

and quantified in the NaI follow-up, although no MDA is available for those nuclides. The Savannah 

River Site lists an MDA of 9.3 nCi for that period (ORAUT 2005b). However, it is not possible to 

reconstruct internal dose or potential internal dose from 90Sr for workers with potential exposure to 

ionizing radiation associated with RaLa operations at TA-10 (Bayo Canyon Site), TA-35 (Ten Site), 

and Buildings H, Sigma, and U (in TA-1).

Actinium

Notebook 6489, containing 227Ac in vitro bioassay results, has been found. Only 15 analyses are 

recorded in the notebook. All analyses were performed in 1954. No other records of 227Ac bioassay 

have been found. Results were 0 to 0.6 dpm/sample.

Potential worker encounters with 227Ac occurred during the decommissioning of TA-21-153 in 1978. 

However, all workers on the decommissioning wore full-face particulate respirators and 

anticontamination clothing. Workers were regularly monitored with nasal swipes and in vivo bioassay 

during the project. Results of the bioassay and nasal swipes were below detection levels (Harper and 

Garde 1981).

Shipman et al. (1952) mentioned the determination of 226Ra as the only practical method “at present” 

for estimating individual exposures to actinium.

No other information is available. Results for any bioassay that might have been performed for the 

individual will not be present in the LABDR. These results will be available only by special request.

Phosphorus-32 and Carbon-14

Both of these radionuclides were encountered as labeled compounds. Standard biokinetics do not 

apply to labeled compounds. A bioassay procedure for 32P in urine lists an effective date of August 

1975 and an MDA of 40 pCi/L. This is also the MDA listed in a procedure in 1983 (Gautier 1983). No 

bioassay procedure has been found for 14C.

Results for any bioassay that might have been performed for the individual will not be present in the 

LABDR. These results will be available only by special request.

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Thorium

A procedure for a coprecipitation technique and a colorimetric final determination was listed as 

effective January 1963. The MDA for this analysis was listed as 20 μg/L. Dummer (1958) listed the 

MDA for the colorimetric procedure as 0.01 μg/L. While plutonium can coprecipitate with thorium, the 

colorimetric determination prevented interference from the plutonium. An earlier procedure titled “The 

Determination of Thorium230 in Urine” was listed in McClelland (1958). This procedure used lowbackground alpha proportional counting and lists a sensitivity of 0.05 dpm (no volume units) of 230Th. 

The procedure stated that natural thorium, plutonium, some americium, curium, actinium, and 

neptunium are carried over with the precipitate.

Only the results of the radiometric analysis of 44 urine samples for 230Th in 1958 have been found 

thus far. A sensitivity of 0.2 dpm (no volume units) was listed for these analyses. Dummer (1958) 

suggested the use of alpha proportional counting for 230Th bioassay samples. Thorium-230 is also 

referred to as ionium. Results for any bioassay that might have been performed for the individual will 

not be present in the LABDR. These results are available only by special request.

Shipman et al. (1952) discussed monitoring of thorium progeny in air from the decay of 232Th. 

Casting, machining, and other operations with thorium metal were undertaken in 1951. This suggests 

that 232Th was in use during this period. Thorium-232 has been identified in environmental emissions 

from waste sites at LANL. After 1980, some work with thorium-based materials was performed. The 

thorium was probably 232Th.

The isotopic mixture should be based on the facility. Conversations with former LANL personnel 

suggest that most exposure to thorium would be incidental to the exposure to decay products of 

uranium. However, statements about work with thorium metal would seem to indicate otherwise. 

Building 159, Sigma Complex, is designated as a Thorium Storage Building for the storage of ingot 

and oxide forms, which would suggest 232Th and progeny. Therefore, the dose reconstructor should 

use best judgment for selection of the isotope or mixture. Thorium-230 is the maximizing conversion 

from mass units. However, it is unlikely, given the current structure of the database, that results from 

thorium bioassays will ever be submitted for a worker unless specific requests are made for the 

results. In that situation, the isotope that was encountered would probably be known.

The Tiger Team Assessment report (DOE 1991) indicated that line managers were not aware that 

thorium and its decay products were internal radiological hazards and that workers handling gram 

quantities of dispersible thorium oxide powders and other thorium compounds should be identified by 

the checklist system for participation in the bioassay program.

