ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Los Alamos National 

Laboratory – Occupational Internal Dose

Document Number:

ORAUT-TKBS-0010-5 

Effective Date: 12/21/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 96

Subject Expert: Rowena S. Argall

Document Owner

Approval: Signature on File Date: 11/16/2004

Jack E. Buddenbaum, TBD Team Leader

Approval: Signature on File Date: 11/15/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 11/15/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 12/21/2004

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

5

Acronyms and Abbreviations ...............................................................................................................6

5.0 Occupational Internal Dose......................................................................................................9

5.1 Introduction ..............................................................................................................................9

5.1.1 Bioassay Results of Individuals ........................................................................................ 11

5.1.2 Bioassay Program............................................................................................................ 12

5.2 In Vitro Bioassay ....................................................................................................................13

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

5.2.2 Americium ........................................................................................................................22

5.2.3 Tritium..............................................................................................................................23

5.2.4 Uranium ...........................................................................................................................25

5.2.5 Fission and Activation Product Analysis ........................................................................... 29

5.2.6 Polonium ..........................................................................................................................31

5.2.7 Other Limited-Exposure Radionuclides ............................................................................ 32

5.3 In Vivo Minimum Detectable Activities, Analytical Methods, and Reporting Protocols ............ 34

5.3.1 Whole-Body Counters ...................................................................................................... 34

5.3.2 Lung Burdens...................................................................................................................38

5.3.3 Wound Monitoring ............................................................................................................ 40

5.4 Interferences and Uncertainties.............................................................................................. 41

5.4.1 Contamination..................................................................................................................41

5.4.2 Uncertainties ....................................................................................................................41
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5.5 Air Concentration in Selected Buildings.................................................................................. 42

5.5.1 Respiratory Protection Program ....................................................................................... 42

5.5.2 Maximizing and Best Estimate Intake Parameters............................................................ 46

5.6 Unmonitored Workers ............................................................................................................ 48

5.6.1 Workers Pre-1947 ............................................................................................................ 51

5.6.2 General Guidance ............................................................................................................ 53

References ........................................................................................................................................54

Glossary ............................................................................................................................................62

Attachment 5A Occupational Internal Dose for Monitored Workers..................................................64
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LIST OF TABLES

Table

5-1 Historical in vitro bioassay ....................................................................................................

Page

13

5-2 Plutonium respiratory tract absorption type........................................................................... 14

5-3 Activity composition of nominal 3% plutonium mixture.......................................................... 16

5-4 Activity composition of reference weapons-grade 6% plutonium mixture .............................. 16

5-5 Plutonium bioassay sensitivity as listed in procedures and reports ....................................... 17

5-6 Fresh plutonium mixture ....................................................................................................... 19

5-7 Activity composition of reference fuel-grade (12%) plutonium mixture .................................. 19

5-8 241Am bioassay techniques and sensitivities ......................................................................... 23

5-9 Tritium urine bioassay sensitivity levels ................................................................................ 24

5-10 Uranium conversion tables ................................................................................................... 25

5-11 D-38 uranium – nominal weight composition and fractional activities.................................... 25

5-12 Or-93 uranium – nominal weight composition and fractional activities .................................. 25

5-13 Routine uranium urinalysis detection levels .......................................................................... 27

5-14 Natural uranium....................................................................................................................29

5-15 Routine 210Po urinalysis detection levels............................................................................... 32

5-16 Specific activity of thorium isotopes ...................................................................................... 34

5-17 Routine whole-body counting detection levels ...................................................................... 36

5-18 Mean body burdens of 137Cs from fallout in the United States............................................... 38

5-19 MDA LC values for lung counting .......................................................................................... 39

5-20 Summary of airborne concentrations from selected buildings ............................................... 43

5-21 Respiratory protection apparatus.......................................................................................... 45

5-22 Reported exposure incidents and results.............................................................................. 47

5-23 Summary of February 10, 1977, plutonium incident.............................................................. 48

5-24 Summary of maximum allowable concentrations for selected areas ..................................... 49

5-25 MAC and Action Levels for 1964 .......................................................................................... 50

5-26 Hypothetical chronic intakes for plutonium, 1943-1946......................................................... 51

5-27 Hypothetical intakes for polonium and uranium, 1943-1946.................................................. 52

5A-1 Pathway codes for current in vitro bioassay reports.............................................................. 66

5A-2 Assessment method codes for current bioassay summary reports ....................................... 66

5A-3 In vitro bioassay type and incident verification codes from current dose reports ................... 66

5A-4 Incident line ..........................................................................................................................67

5A-5 Example urine data header information ................................................................................ 67

5A-6 Example bioassay data line .................................................................................................. 67

5A-7 Tritium bioassay in urine (edited databases)......................................................................... 68

5A-8 Uranium bioassay in urine .................................................................................................... 68

5A-9 In vivo report codes .............................................................................................................. 69

5A-10 Current and historical locations and default respiratory absorption type................................ 70

5A-11 Zia employee access to plutonium areas.............................................................................. 82

5A-12 Historical in vitro bioassay .................................................................................................... 82

5A-13 Current in vitro bioassay performed 1997 ............................................................................. 82

5A-14 Routine sampling procedure................................................................................................. 84

5A-15 Plutonium urinalysis—summary of sensitivities and analytical techniques ............................ 84

5A-16 Example urine bioassay results ............................................................................................ 85

5A-17 Routine plutonium sampling frequency ................................................................................. 87

5A-18 Routine sampling frequency 241Am exposures ...................................................................... 89

5A-19 Sampling protocol for tritium ................................................................................................. 89

5A-20 Uranium routine sampling protocol ....................................................................................... 89

5A-21 Sampling protocol for accidental plutonium exposures ......................................................... 90
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5A-22 Sampling protocol for accidental plutonium exposures ......................................................... 90

5A-23 Sampling protocol for accidental plutonium exposures ......................................................... 91

5A-24 Sampling protocol for accidental 241Am exposures................................................................ 92

5A-25 Uranium nonroutine sampling protocol ................................................................................. 92

5A-26 Summary of in vitro bioassay (except plutonium and americium) sensitivity ......................... 93

5A-27 Summary of whole-body counting detection levels ............................................................... 93

5A-28 MDA, LC values for lung counting ......................................................................................... 95

5A-29 Solubility class and absorption type assigned to uranium compounds .................................. 96

5A-30 In vivo bioassay routine frequency........................................................................................ 96
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 12/19/2003 00-A New Technical Basis Document for the Los 

Alamos National Laboratory – Occupational 

Internal Dose. Initiated by Jack E. Buddenbaum.

Draft 06/02/2004 00-B Incorporates NIOSH review comments. Initiated 

by Jack E. Buddenbaum.

Draft 07/30/2004 00-C Incorporates NIOSH and internal review 

comments. Initiated by Jack E. Buddenbaum.

Draft 08/31/2004 00-D Incorporates further NIOSH and internal review 

comments. Initiated by Jack E. Buddenbaum.

Draft 10/15/2004 00-E Incorporates further NIOSH and internal review 

comments. Initiated by Jack E. Buddenbaum.

12/21/2004 12/21/2004 00 First approved issue. Initiated by Jack E. 

Buddenbaum.
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ACRONYMS AND ABBREVIATIONS

25 LASL reference to 235U or the standard ~93% enrichment

28 LASL reference to 238U or depleted uranium

49 LASL reference to plutonium-239

AL (Los Alamos) Action Level

ALI Annual Limit on Intake (ICRP 30); Annual Limit of

AMAD Activity Median Aerodynamic Diameter

Intake (ICRP 60)

Bq Becquerel = 1 disintegration per second

BZA Breathing Zone Air Sampler

CAM Continuous Air Monitor

CDE Committed Dose Equivalent

cpm cpm – counts per minute

CMR Chemistry and Metallurgy Research

D-38 depleted uranium mixture

DAC Derived Air Concentration

DIL Derived Investigation Level

DNAA delayed neutron activation analysis

DOE U.S. Department of Energy

DOP dioctyl phthalate

dpm disintegrations per minute

dpm/m3 disintegrations per minute/ cubic meter

DTPA diethylenetriaminepentaacetic acid

DU depleted uranium (less 235U than natural uranium, nominally 0.72%)

EDE Effective Dose Equivalent

ESH Environmental, Safety, and Health Division

EU enriched uranium (more 235U than natural uranium, nominally 0.72%)

GeLi Germanium Lithium drifted semiconductor detector used for gamma spectroscopy.

GM Geiger-Mueller detector

HPGe Hyper Pure Germanium Detector

HRL Health Research Laboratory

HSE Health, Safety, and Environment Division

HT elemental tritium (tritiated gas)

HTO tritium oxide (water or water vapor)

HUMCO “Human Counter”, whole-body counter used between1955-1970

IA Induced activity 

ICRP International Commission on Radiological Protection

IDLH immediately dangerous to life and health

IMBA Integrated Modules for Bioassay Analysis (computer program)

ionium 230Th –the decay product of 234U 

keV kilo electron volt

Kr krypton
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kW kilowatt

Lc Critical Level

LAMPF Los Alamos Meson Physics Facility

LASL Los Alamos Scientific Laboratory (until January 1981)

LANL Los Alamos National Laboratory (1981 to present)

LSC liquid scintillation counter used for tritium urinalysis starting in 1970

MAC Maximum Allowable Concentration

MDA Material Disposal Area

MDA Minimum Detectable Activity

MDTA Minimum Detectable True Activity 

MeV million (mega) electron volts 

MPBB maximum permissible body burden: limit used from startup until adoption of ICRP 30 

methods, approximately 1990

MPC Maximum Permissible Concentration

MPL Maximum Permissible Level

mR milli (1/1000) R

mrem milli (1/1000) rem

MSMA Minimum Significant Measured Activity

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

MW megawatt

NaI Sodium iodide gamma/ photon scintillation detector, a.k.a. NaI(Tl)

nCi nanocurie 

NCRP National Council on Radiation Protection and Measurements

NDA No Detectable Activity

NIOSH National Institute for Occupational Safety and Health

NTA Neutron Track Analysis

OBT organically bound tritium

OR-93 Oralloy, enriched uranium (Oak Ridge Alloy); either 40% or 93% enrichment 

OU Operable Unit

OWR Omega West Reactor

PF Plutonium Facility

PHA Pulse Height Analysis, used in gamma and alpha spectroscopy analysis

R roentgen (unit of exposure)

rad unit of absorbed dose

RaLa Radioactive Lanthanum Project (radioactive barium and lanthanum implosion tests) 1944

through 1962

RAS radiometric alpha spectroscopy, a.k.a. pulse height analysis (PHA)

rep roentgen equivalent, physical (former unit of absorbed dose)

rem roentgen equivalent, man (dose equivalent)

Sv sievert; the SI unit for dose equivalent. (1 Sv = 100 rem.)

T3O8 uranium oxide based on tuballoy (natural uranium)

TA Technical Area

TIMS Thermal Ionization Mass Spectroscopy
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U.S.C. United States Code

Unat Natural uranium

VFP volatile fission products

WGPu Weapons Grade plutonium 

X10 LANL reference to plutonium

ZnS zinc sulfide scintillation detector for alpha counting

μCi microcurie

μg microgram

α alpha radiation

β beta radiation

γ gamma radiation
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5.0 OCCUPATIONAL INTERNAL DOSE

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning 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 the National Institute for Occupational Safety and Health 

(NIOSH) 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 facility” as defined in the Energy Employee 

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 7384l (5) and (12)). 

Occupational internal dose is the dose received by an individual from an intake of radioactive material 

while performing tasks within buildings and structures of the Los Alamos National Laboratory (LANL), 

which was called the Los Alamos Scientific Laboratory (LASL) until January 1981, or from activities 

outside buildings, such as burial of waste and monitoring of tests, where intakes of radioactive 

material could occur. This document contains information for reconstruction of occupational internal 

doses at LANL facilities throughout its history. 

5.1 INTRODUCTION

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

contamination swipes. Swipes with a lightly oiled filter paper were made 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 more 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. 

Radiation hazards were minimal during 1943. In the spring of 1944, the first few milligrams of 

plutonium arrived at LANL. Until that time, the laboratory was in the construction phase. In 1943, 

hazards of the project were limited to external radiation from the cyclotron, the Van de Graaff, radium 

sources, and a few micrograms of plutonium that arrived in the summer of 1943. In addition, there 

were some 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, showering 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 developing 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 burdens were reduced significantly. Maximum permissible body burdens (MPBBs) began 
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at 5 μg (0.33 μCi) of plutonium in 1943. These were reduced to 1 μg (0.07 μCi) in October 1944 and 

reduced again to 0.6 μg (0.04 μCi) in 1951 (Langham et al. 1962).

Early safety efforts were based on working in safe contamination levels. In 1954, contamination-level 

measurements were based on contact with a shield open (Geiger-Mueller) GM tube for beta/gamma 

and a Pee Wee probe of 55-cm2 area for alpha. The efficiency of these portable probes is 

approximately 10% to 15%. Swipes counted on a fixed proportional counter had an efficiency of 

approximately 40% to 50%, although the oil on the swipes, used to increase 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). 

The tolerance for wounds was 10 cpm for alpha and 0.15 mrep/hr beta/gamma, except wounds 

contaminated with 90Sr for which the tolerance was 0.05 mrep/hr. The skin contamination tolerance 

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

for alpha emitters was 1,000 cpm for polonium, 500 cpm for Tuballoy or Oralloy, and 250 cpm for 

plutonium. The tolerance for floor counts was 100 cpm/swipe alpha but definitely less than 500 

cpm/swipe alpha, and bench tops were usually 50 cpm/swipe alpha or less. Floors in laboratories 

were mopped once or twice a day to maintain safe contamination levels (Hempelmann, Richmond, 

and Voelz 1973). Early Health Group reports indicate contamination inside many of the respirators 

indicating improper storage and handling and poor fit during use (LASL 1944a).

Kilogram quantities of plutonium began to arrive at LASL 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 online, 

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

From 1943, the Health Group was responsible for establishing 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. 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, Environmental, Safety, and Health (ESH) and Health, 

Safety, and Environment (HSE) Divisions], with multiple groups therein 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 

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, as well as 

performed physical examinations and treated industrial accidents.

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

equipment, and safety procedures. Nevertheless, the potential for monitored and unmonitored 

intakes has existed throughout the history of the site. 
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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)

• Americium (241Am)

These radionuclides of primary concern are listed in Laboratory reports from 1943 through 2004. 

They have been associated with work areas and years of operation in Attachment 5A, Table 5A-10. 