Default absorption types for thorium are M and S. The absorption type should be selected based on 

the compound expected or the matrix (in the case of 230Th incidental to uranium exposure).

If RAS rather than total alpha results of 232Th are encountered, the contribution from decay product 

nuclides should be included. Thorium-232 decays with a 6.7-year half-life to 228Ra, which decays with 

a 5.75-year half-life to 6.13-hour 228Ac, which decays to 1.91-year 228Th. See Table B-1.

Table B-1. Specific activity of thorium isotopes.

Isotope Specific activity (pCi/μg)

Th-228 8.1946E+08

Th-230 2.0184E+04

Th-232 1.0966E-01

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Protactinium-231

A procedure for a coprecipitation technique and proportional alpha counting is documented in 

McClelland (1958). The MDA is listed as 0.88 dpm/24-hr sample. No other mention of this analysis is 

found in later procedure documents or reports.

Results for any bioassay that might have been performed for the individual will not be present in the 

LABDR. These results are available only by special request.

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BIOASSAY FOR RADIONUCLIDES OTHER THAN PRIMARY RADIONUCLIDES

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ATTACHMENT C

RESPIRATORY PROTECTION PROGRAM

Respiratory Protection Program

The first reference to the need for respiratory protection for LANL workers was in a telex that 

requested “assault-type gas masks for testing in connection with toxic effects of X10 product 

[plutonium] dust” (Oppenheimer 1944). The May 31, 1944, Health Safety Rules for Buildings H and 

V-8 specified that “[a]n approved respirator or gas mask must be worn during operations requiring the 

exposure of the dry material to the air and during wet operations involving spray or splatter” (LASL 

1944a). Positive pressure masks were being introduced into the program at that time (LASL 1944a). 

Dry boxes and respirators were a satisfactory method of handling dry material; however, individuals 

who worked with solutions consistently received the highest exposure despite all precautions.

Respiratory protective devices were provided during the Manhattan Project by the U.S. Army 

Chemical Warfare Laboratories. Use of the M-9 mask with either the M-11 canister for gases and 

particulates or the M-14 canister for particulates began in 1946. The efficiency of the canister was 1 

in 100,000 particles of dioctyl phthalate (Adley et al. 1963). This mask continued to be used as a 

standard until commercially produced masks were available. By the time the handbook on radiation 

protection (Dummer 1958) was published, a variety of U.S. Bureau of Mines-approved commercial 

respirators and self-contained breathing apparatus was available. Approved respirators were used 

where possible. The Industrial Hygiene group maintained facilities for testing respiratory protection 

equipment and could make recommendations on equipment. The use of Wilson-type respirators was 

discussed in 1973 (Hempelmann, Richmond, and Voelz 1973).

The Industrial Hygiene group at LANL continued to be at the forefront of development of respiratory 

devices and programs. Respiratory protection programs are documented beginning with Dummer 

(1958). The programs included training in the use of and fit testing for air-purifying respirators and 

self-contained breathing apparatus. Protection factors assumed for respiratory protection equipment 

are listed in Table C-1. The protection factors appear to remain consistent throughout the history of 

the program within the exceptions noted.

Table C-1. Respiratory protection apparatus.

Equipment Protection factor Remarks

Half-mask Not more than 10 × MPCa b, c

Full-face mask Not more than 50 × MPC (Dummer 1958)d b

Full-face mask Not more than 100 ×, MPC (Healy 1970; LANL 1984)e b

Full-face mask with fit field tested Not more than 1,000 × MPC (Healy 1970) b

Full-face mask with chemical canister Used for organic gases and vapors up to 2% b

Air-line equipment Used above limits of air purifying respirators; no specific 

protection factors listed

Self-contained breathing apparatus Used above limits of air purifying respirators; no specific 

protection factors listed

a. Assumed 20% penetration because of fitting limitations and 50% workday usage.

b. Air purifying respirators shall never be used in an atmosphere immediately dangerous to life and health.

c. Half-face respirators were not permitted to be used in airborne highly toxic dust such as plutonium (Shipman 1964). 

LANL (2004) stated that half-face respirators are not approved for use with radiological materials or airborne 

radioactivity.

d. Only one style was available at that time; masks were assumed to be only 98% to 99% efficient.

e. Powered air-purifying respirators not more than 3,000 times with high-efficiency particulate air filters.