Information applicable to both monitored workers (e.g., units, tolerance, reporting levels, and 

frequencies) and unmonitored workers (e.g., establishing maximizing and plausible intakes) is also

listed in Attachment 5A.

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

facilities, laboratory facilities, reactors, accelerators, and others. Many of the exposure histories and 

work records are not specific about the assigned work areas of individuals. However, when 

information concerning the work location is available, Table 5A-10 in Attachment 5A can be used to 

determine the probable nuclide(s) and inhalation absorption type. Section 5.2 discusses in vitro

methods for specific radionuclides. Excreta bioassay methods for determining internal exposures 

were developed in late 1944 for plutonium (fully implemented in April 1945) and polonium, in 1949 for 

uranium, and in 1950 for tritium. Only workers with a significant potential for exposure were 

monitored. Although the number of individuals monitored has increased, not all individuals working at 

LANL are currently monitored. A survey taken in 1986 estimated that approximately 350 persons had 

known burdens of plutonium. 

Blood tests were performed following potential exposures starting in 1944. However, these blood 

tests were performed for blood count levels related to external radiation exposure rather than the 

radioactive content of the blood. Therefore, any records of blood counts performed on an individual or 

mentioned in the claimant interviews will not be directly applicable to internal dose calculations. Air 

samples, identifiable with an individual’s record, were performed beginning in 1944. However, it is not 

apparent how the records of these samples might be available in the individual’s dose records. 

Records of chelation therapy for plutonium and americium are noted in the bioassay record. 

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

begun in 1955, chest counting was begun in 1970, and wound counting was performed beginning 

before 1967. Section 5.3 discusses in vivo bioassay methods used currently and historically at LANL. 

5.1.1

At the time of this technical basis document, LANL had submitted bioassay data for only a few 

individuals. Only summary dose reports have been submitted for all claims. The submitted bioassay 

results were for plutonium only and in an interim format excerpted from the current database. The 

format of these results is not in the final format expected for the data. All descriptive information 

provided in this document is based on speculation of the final format of the data that will eventually be 

supplied. The descriptions listed in Attachment 5A are based on the current structure of data in 

databases. These records are expected to be supplied as text or spreadsheet files. No information is 

currently available on any hardcopy reports that might be supplied for in vitro analysis results. In vivo

reports are available and are described.

Bioas s ay Res ults of Individuals
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Codes and text that describe the assigned intake scenario are often listed in the database of 

plutonium and americium bioassay results. Descriptive codes and format descriptions are also 

addressed in Attachment 5A, Section 5A.1. 

When a dose has been assigned to an individual, the dose summary report might contain the intake 

or incident date used in the calculation. An assumed intake date might also be available on the 

incident line in the current data base. An assumed intake date might be available on the “white card” 

that accompanies the in vivo reports.

If the report of an individual indicates that chelation therapy has been administered, then initially the 

excretion patterns cannot be correlated with the International Commission on Radiological Protection 

(ICRP) or other standard methodology. Historically, approximately 35 workers are known to have 

received chelation therapy for plutonium and americium intakes at LANL. Methods of handling these 

data are discussed in Section 5.4. Contact the internal dosimetry supervisor for guidance in 

interpreting bioassay data following chelation therapy. 

5.1.2

Prior to 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, 

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

task involving 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. 

Bioas s ay Program 

Short-term workers, such as summer students, persons participating in postdoctorate work, and 

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

monitored for internal exposure only in unusual circumstances. Near the termination date, the worker 

might have received an in vivo count. The worker might have been required to submit an initial urine 

sample or have an initial in vivo count. 

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

monitoring program from that used for laboratory employees. Zia employees had a special computer 

program that locked out access to plutonium areas if participation in a plutonium bioassay program 

was not recorded within 425 days. Attachment 5A contains a list of criteria and exempt job 

categories. In 1978, in an attempt to reduce the number of sampled Zia employees to 500 per year; 

supervisors who only performed inspections were eliminated from the schedule.

In the 1970s, LANL initiated 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. 

One small, but significant, group of workers is the UPPU club (translated as “You pee Pu”). This 

group, established by Wright Langham in 1951, consists of individuals who had 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.
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5.2 IN VITRO BIOASSAY

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

This section contains a detailed discussion of the techniques that were used for selected 

radionuclides. Table 5A-26, Attachment 5A, contains a summary of in vitro bioassay sensitivities 

listed by technique and year.

Table 5-1. Historical in vitro bioassay. 

Material Start year Comments

Tritium 1950

Uranium 1949 In practice, 1951

Plutonium 1944

Polonium-210 1944

Americium-241 1954

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 Pending data from LANL Not done regularly

5.2.1

The most serious intakes at LANL have involved isotopes of plutonium and 241Am. The first urinalysis 

performed for evidence of plutonium uptake was performed on July 18, 1944, using a chemical 

procedure developed at the University of Chicago Metallurgical Laboratory in June 1944. However, 

Health Group personnel soon discovered that the procedure, using 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. The first recorded accident in which a human was subjected to a possible intake of 

plutonium occurred on August 19, 1944. This accident resulted in LASL being authorized to proceed, 

under the direction of Wright Langham, to develop a more sensitive procedure. The resulting 

cupferron extraction procedure and subsequent procedures are discussed later in this section (Moss 

1990). 

Plutonium 

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

are predominantly 239Pu. (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 currently and historically encountered at LANL are oxide, nitrate, 

fluoride, and metal. Plutonium oxide is the most common form encountered and should be assumed 

if other information is not available. 

Respiratory Tract Absorption Type 

The absorption type assigned for plutonium isotopes is dependent on the chemical process. Table 

5-2 lists absorption types as related to selected processes (Hempelmann, Richmond, and Voelz 

1973). Table 5A-10 provides information on absorption types associated with locations. 

Sample Collection Procedures

At the onset of the LASL 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 
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Table 5-2. Plutonium respiratory tract absorption type. 

Process/area Isotope Compound Type Comments

Recovery area Pu-239 Plutonium nitrate M In addition, the area might have

oxalates, oxides, hydroxide and 

others based on source of input

Purification Pu-239 Nitrate, oxalate, sodium plutonyl acetate M

Fluorination Pu-239 Oxalate, fluoride, oxide M-S Powder dust, respirators worn

Reduction Pu-239 Metallic, fluoride, chloride S Welding

Heat source production Pu-238 High fired oxides S Possible extended retention and 

delayed excretion

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

procedure was later deemed extremely expensive and was eliminated in 1952. A procedure that 

allowed collection of an equivalent 24-hour sample while the employee was off the site was initiated. 

A modified procedure to collect a 24-hour equivalent (two morning and two evening voidings) sample 

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 dpm/day results. All sample results are expected to be reported in units per 24-hour 

day.

LASL instituted a sample validation procedure (Lawrence 1978). Attachment 5A, Section 5A.3, lists 

the validation procedure used for plutonium urine bioassay samples. This procedure is more rigorous 

than can be implemented without the computer software. However, certain samples in the database 

have been marked invalid based on these criteria. The dose reconstructor should not attempt to 

implement the validation procedure but should use best judgment when evaluating samples for 

inclusion in the intake calculations. 

Missed Intakes

Intakes of plutonium could 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 LASL workers with high and low potential for exposure to plutonium has 

shown that measurable amounts of plutonium, above that which is expected from global fallout, could 

occur in low exposure potential individuals (Foreman, Moss, and Langham 1960). 