Respirator mask fit and proper usage remained a significant challenge throughout the development of 

the program. In addition, the decision to don a respirator is often based on the alarm of a continuous 

air monitor in the area.
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Maximizing and Best Estimate Intake Parameters

Intake parameters can be derived from airborne contamination levels for buildings with the highest 

exposure potentials or highest intakes for various periods. Incidents and intakes are listed here in 

Table D-1. 

Table D-1. Reported exposure incidents and results.

Year Incident

1944–1946 26 workers received intakes estimated to be 6 to 80 nCi while involved in plutonium operations: 

purification, fluorination, metal reduction, and recovery. Average intake = 58 nCi (Voelz et al. 

1979); 9 of 12 persons with highest exposures were working with water-soluble plutonium salts in 

1945 (Hempelmann and Langham 1953).

1944 Periodic overexposure of individuals working on RaLa Project occurred for 6 mo at start of project 

until bugs were worked out of remote chemical handling procedures.

Highest nasal swipe for May was 11,372 cpm (assume alpha).

August – Worker opened a sealed container of active material without a respirator or face shield.

1945 August – Building 52 handled large amounts of polonium. Note at bottom of 52 air concentration 

data table stated: “0.75 cpm/L is used as 2-year tolerance value assuming 100% retention by 

lungs.”

4 persons exceeded safe amount of plutonium in body, 1 μg (Hempelmann 1946).

Tolerance value for polonium in urine samples 1,500 cpm/d (assume 50% efficiency for counter), 

exceeded by only 2 individuals. All persons working with polonium were monitored (Hempelmann 

1946).

At the Water Boiler at Omega, several instances of mild to moderate overexposure when gas 

exhaust lines leaked or during decontamination of active material.

Summer – 14 workers removed from plutonium production because urine assays indicated body 

burdens at or above the MPL of 7 cpm/24-hr sample (Hempelmann, Richmond, and Voelz 1973).

Workers entered contaminated side of the air filters areas at DP East without approval (Cox 

1945).

1946 April 15 – Tuballoy fire, Tech Area, firemen exposed without protective equipment (Tribby 1946b).

1948 November 2 – worker from GMX ingested sufficient amount of RaLa to give a reading on detector 

(LASL 1949a).

Radioactive specks found near Bayo Canyon ranged from 0.1 to 10 μCi of RaLa at time of origin 

(LASL 1948).

1949 Difficulties with RaLa separations and extremely large quantities of material being worked with in 

Bayo Chemistry Area cause spread of contamination and few slight overexposures (LASL 1949b).

1949–1951 Number of airborne contamination results above plutonium MAC (0.0044 cpm/L) decreased from 

40% of samples in 1949 to 7% in 1951 for DP West and 25% to 17% for Tech Area. Above-MPL 

nasal swipes decreased from 40% in 1949 to 15% in 1951 for DP West and 60% to 15% for Tech 

Area (Shipman et al. 1952).

1951 1,876 plutonium urine samples below MAC and 492 polonium urine samples from DP East were 

below tolerance. All routine tritium urine samples were below MAC (Shipman et al. 1952).

DP West dismantle 408 and 413; high airborne plutonium exposures were kept to minimum 

(Shipman et al. 1952).

DP East experimental program with actinium. No overexposure; however, adequate urine 

analysis is not available, air samples difficult to evaluate (Shipman et al. 1952).

Pajarito Warehouse contaminated by leaking polonium source. TA-33, Area 6 required cleanup 

from polonium spill (Shipman et al. 1952).

Ruptured plutonium slug at GT site vault. Contamination spread to main building. Air tolerance 

limits reached (Shipman et al. 1952).

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DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL DOSES

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Year Incident

Airborne alpha activity in Room 513, DP West was consistently above MAC. Source could not be 

localized; however, cleaning attic and sealing openings to room decreased airborne levels to 

satisfactory (Shipman et al. 1952). MAC total alpha plutonium as of April 1951 was 0.0044 cpm/L.

Kilocurie quantities of tritium handled in CMR areas. “Health program for this work has been 

satisfactory” (Shipman et al. 1952).

Late 1950s Radioiodine Experiments: LANL biomedical researchers orally administered I-125 and I-131 to 19 

subjects. 

1955 A vial containing americium and curium (91%) exploded, contaminating 2 workers on the face and 

clothing (LASL 1955).