The long-term excretion pattern of plutonium isotopes can permit 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 be directly related

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

earlier. Illustrating the magnitude of the potentially missed intake is one case cited in Hempelmann, 

Richmond, and Voelz (1973) that discusses an individual following a potential intake in 1944. A vial 

containing 10 mg of plutonium chloride exploded in a young chemist’s face. The face was thoroughly 

scrubbed and the mouth washed out. Heavy contamination of the face persisted for several days. 

From the ionization of expired air, it was estimated that the level of contamination of the mouth was 

approximately 10 μg. The individual’s urine analysis was less than detection levels in 1944, but more 

recent urine analysis determined that the individual did have a plutonium body burden. In this 

situation, the intake date would be 1944 and the dose should be assessed from that date. 

The first urine tests for plutonium were developed in 1944-1945. Urine analysis was difficult and timeconsuming. Therefore, personnel with the most positive nasal swipes had the most urine bioassays

(ENSR 2002). 
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Routine Sample Frequencies

Tables 5A-17 through 5A-20 in Attachment 5A list the routine sampling frequencies by historical 

period. The routine sampling frequencies can provide an upper bound of potential exposure for an 

individual, allow the appropriate grouping of samples into intake regimes, and provide a better 

estimate of intake dates. In 1978, sampling frequencies were reevaluated for LASL 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.

Sample Analysis Procedures

Initially the urine bioassay analysis procedure could not distinguish adequately between plutonium 

and polonium. 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 was in use between October 1949 and January 1957. No corrections were 

made for chemical blanks, recovery factors or counting geometry until 1957. 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) 

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. 

Use of a ZnS counter began in 1966. All of these methods measured total alpha activity from 

plutonium isotopes. Prior to 1967, when alpha spectroscopy was begun, isotopic fractions can be 

selected based on process knowledge, as described below in Plutonium Isotopic Ratios (Mixtures). 

Predominantly pure 238Pu was not encountered at LASL prior to 1968. Plutonium-241 and 241Am were 

not measured by total plutonium alpha procedures and the dose contribution from these nuclides 

might need to be accounted for using the mixtures in Tables 5-3, 5-4, and 5-7, unless the plutonium in 

the intake was known to be a pure isotope. 

From 1967 to present-day operations, counting has been performed by alpha spectroscopy for 

specific plutonium isotopes. In 1997, thermal ionization mass spectroscopy (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 radiometric alpha-spectroscopy (RAS). The sensitivity 

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

does not measure 240Pu. Therefore, the amount of 240Pu associated with the mixture should be 

accounted for using Table 5-3, 5-4, or 5-7. 

Minimum Detectable Activities (MDAs) reported in procedures and reports for the various 

radiochemical and counting methods are listed in Table 5-5. 

Urine, fecal, and tissue samples were analyzed for plutonium. Most in vitro bioassay samples are 

urine. Fecal and tissue samples are performed only, if requested, in special circumstances. 

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 reported results for all periods should be interpreted as 
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Table 5-3. Activity composition of nominal 3% plutonium mixture (Kinderman et al. 1953). 

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

Years of aginga: 0 5 10 15 20 25 30

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-239 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

Activity ratios

Pu-239+240:Am-241 NA 66 37 27 23 20 19

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

Pu-239: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 NA 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.

Table 5-4. Activity composition of reference weapons-grade 6% plutonium mixture. Mixture designation: Fresh 5-year 10-year 15-year 20-year 25-year 30-year

Years of aginga: 0 5 10 15 20 25 30

Specific activity in mixture (Ci/g)

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

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

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

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

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

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

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

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

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

Activity ratios

Pu-239+240:Am-241 NA 12.2 6.87 5.13 4.28 3.80 3.50

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

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

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

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

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

Pu alpha:Am-241 NA 13.6 7.62 5.68 4.73 4.19 3.84

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

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

units/24-hr sample unless otherwise specified in the record. Some samples taken in response to 

suspected acute intakes might have been “spot” samples of less than 24-hr excretion. These sample 

results are expected to be either noted in the database or the results normalized to 24-hr excretion. 

Detection Sensitivities, Reporting Limits, Tolerance Limits

Detection sensitivity was not a primary concern in the early years of operation of LASL. The concern 

was whether a tolerance limit was exceeded. Results below MDA were originally reported as “lessthan” values (“LX.XX”) or 0 until the 1980s, when actual results, positive and negative, were listed in 

the database. In the current database, all results, from all periods, have been listed as actual values, 
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Table 5-5. Plutonium bioassay sensitivity as listed in procedures and reports.

Nuclide Year(s)

Sample 

typec Technique (“era”)

MDA level Reporting limit Tolerance limitj

Unit/24-h samplea Unit/24-h samplea Unit/24-h samplea

Pu–Total 

alpha

1944–1949 U Cupferronh 0.8 pCie (1 dpm DL (Moss 1990)) >0.8 pCi 7 cpm/24 hr or 6.3 pCi

1949–1/1957 U Bi-phosphate/alpha 

counting

0.4 pCi e 2 dpm or 0.9 pCi 7 dpm/24 hr or 3 pCi

1/1957–1965 U Aluminum nitrate/NTA 0.05 dpm at 99% confidence 

or 0.03 pCif

0.2 dpm or 0.09 pCi 7 dpm/24 hr or 3 pCi

1966 U ZnS 0.07 dpm or 0.03 pCig

Pu-239 1967–1997 U RAS (PHA) only starting in 

1971

0.03 pCii(1mBq)

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 earth

Oxalate 

Rapid

Alkaline earth- 0.015 pCi/800 cc

Oxalate – 0.015 pCi/700 cc

Rapid – 1.6 pCi/L

1997–2002 U TIMS/RAS (alpha specb ) 7.6 μBq/24-hrd

1.03E-4 pCi/ 24-hr sensitivity

1997–present U RAS (alpha spec) 8.1E-3 pCi or 300 μBq

2003–present U TIMSk 3E-4 pCi 

Pu-238 1967–1971 U RAS (alpha PHA) 0.03 pCii 0.2 dpm/24hr 

investigate

1971–1976 U RAS (alpha PHA) 0.07 dpm or 0.03 pCii

1977–1997 U RAS (PHA) 0.03 pCii

1977–1981 F RAS (PHA) 0.4 nCi/ sample

1981–1983 F Phoswich detector, 4-cm 

sample thickness, 

0.2 nCi / sample, (17-keV X-rays)

1997–present U RAS (alpha spec) 8.1E-3 pCi or 300 μBq

a. Unless otherwise noted.

b. Source: Inkret et al. (1999a). 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. 

c. U=urine, F=fecal

d. 2x sensitivity

e. Source: Lawrence (1978).

f. Source: McInroy et al. (1991).

g. Source: Moss et al. (1969).

h. “A successful method of analyzing urine was developed in Jan. 1945 but could not be used as a routine test until a contamination free laboratory (ML Building) was 

ready for use in Feb. 1945.” (Hempelmann 1945)

i. Source: LASL (1978a)

j. Tolerance Limit – Considered the level under which it was ‘safe’.

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. 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 18 of 96

positive or negative. When not specified, two times the detection level (or uncertainty of the blank) is 

considered MDA. Average chemical recoveries and matrix blanks were not reported prior to 1957. 