1956 Some high urine counts in uranium areas for employees working with incinerators without 

respirators: Sigma Building, TU Building, Tuballoy Shop, M Building, and Building 96 (McKown 

1957).

1957 TA-18 air samples in vicinity of Honeycomb assembly were one-half of tolerance level, and in 

vicinity of Lady Godiva assembly air samples were “excessively high.”

February 12 – Godiva reactor went supercritical and contaminated the kiva (Buckland 1958).

May 17 – Spill of Sr-90 within H-7 undetected. Contamination spread to homes (Buckland 1958).

June 13 – Spill of Dy-165 resulted in airborne and personnel contamination (Buckland 1959). 

Monitoring of individual for urine, blood, and human counter showed only positive for the human 

counter. Suspected the human counter to have surface contamination. Worker’s badge read 

2 rep (Buckland 1958).

Dismantling tritium equipment in TA-33 resulted in the contamination of three workers, urine 

concentrations up to 650 μCi/L (Buckland 1958)

1958 May 9 – Spill of Dy-165 resulted in airborne and personnel contamination, worker had 20 mR/hr at 

nostril (Buckland 1959).

Total body counts performed on cyclotron personnel indicated measurable Zn-65 (LASL 1959).

1959 April – Radioactive gas leak at LAPRE II resulted in thyroid uptakes by at least five workers of I131, I-133, and I-135 resulting in thyroid burdens equivalent to approximately 3 μCi I-131 (Van 

Dilla 1959).

Laboratory air sample 3.8 dpm, 16% of MPC for 1-wk sample, highest value February to July 

(LASL 1950–1960).

August – Three employees >MPL, 11 >50% MPL, 90% of people with ≤10% MPL for plutonium 

(0.04 μCi) of 1,325 monitored (LASL 1959)

1960 January 11 – Tritium exposure, 476 μCi/L-urine sample (Buckland 1961).

March 2 – A power outage disabled exhaust ventilation resulting in the contamination of four 

laboratories (Buckland 1961). 

March 15 – Wind carried plutonium contamination from an assembly that was being identified to a 

member of H-3. Gross contamination and an intake of plutonium resulted. Contamination spread 

to the Administration building before being detected (Buckland 1961).

December 1 – Activated Na-24 in glass wool contamination (Buckland 1961). 

1971 Wing 9 of CMR – Inhalation of Pu-238, minimal urine excretion for 100 days, then rose to large 

values. ICRP 30 model modified to time constant of 10,000 days (ICRP 1979; Miller et al. 1999); 

half-time for ICRP 30 model is 10,000 days, AMAD is 0.2 μm. Intakes 2,150 to 210 nCi, found to 

fit other Wing 9 intakes.

1972–1973 Demolition of Building 12, DP Mesa, produced no bioassay above detectable and only four 

detectable nose swipes. Highest nose swipe was 85 dpm (Christensen, Garde, and Valentine 

1975)

1977 January – Oxide of plutonium caught fire, airborne 4.5 to 11.65 dpm/m3, no positive nasal swipes, 

no bioassay (University of California 1977b).

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DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL DOSES

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Year Incident

January – H-3 urine bioassay after January 1 incident at LAMPF <1 μCi/L (University of California 

1977a).

February – One employee received plutonium intake estimated at 1 MPBB (0.04 μCi) (University 

of California 1977a,d).

February 10 – Incident resulted in <16-nCi lung burdens for five employees; one employee 

received diethylene triamine pentaacetic acid (DTPA) chelation therapy. Urine sample plutonium 

intake estimates are 1 nCi (one worker), 4 nCi (two workers), 9 nCi (one worker).

May – U-235 airborne concentrations of >1,500 dpm/m3 and >25,000 dpm/m3 for DP Bldg 4, 

Room 412 (University of California 1977b).

May – TA-35 and TA-41 alpha contamination escaped from pressure vessel, caused personnel 

contamination (University of California 1977d).

June – Alpha contamination (up to 100,000 cpm) from damaged uranium target found on several 

persons at LAMPF switchyard (University of California 1977d).

June – Tritium in urine 101 μCi/L after airborne release DP, Bldg 5.

20,000 dpm alpha airborne (U-235 and U-238 operations). 

November – Tritium bioassay (confirmed) of [redacted] employee 798 μCi/L (University of 

California 1977c).