Therefore, the values reported in the tables for MDAs prior to 1957 are actually the results that would 

be reported if an analyzed personal 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 on whether the value was positive; therefore, these are being treated as MDAs in this 

document (Lawrence 1978). 

Prior to the start of RAS analysis in 1967, plutonium results were analyzed as total alpha

plutonium. These results are found in the 239Pu database, 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, without the contribution from 240Pu. Before using 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. 

Tables of plutonium bioassay sensitivities, as listed in procedures and reports, are found in Table 5-5. 

Also listed in Table 5-5 are Reporting Limits and Tolerance Limits for certain years. Tolerance Limits 

were defined as the level that could be accepted as “safe.” Reporting Levels are those values above 

the MDA that are considered significant for recording or follow-up. However, the database, when 

completed, should have the actual bioassay results even if they are below the reporting level. 

Calculating pCi/24-hr for Urine Bioassay Results

All plutonium bioassay results records are processed through the current database prior to reporting. 

Therefore, all results, from all periods, are expected to be reported in pCi/24 hr, with associated 

uncertainties reported as one standard deviation. However, should the need arise to calculate 

plutonium urine bioassay analysis results that are not fully processed, equations in Lawrence (1978) 

can be used to convert the listed units to pCi/24 hr. Select the method based on the “era” of the 

samples.

Plutonium Isotopic Ratios (Mixtures)

Dose assessments performed at LANL were and are currently based solely on the isotope(s) 

identified in the analysis. When total alpha measurements were made for plutonium, 239Pu was 

chosen based on process knowledge. Results of these total alpha measurements prior to 1967 

are found in the database and listed as 239Pu. Since 1967, plutonium analysis is performed with 

alpha spectroscopy. 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. Thus, when 

quantities of 238Pu arrived at LASL in 1968, the analysis was capable of distinguishing between the 

isotopes of plutonium. Since 1997, TIMS analysis is available and is capable of distinguishing 

between 239Pu and 240Pu. Results listed for TIMS analysis will be only 239Pu. Typically, dose 

calculations performed in dose reconstructions, including those performed with the Integrated 

Modules for Bioassay Analysis (IMBA) computer program, assume a combination of 239+240Pu, 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.

No definitive historical information on the 240Pu:239Pu atom ratios of LASL/LANL sources exists, nor is 

there any information on how the ratios vary with time and location. However, prior to 1970, 

plutonium at LASL was Hanford-derived (Gallaher and Efurd 2002). According to environmental 

impact studies, atom ratios from LASL plutonium released prior to 1970 would have been 0.01 to 0.03 

240Pu:239Pu. LASL plutonium was Savannah River Site-derived starting in 1970. The atom ratio of 

LASL plutonium release after 1970 was expected to be 0.05 to 0.07, although ratios up to 0.13 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 19 of 96

indicated the sediment was affected by LASL plutonium rather than purely global fallout. Ratios 

above 0.07 in environmental samples were indicative of a mixture of global fallout and LASL 

plutonium. However, Table 5-6 describes the plutonium mixture found in 1960 in dirt 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, pre-1970 

mixtures of 3% listed in Table 5-3 are suggested when 

isotopic ratios are not known (Gallaher and Efurd 2002). 

Table 5-4 lists activity and weight ratios as referenced for U.S. Department of Energy (DOE) sites for 

6% weapons-grade plutonium (WGPu). These ratios are based on the presumed age of the material 

for DOE sites. There was practically no americium present in the plutonium handled at LASL in 

the1944-1945 period (Voelz, Grier, and Hempelmann 1985). Throughout the history of the site, 

plutonium mixtures are usually considered to be fresh and, therefore, do not initially contain a 

significant amount of americium. 

Although trace quantities (by weight) of 238Pu have always been present in WGPu, relatively pure 

238Pu did not arrive at LANL until 1968. At LANL, the relatively pure 238Pu has always been processed 

in separate dry boxes and gloveboxes, in later years, from WGPu to prevent cross-contamination of 

the two forms. Tables 5-3, 5-4, and 5-7 list the relative activities of plutonium isotopes and 241Am, 

which grows in from 241Pu, for 3% 240Pu (Gallaher and Efurd 2002), 6% 240Pu and 12% 240Pu mixtures 

(Carbaugh 2003). In these tables, “aging” refers to the time since the 241Am was separated from the 

plutonium.

Table 5-7. Activity composition of reference fuel-grade (12%) plutonium mixture. 

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

Years of aginga: 0 5 10 15 20 25 30

Specific activity in mixture (Ci/g)

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

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

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

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

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

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

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

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

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

Activity ratios

Pu-239+240:Am-241 NA 3.64 2.05 1.53 1.28 1.13 1.04

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

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

Pu-241:Pu-239+240 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 NA 4.39 2.46 1.82 1.51 1.33 1.22

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

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

Since 1967, 238Pu and 239+240Pu were measured separately; therefore, if positive results are identified 

in the sample, the ratio of one to the other can be used to estimate (1) the category of the plutonium 

mixture and (2) the activities of 241Pu and 241Am. 

Table 5-6. Fresh plutonium mixture 

(Christensen, Garde, Valentine 1975) 

Isotope % by weight

Pu-239 93.5%

Pu-240 6.0%

Pu-241 0.5%
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Prior to 1967, the measured quantity was total alpha from plutonium, which means the total of 238Pu 

and 239+240Pu. Total plutonium results are treated as 239Pu in the database. Most plutonium mixtures 

handled at LANL should be assumed to be nominally weapons-grade (6%) or, prior to 1957, 3%. 

The nature of the process work at LANL, however, involved research-grade plutonium mixtures 

usually considered to be fresh. While the plutonium mixture encountered in processes should be 

assumed to be fresh, aged plutonium might have been encountered during building demolitions and in 

other areas where plutonium or plutonium contamination might have been in place for many years. 

Any 241Am observed in lung counts performed years after the intake can be assumed to be the result 

of the in-growth of 241Am from the 241Pu in the mixture over time or the 241Am in the initial plutonium 

mixture unless the incident report specifically indicates potential exposure to pure 241Am or urine 

bioassay for 241Am is found for the intake period. Assume 3% or 6% plutonium mixture unless other 

information is available in the dosimetry records. 

In addition, pure 238Pu might be encountered. A typical plutonium isotopic composition, by 

weight, with a specific activity of 14.1 μCi/μg, of the material released to the atmosphere in 

1970 during an accident in Technical Area (TA)-21 (DP West) involving pure 238Pu is 238Pu 

(80%), 239Pu (16.9%), 240Pu (2.8%), 241Pu (0.2%), and 242Pu (0.1%) (Meyer 1970). Pure 238Pu is 

typically found in areas involving heat source technology. 

Fuel-grade plutonium (12%) was not routinely encountered at LANL. However, for maximizing dose, 

10-year-old, 12% fuel-grade mixture can often be claimant-favorable. For intakes since about 1996, a 

20-year-old mixture could be assumed. See Table 5-7. 

Validation of Samples

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. The dose reconstructor should review the sample information 

carefully and make a decision on the relevance of the sample result(s). Section 5A.3 in Attachment 

5A lists the protocol followed by the LASL computer software to validate a sample result. This section 

is provided for information only, to allow the dose reconstructor to possibly understand the criteria 

that were used to invalidate the sample result(s). This method should not be used by the dose 

reconstructor in the validation process.

The excretion rate following an intake 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 2001. Attachment 5A, Section 5A.9, contains 

actual bioassay results (urine bioassay and nasal swipes) 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. These results can be useful to the dose reconstructor when judging the 

variability of results to determine sample validity. 

Excretion Rate

Pulmonary absorption type should be assumed according to work location or material composition, if 

known. 

Special cases of excretion kinetics have been observed at LANL. The 1971 Wing 9 incident involving 

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

A gradual rise in excretion rate was observed following that time, with no additional intake suspected 

(Miller et al. 1999). In addition, infrequent cases of plutonium inhalations at other DOE sites, probably 

associated with calcined or high-fired oxides, have been known to exhibit a low urinary excretion rate 

and a long retention time in the lung, consistent with the “bound state” in the ICRP (1994a) lung 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 21 of 96

model. The dose reconstructor should be aware of this possible type of behavior when evaluating a 

series of bioassay results. This is especially important when establishing the intake date because 

there might be several negative bioassay samples following the incident before the first positive 

sample. In these situations, the plutonium-to-americium ratio of the bioassay results is significantly 

different than the ratio of the inhaled product. Further, results of lung counts are significantly different 

from the lung burden predicted by the urine bioassay results (Hammond, Lagerquist, and Mann 1968). 

If chest count data show a long retention time, the dose reconstructor should use their best judgment 

on forcing the data to fit a type-S model or decreasing the absorption rate constants in the lung model.

Sampling Protocol for Special Accidental Exposure

Sampling protocols for special and accidental exposures are listed in Attachment 5A. The information 

might be useful to the dose reconstructor in estimating the date of intake if no date is apparent in the 

records. The incident line(s) in the plutonium database might list information concerning any incident 

related to the results. These records might provide more precise information on known or suspected 

intake dates. In addition, the record of nasal swipes, positive or negative, can aid the dose

reconstructor in determining a plausible intake date. However, absence of a positive [above the 50-

dpm Maximum Permissible Level (MPL)] nasal swipe should not invalidate an intake. In practice, a 

negative nasal swipe might have precluded further bioassay. The sampling protocol can provide 

guidance to the dose reconstructor for grouping bioassay samples into an intake regime. Tables 

5A-21 to 5A-23 in Attachment 5A list the historical and current protocols for sampling for plutonium 

exposures. 

Nasal Swipes

Until the first urine bioassay analysis was perfected and available in February 1945, “nose counts” 

were relied on to be a qualitative indicator of plutonium intake (Hempelmann 1945). After the 

development of urine bioassay techniques, nasal swipes were used to indicate the need for follow-up 

bioassay, although bioassay was not always performed immediately following a positive nasal swipe

(McInroy et al. 1991). The MPL was 50 dpm, alpha, per nostril. 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 follow-up bioassay being performed by LANL. In 

addition, in 1944 LASL began the practice that if the two nasal counts varied significantly, the higher 

nasal count was considered spurious (LASL1944b). Information concerning a positive nasal swipe 

can often be found in the incident line of the database entry 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.

Computer-Based Calculations and Sorting

Beginning in 1959, LASL routinely employed 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. 

Of relevance to dose reconstruction, these computer codes also 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 5A.4 of Attachment 5A.
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All plutonium and americium data currently reside in the Los Alamos Bioassay Data Repository 

database. Data results, current and historical, have been converted to units of activity (typically pCi) 

per 24 hours. All results are reported in actual values (e.g., positive, negative, or zero). Uncertainty 

values are listed for each measurement. Codes for the interpretation of the records are in Section 

5A.1 of Attachment 5A. 

5.2.2

At LANL 241Am is usually encountered as a trace contaminant in plutonium. However, there is 

potential for exposure to pure 241Am at LANL. No exposure to pure 241Am was likely prior to the 

beginning of the americium bioassay program in 1954, although a procedure for determining 241Am in 

urine was in development in 1948 (LASL 1948). 

Americium

Current and historical bioassay results are stored in the Los Alamos Bioassay Data Repository

database with the plutonium results. Codes and formats of these results are likely to be identical to 

those described in Attachment 5A, Section 5A.1, for plutonium bioassay results.

There is an 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. Conversely, if the selected plutonium mixture, based on plutonium 

bioassay results, indicates the presence of americium, the absence of americium bioassay should not 

preclude the calculation of the dose from the americium contribution to the mixture. 

Prior to the mid-1990s, very few individuals submitted urine samples that were analyzed for the 

presence of 241Am. Samples were submitted in response to incidents involving exposures to sources 

of 241Am not contained in the plutonium mixture. Estimates of internal doses are based on the results 

of chest counts and urine bioassay (Inkret et al. 1998a). Urinalysis is the principal bioassay method 

used for the assessment of intakes. ICRP (1994b) lists the respiratory absorption type for americium 

as type M. However, if 241Am is determined to be part of a plutonium mixture or in-growth from the 

intake of a plutonium mixture, the absorption type assigned to the plutonium mixture should be 

assumed for the 241Am.

The current routine and special sampling programs have protocols similar to those for the plutonium 

program. Table 5A-18 in Attachment 5A describes the routine sampling frequency. Table 5A-24 

describes the current program of special sampling frequencies. The sampling protocols might aid the 

grouping of samples into intake regimes or identification of an intake date. 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. Lack of a positive result on a nasal swipe should not eliminate the possibility of an 

intake.

Minimum Detectable Activities 

Historically, emphasis was not on individuals being below the MDA, but rather being below the 

tolerance level. All bioassay results are expected to be reported in pCi/24-hr sample; therefore, the

listed urine bioassay MDAs have been normalized to pCi/24-hr sample. 

Historically, results below MDA were reported as less-than values (“LX.XX”) or 0 until the 1980s, 

when actual results, positive and negative, are listed in the database. In the current database all 

results from all periods have been listed as actual values, positive or negative. When not specified, 

two times the detection level (or uncertainty of the blank) is considered MDA. Table 5-8 lists MDAs. 

MDAs are based on MDAs listed in procedures and reports. 
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Table 5-8. 241Am bioassay techniques and sensitivities.

Sample 

type Year Method MDA

Urine 1954 – 1957 Unknown 9.E-01 pCi/24hra

Urine 1958 – 1982 Chemical extraction/ 

proportional counting

2.E-01 pCi/24hrb

Fecal 1977 Phoswich 4.E-02 nCi/sample

Urine 1983c – 2002 Co-precipitation/alpha spec 1.5E-02 pCi/24hr

Fecal 1983c Am/Pu screening/Phoswich 1.E-02 nCi/sample

Urine 2003 – present Alpha spec 1.0E-02 pCi/24hr

a. No MDA available, use DIL; Tolerance level 3.1 pCi/sample. 

b. Source: Milligan et al. (1958); method can carry over thorium, plutonium, curium, 

actinium, and neptunium. Exact end date to the start of this MDA is not known.

c. Source: Gautier (1983); exact start date of the MDA is not known.

d. Source: Inkret et al. (1998a). 

e. Sent to an offsite laboratory.

When no MDA value is provided with the sample, dose reconstructors should use the values in Table 

5-8 to determine whether a listed bioassay result is positive or to calculate missed dose using 

detection levels. It is not expected that MDAs will be provided with results because all results are now 

reported in actual values. MDAs in Table 5-8 are listed in pCi/24-hour sample. At present, the 

electronic database contains results in units of pCi/24-hour sample. 