1978 UF6 tube ruptured (5-10 g) spreading HF fumes and depleted U-238 (University of California 

1978a).

[Title redacted] replacing leaking hose at TA-33 stack had tritium bioassay of 4 μCi/L; 449 Ci HT 

released (University of California 1978b).

1979 [Redacted] employee with nasal swipe of 553/70 dpm TA-21 (University of California 1980a).

1980 TA-33 tritium exposure to employees.

54-μCi/L tritium bioassay TA-33 (University of California 1980b).

153-μCi/L tritium bioassay TA-33 (University of California 1980c).

1983 Plutonium spill in TA-55, 10 persons, no significant exposures (LANL 1983a).

High airborne contamination results in high nasal swipes TA-55 plutonium (LANL 1983b).

21 plutonium wound counts, 3 = 0.1 nCi, 1 had 11.2 nCi Pu-239 and 0.4 nCi Am-241 and was 

surgically excised (LANL 1983c).

Quarter 4 – One wound above 20 nCi, numerous positive (>50 dpm) nasal wipes (LANL 1983b).

Particulate filtering respirators were available and were used from the beginning of the program. 

Therefore, ambient air concentrations might not reflect the actual breathing air concentration of the 

workers.

Research continues to identify the intakes for both maximizing and general conditions. Examples of 

intakes identified thus far are listed in Table D-1; Table D-2 summarizes a particular incident. As the 

state of the respiratory protection program and engineering controls improved, the potential for 

chronic and acute intakes was reduced.

NOTE: While an attempt has been made to report only incidents with quantitative results, some 

incidents for which only qualitative comments were available in the records have been included in the 

list. Information is not available for every year of operation.

ATTACHMENT D

DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL DOSES

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Table D-2. Summary of February 10, 1977, plutonium incident (University of California 1977d).

Worker

Total systemic burden (increase) Nasal swipe (left/right) (dpm) Chest burdena

Pu-239 nCi Pu-238 nCi Total alpha Pu nCi

[Redacted] 1 8.2 (3.3) 1.7 (0.1) 12,759/13,385b 7.5

[Redacted] 2 7.6 (5.8) 0.7 (0.5) 801/1,849 13.

[Redacted] 2 42.9 (7.5) 3.2 (0.6) 1,091/1,193 6.

[Redacted] 3 21.8 (18.0) 1.1 (0.4) 611/1,193 10.

[Redacted] 4.2 (0.3) 0.7 (0.0) 226/577 0

a. No worker had chest counts above MDA (16 nCi).

b. Received DTPA chelation therapy.

Rather than preventing intakes completely, the focus of the respiratory protection program, especially 

in the early years of operation, was to not exceed allowable concentration limits. The MPBB for 

plutonium was 5 μg from 1943 to 1945. During 1945, when many of the exposures occurred, the 

tolerance dose of plutonium was assumed to be 1 μg. Therefore, a suitable method of measuring 

personnel exposure had to be able to detect from 2 to 14 dpm of plutonium in a 24-hour sample of 

urine.

The MACs for buildings with a high potential for exposure were established by 1947 and changed 

over the years as technology and understanding of the hazards increased. MACs were eventually 

replaced by MPCs and derived air concentrations (DACs) as controlling values. It appears that at 

some time before 1964, the MAC values were changed to ALs with larger, more specific MAC values 

established for nuclides, enrichments, and mixtures.

ATTACHMENT D

DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL DOSES

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ATTACHMENT E

SCALING OF COWORKER DOSE INTAKES CONSISTENT WITH 

TYPE AND DURATION OF EXPOSURES

Scaling of Assigned Intakes Using Case-Specific Information

The major presumptive exposures at LANL were made up of the primary radionuclides plutonium, 

uranium, tritium and, in early days, polonium. Other radionuclides may 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 dry boxes, potential exposures are relatively unlikely. NIOSH has 

determined that it is not feasible to reconstruct doses for these radionuclides before the year 1996. 

Work locations are another important consideration when evaluating for a potential unmonitored 

intake. 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 the table below.

Table E-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 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, maintenance, 

health physics personnel who are 

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.