5.2.3

Tritium was encountered in several forms: tritiated water (HTO), tritiated gas (HT), organically bound 

tritium (OBT), and metal tritide (MT). Each form has unique characteristics. Each year approximately 

100 individuals are monitored for tritium intakes at LANL. If the form is not indicated or the work area 

is not specific, then, with the exception of metal tritides, assume HTO, which is the form generally 

encountered. However, in the HRL and IBF Buildings, the predominant form of tritium was OBT.

Tritium

Tritium incidents have been recorded for accelerator areas. 

Organically Bound Tritium

The first approximation of the dose from OBT (labeled) compounds is the tritium in body water dose. 

However, the absorption, distribution, and excretion of tritium-labeled compounds are specific to the 

chemical and physiological behavior of the particular compound. Specific guidance is available in 

OCAS-TIB-002 (OCAS 2003a) when the dose records indicate an exposure to tritium-labeled 

compounds.

Locations such as the HRL and IBF list labeled compounds as the principal type of radionuclides 

encountered at these facilities. Intakes of labeled compounds do not follow the same biokinetics as 

the ICRP Publication 68 defaults (ICRP 1994b). OCAS-TIB-002 (OCAS 2003a) discusses the IMBA 

model for OBT. 

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. If urine bioassay 

results are interpreted as HTO, when the actual exposure was to MT, the calculated dose to the lung 

and the whole body will be an underestimate. Inkret et al. (1999b) shows that the particle size at 

LANL is approximately 2 to 10 μm. In simulated lung fluids, the dissolution rate of 4.7 × 10-6/day and 

a corresponding biological half-time in the lungs of 1.5 × 105 days were observed. Long-term 

excretion of the dissolved tritium component can be observed for periods up to 50 days after an acute 
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intake. The derived investigation levels (DILs) range from 0.005 - 5 μCi/L compared with 5 μCi/L for 

HTO.

OCAS-TIB-002 (OCAS 2003a) provides guidance on the evaluation of MT intakes. Metal tritides are 

referred to as Tritium Particulates. Information on building locations and years of operation when MT 

might have been encountered is not currently available. The claimant telephone interview can provide 

indications that a person was exposed to MT. If the dose reconstructor encounters a case involving 

exposure to MT, discussion among the task leadership might be needed to determine the appropriate 

biokinetic model. 

Interpretation of codes and units of the bioassay data for tritium is discussed in Attachment 5A, 

Section 5A.1. 

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 (McClelland and Milligan 1954), the tolerance for tritium in urine was 250 μCi/L. 

The analytical range was 1 to 250 μCi/L (5.18 × 104 to 1.30 × 107 Bq/24 hr). In 1958, an internal GM 

counting technique was used. The reported MDA for this method was 1 μCi/L (5.18 × 104 Bq/24 hr). 

In 1970, liquid scintillation counting began and is still in use today. With the introduction of liquid 

scintillation counting, the detection limit was 1 μCi/L (5.18 × 104 Bq/24 hr). The MDA consistently 

improved. 

For tritium results, the denominator used for reporting purposes was per liter (/L) of urine. All data 

from 1950 through 1992 were converted to μCi/L in 1992. However, results in the current database 

are listed as Bq/24-hr sample. At present, all results are likely to be reported in units of Bq/24-hr 

sample. Tritium bioassay results can be found as early as 1946.

Results for samples can be listed as “0” or coded as “LX.XX” 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 being entered into bioassay 

results when the value was below minimum detection levels (Lawrence 1990a). MDA values are 

listed in Table 5-9 and summarized in Attachment 5A, Table 5A-26. All results in the current database 

are likely to be reported in actual values, positive, negative, or zero.

Table 5-9. Tritium urine bioassay sensitivity levels.

Time Detection level Reporting level Counting method

1950 – 1951 5.18E4 Bq/24 ha (1 μCi/L) Electroscope

1952 – 1953 5.18E4 Bq/24 ha (1 μCi/L) GM counter

1954 – 1957 5.18E4 Bq/24 h (1 μCi/L)

1958 – 1968 5.18E4 Bq/24 h (1 μCi/L) Internal GM

1969b – 1987 1.04E3 Bq/24 h (0.02 μCi/L)b 5.18E4 Bq/24 h (1 μCi/L) LSC

1988 – 1998 5.18E2 Bq/24 h (0.01 μCi/L) 5.18E3 Bq/24 h (0.1 μCi/L) LSC and 1 ml raw urine

1999 – present 2.59E2 Bq/24 h(0.005 μCi/L) LSC

a. Expected to be the same as 1954

b. Source: Gautier (1983).

Routine Sampling

The routine sampling protocols for tritium bioassay throughout the history of the program are listed in 

Attachment 5A, Table 5A-19. Routine sampling protocols can provide an upper bound to intake and 

assist in grouping of samples into intake regimes.
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It should be noted that the Kanne Chamber used to monitor tritium releases at the accelerator was 

recalibrated in 1977 and found to be calibrated high by a factor of x10. The previous calibration had 

been done 15 years earlier.

5.2.4

Historically, uranium was primarily either depleted (D-38, 28) or enriched (Or-93, 25) at LANL with a 

variety of isotopic ratios (Little, Miller, and Guilmette 2003a). Table 5-10 provides generic uranium 

conversion factors derived from Rich et al. (1988). Tables 5-11 and 5-12 describe the nominal weight 

composition and fractional activity for each mixture (Lawrence 1990b). However, if specific isotopic 

information is not available the default values in IMBA can be used in the dose reconstruction.

Uranium 

Table 5-10. Uranium conversion tables.

Enrichment 

percent

Fraction by activity Fraction by mass Total

U-234 U-235 U-238 U-234 U-235 U-238 U pCi/μg

0.1 0.225 0.010 0.765 0.00001597 0.001 0.99898 0.438

0.17 0.260 0.010 0.730 0.00002 0.0017 0.999 0.465

0.2 0.285 0.010 0.705 0.00002178 0.002 0.997978 0.476

0.3 0.340 0.010 0.650 0.00003 0.003 0.997 0.514

0.5 0.410 0.020 0.570 0.00004 0.005 0.995 0.591

0.7 0.480 0.020 0.500 0.00005 0.007 0.993 0.668

0.711 0.486 0.022 0.492 0.00005 0.007 0.993 0.672

0.72 0.486 0.022 0.492 0.00005 0.007 0.993 0.675

3 0.745 0.045 0.210 0.00019 0.030 0.970 1.57

90 0.970 0.029 0.001 0.010 0.900 0.090 62.1

93 0.972 0.027 0.001 0.010 0.930 0.060 65.1

95 0.973 0.026 0.001 0.010 0.950 0.040 67.2

Table 5-11. D-38 uranium – nominal weight composition and fractional activities

(Lawrence 1990b).

Isotope U-234 U-235 U-236 U-238 Total

Weight fraction 0.00002 0.003 0.000003 0.997

Curies in 1g 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 5-12. Or-93 uranium – nominal weight composition and fractional activities

(Lawrence 1990b).

Isotope U-234 U-235 U-236 U-238 Total

Weight fraction 0.011 0.933 0.002 0.054

Curies in 1g 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

Depleted (D-38) is the most common form of uranium encountered at LANL (Inkret et al. 1998b), 

although natural uranium (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 natural 

uranium.) 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). However, the partition between Class D and W could not be determined. 