50th

Dose Assignment

Fiftieth-percentile intakes are used to calculate organ doses based on the listed lognormal distribution 

of intakes, along with the listed GSD. Ninety-fifth-percentile intakes are assumed to be constant 

values (distribution-free). The actual intake values are calculated using the relationship below:

95th-percentile intake = 50th-percentile intake × GSD1.645 (E-1)
Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 114 of 117

TABLE OF CONTENTS

SECTION TITLE PAGE

F.1 Purpose ............................................................................................................................... 114

F.2 Inhalation Exposures............................................................................................................ 114

F.3 Methodology ........................................................................................................................ 114

F.3.1 Breathing Rate ......................................................................................................... 115

F.3.2 Assumptions............................................................................................................. 115

F.4 Intakes ................................................................................................................................. 115

F.5 Results and Recommendation ............................................................................................. 116

F.1 PURPOSE

This attachment presents bounding estimates of potential organ doses for exposure to airborne 

radioactive materials on the Los Alamos site during the Cerro Grande fire in May 2000. 

Measurements of airborne radioactivity were conducted throughout the event as described in 

Measurements of Air Contaminants during the Cerro Grande Fire at Los Alamos National Laboratory

(Eberhart 2010). Low values for measured concentrations (in attocuries per cubic meter) resulted in 

small intakes and low doses.

F.2 INHALATION EXPOSURES

Airborne concentrations of radioactive material from the fire are described in Eberhart (2010). The 

most abundant radioactive materials resulted from resuspension of radon progeny that had 

accumulated for many years on vegetation and on the forest floor. Concentrations of plutonium, 

americium, and uranium were also measured, but these results were, in general, consistent with 

measurements that were made outside the fire period (Eberhart 2010). Gamma spectroscopy 

identified only naturally occurring radionuclides. For the purpose of an overestimate of potential 

internal dose to Los Alamos workers, intakes for plutonium, americium, and uranium were calculated 

based on the highest measured values without considering likely background intakes. That is, 

material of natural origin was not rejected for the purpose of the maximum dose calculation described 

here.

F.3 METHODOLOGY

Intakes were estimated using the maximum measured concentrations of 241Am, 238Pu, 239Pu, 234U, and 

238U as follows:

( )

3

3

dpm m hr

m hr

Total intake Concentration Inhalation rate Exposure time    = ××        

(F-1)

ATTACHMENT F

BOUNDING ESTIMATES OF ORGAN DOSES FROM POTENTIAL INHALATIONS OF 

RADIOACTIVE MATERIALS DURING THE CERRO GRANDE FIRE AT LOS ALAMOS

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Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 115 of 117

F.3.1 Breathing Rate

Standard breathing rate assumptions could underestimate actual rates for emergency workers who 

responded to the fire. For this reason, the breathing rate was adapted from ICRP (1975), Table 120

as follows:

m T V f V time =× × (min) (F-2)

where: 

Vm = volume of air per minute

f = frequency of inhalation breaths per minute (1/min)

VT = tidal volume, which varies with activity level

In this case, f and VT were selected for an adult performing heavy work (ICRP 1975, Table 120), with 

the following result:

( )

3 1 min mL m 21 2,030 mL 60 2.56E 06 2.6 min hr hr hr

Vh 

      

= × × =+ =              

(F-3)

where:

Vh = volume of air per hour

F.3.2 Assumptions

Additional assumptions were:

• A bounding intake was used to overestimate worker internal dose consequences,

• A 60-hour exposure period was assumed,

• Individuals were assumed to spend the entire 60-hour period at the location of the highest 

potential intake,

• The highest measured values for each alpha emitter were identified from the data in Eberhart 

(2010) (these values are listed in Table F-1),

• The heavy work breathing rate was assumed for the entire 60-hour period, and

• Respiratory protection was not worn.

F.4 INTAKES

Intakes calculated from the methodology described above are listed in Table F-1.

ATTACHMENT F

BOUNDING ESTIMATES OF ORGAN DOSES FROM POTENTIAL INHALATIONS OF 

RADIOACTIVE MATERIALS DURING THE CERRO GRANDE FIRE AT LOS ALAMOS

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Document No. ORAUT-TKBS-0010-5 Revision No. 02 Effective Date: 04/18/2013 Page 116 of 117

Table F-1. Highest measured air concentrations of alpha emitters and consequent intakes.

Radionuclide Date Location

Concentration Intake

(aCi/m (dpm) Concentration Source 3) (dpm/m3)

Am-241 May 12 TA-54 125 2.78 E-04 0.043 Eberhart 2010, Table A-16

Pu-238 May 13 McDonald’s 65 1.44 E-04 0.023 Eberhart 2010, Table A-17

Pu-239 May 13 Los Alamos Inn 472 1.05 E-03 0.17 Eberhart 2010, Table A-18

U-234 May 13 TA-5 (formerly 

TA-52, Beta Site)

1,709 3.79 E-03 0.59 Eberhart 2010, Table A-19

U-238 May 13 TA-5 (formerly 

TA-52, Beta Site)

4,188 9.30 E-03 1.45 Eberhart 2010, Table A-21

F.5 RESULTS AND RECOMMENDATION

Organ doses that would result from the listed intakes are likely to bound doses due to conservative 

assumptions for location, exposure time, and breathing rate. Doses were estimated using IMBA and 

assuming an acute inhalation of each of the listed intakes, over a number of organs, as well as a 

cancer diagnosis in May 2010. For radionuclides with multiple absorption types, the type that resulted 

in the highest dose was selected.

Calculated doses to a several organs are listed by radionuclide in Table F-2. The highest total dose 

from a single radionuclide to any organ was 2.92 × 10-04 rem (0.29 mrem) to the thoracic lymph nodes 

(LN(TH)) for an assumed absorption type S intake of 238U. Doses to all other organs, including the 

lung, thoracic lymph nodes, and red bone marrow, were less than this value. Two isotopes of 

uranium were measured, and both are present in the background and from anthropogenic sources. 

Doses were totaled for each organ over the five radionuclides as shown in Table F-3. The largest 

total dose was 4.75 × 10-04 rem (0.48 mrem).

Because the dose results listed above are below the level regarded as significant for the calculation of 

probability of causation, 0.001 rem (1 mrem), and because dose reconstruction practice includes the 

assignment of a larger dose based on assumptions for environmental internal dose that are favorable 

to claimants, it is unnecessary to include dose estimated for exposure during the Cerro Grande fire.

ATTACHMENT F

BOUNDING ESTIMATES OF ORGAN DOSES FROM POTENTIAL INHALATIONS OF 

RADIOACTIVE MATERIALS DURING THE CERRO GRANDE FIRE AT LOS ALAMOS

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Table F-2. Calculated doses by radionuclide and organ.

a

Radionuclide Organ

Dose (rem)

(absorption type) Radionuclide Organ

Dose (rem)

(absorption type)

Pu-238

Lung 2.38 E-06 (S)

Pu-239

Lung 1.53 E-05 (S)

RBM 8.09 E-07 (M) RBM 5.80 E-06 (M)

LN(TH) 5.79 E-06 (S) LN(TH) 4.19 E-05 (S)

Bone Surface 1.05 E-05 (M) Bone Surface 7.54 E-05 (M)

Liver 2.21 E-06 (M) Liver 1.60 E-05 (M)

ET2 3.09 E-06 (S) ET2 2.25 E-05 (S)

U-234

Lung 4.51 E-05 (S)

Am-241

Lung 2.49 E-06 (M)

RBM 8.00 E-07 (F) RBM 1.33 E-06 (M)

LN(TH) 1.34 E-04 (S) LN(TH) 3.61 E-07 (M)

Bone Surface .11 E-06 (F) Bone Surface 2.18 E-05 (M)

Liver 8.43 E-07 (F) Liver 2.87 E-06 (M)

ET2 7.35 E-05 (S) ET2 1.01 E-06 (M)

U-238

Lung 8.68 E-05 (S)

RBM 1.82 E-06 (F)

LN(TH) 2.92 E-04 (S)

Bone Surface 1.55 E-05 (F)

Liver 1.84 E-06 (F)

ET2 1.56 E-04 (S)

a. Italics indicate the highest single-radionuclide dose to any organ.

Table F-3. Total organ doses for all evaluated 

radionuclides.

Organ Total dose (rem)

Lung 1.52 E-04

RBM 1.06 E-05

LN(TH) 4.75 E-04a

Bone Surface 1.30 E-04

Liver 2.37 E-05

ET2 2.56 E-04

a. Highest total dose to any organ from all five radionuclides.

ATTACHMENT F

BOUNDING ESTIMATES OF ORGAN DOSES FROM POTENTIAL INHALATIONS OF 

RADIOACTIVE MATERIALS DURING THE CERRO GRANDE FIRE AT LOS ALAMOS

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