Historically, Class W was used for reporting results (Lawrence 1992a) because it produced larger 
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doses. The equivalent absorption type M is recommended unless other forms are specified in the 

individual’s dose record (Little, Miller, and Guilmette 2003a). 

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, so type M should be used as the default. Therefore, absorption type F or M should be 

considered unless type S is indicated by the fitting of the bioassay or by process knowledge. 

Statements about welding on uranium, both depleted and eventually enriched, which would have 

been a potential for more insoluble forms of uranium are found in LASL (1951). However, if specific 

isotopic information is not available, the default values in IMBA can be used in the dose 

reconstruction.

If the specific compound of uranium is known, Table 5A-29 in Attachment 5A can be used to 

determine the solubility class and/or absorption type. 

The LANL program calculated intakes for individuals having at least one positive urine analysis result. 

The potential for intake was indicated by a positive nasal swipe. A positive nasal swipe was 50 dpm 

in either nostril. Lack of a positive result on a nasal swipe should not eliminate the possibility of an 

intake.

Table 5-10 lists activity and weight fractions for several typical enrichments (Rich et al. 1988). 

Otherwise, dose reconstructors can use the nominal compositions in Tables 5-11 and 5-12 as default 

compositions when no other information is available on the enrichment (mass or weight fraction of 

235U multiplied by 100%) of the uranium, although IMBA defaults are also appropriate. These are the 

compositions found at LASL from 1970 through 1990, at least (Lawrence 1990b).

Sampling Protocol

The routine and non-routine sampling protocols provide information that is helpful in estimating intake 

dates and grouping samples into intake regimes. Historical sampling protocols are listed in 

Attachment 5A, Table 5A-20. The program was initiated in 1949 with analysis for uranium mass, 

purely as an indicator of exposure. Samples were collected on Fridays just before leaving work 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 going into the work area. The sample might also have consisted of the last voiding on a 

Sunday evening and/or the first voiding on a Monday morning, if extra volume was necessary. This 

protocol ensured that the large fraction (approximately 80%) of rapidly excreted uranium experienced 

on the first day following an intake was excluded from the sample (Lawrence 1984). Beginning in July 

1993, the larger volume of sample required by the alpha spectroscopy method of analysis made the 

Sunday evening/ Monday morning collections standard protocol (Lawrence 1992b). 

Retroactive calculations were made by Lawrence of the data beginning in 1949. Data for 1951 was 

not included in the retroactive calculations performed in 1992 by Lawrence because the bound Los 

Alamos Notebooks (LANB) could not be found. These data might now be available. For 235U, 

enriched uranium, the conversion of data into the computer files and retroactive calculations started 

with the 1955 data. Assays identified as 238U were converted to μg/day with values less than MDA or 

0 set to the ”minimis” MDA. Assays identified as 235U were converted to pCi/day with values less than 

MDA or 0 set to the “minimis” MDA. 
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Uranium Analysis Techniques

Several analysis techniques were employed over the history of LANL. Techniques include 

fluorophotometric, ion exchange, or extraction chemistry with radiometric alpha proportional counting, 

delayed neutron activation analysis (DNAA), and radiometric alpha spectroscopy (RAS). 

Table 5-13 summarizes the routine urinalysis detection levels for various periods. The results of urine 

bioassay for uranium have been entered into a database. However, no examples are currently 

available of the bioassay data format that will be provided for the claimant from this database. 

Therefore, units, codes, normalization of sample volume, and other information cannot be described 

for the dose reconstructor. The results in 1991 or before might be reported in actual values, positive 

or negative (Lawrence 1992a). There was discussion of starting the practice of establishing a 

minimum detectable activity and reporting results less than the minimum detectable activity as “0” 

after 1992. 

Table 5-13. Routine uranium urinalysis detection levels. 

Period Methoda MDA

Decision 

level

Reporting 

levelb

1949 – 1967 Fluorophotometric (DU or Unat) None listedc 50 μg/L >100 μg/L

1968 – 2/1976 Fluorophotometric (DU or Unat) 4 μg/Lc,d U

3/1976 – 1978 Fluorophotometric (DU or Unat) 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)

? dpm/Lc,e 50 dpm/L >100 dpm/L 

3/1971 – 1976 Extraction/alpha proportional counting 15 dpm/Lc

6/1976 – 1/1977 Extraction/alpha proportional counting 10 dpm/Lc

2/1977-12/1977 Extraction/alpha proportional counting 4 pCi/L

1978 – 7/1982 DNAA/delayed neutron countingfOnly 

U-235 counted

1 μg/L U-238 

0.17% enriched

7/1982 – 6/1991 DNAA/delayed neutron countingf

Only U-235 counted

4 pCi/L U-235f 

93% enriched

1982 – 6/1991 DNAA/delayed neutron countingfOnly 

U-235 counted

4 μg/L U-238 

0.17% enriched

1982 – 1992 Anion exchange/colorimetric (used to 

confirm DNAA >30 μg/L

1 μg/50 cm3

7/1991g – 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. Method listed.

b. Exceeding Reporting Levels required investigation and evaluation (Lawrence 1984).

c. Lawrence (1984). 

d. 50 μg/L considered positive indication of natural uranium material in the body (Dummer 1958).

e. Specific for U-235 and U-233, 50 dpm/24-hr sample considered positive indication of enriched uranium in the body (Dummer 

1958).

f. U-235 results should be divided by 0.0284 to calculate the 234U activity. See Section on Delayed Neutron Counting 

below for full explanation of the DNAA results including minimum reporting levels.

g. To maintain consistency during the transition from DNAA to RAS, urine results for 1991 were reported as DNAA results. For 

samples analyzed starting July 1991, summation of the U-234, U-235, and U-238 results was made. If the sum was less than 

4 pCi/L, a single result was produced with the value “L4.00 pCi/L” and labeled U235, simulating the DNAA results. In addition, 

each result was converted to μg/L and summed to obtain the total uranium mass. If the sum was less than 4.0 μg/L, a single 

result was produced with the value of “L4.00 μg/L” and labeled U238. 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. 

Delayed Neutron Counting

DNAA was the analytical technique used between 1978 and 1992. Interpretation of the results during 

this time was dependent on identifying the radioisotopic mixture to which the person might have been 
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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 results reported by the DNAA method for enriched uranium analyses are labeled U235 in 

the records, but have always included the alpha activity of 234U that is about 34 times that of 

235U. Similarly, results for depleted uranium analyses are labeled U238, 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.

By the DNAA method, the minimum reporting value was 4 μg/L for depleted uranium (238U at 0.17% 

enrichment) and 4 pCi/L for enriched uranium (93% enriched). For D-38 (depleted uranium), a 4-μg/L 

sample would contain 0.0147 pCi/L of 235U [(4μg/L * 0.0017 * 2.163 × 10-6 Ci/g * 1 × 1012 pCi/Ci)/1 × 

106 μg/g]. The enrichment of 235U assumed for depleted uranium at this period was 0.17%. A 4-pCi/L 

sample of enriched uranium (93%) would contain approximately 0.116 pCi/L of 235U [0.116 pCi/L 

divided by 0.0284 from Table 5-12 equals 4 pCi/L]. If the person’s actual urine sample contained 

0.116 pCi/L of 235U, the reported value for enriched uranium would be 4 pCi/L. If the analysis was 

designated for both enriched and depleted uranium, the D-38 value would be reported as 

approximately 32 μg/L [(0.116/0.0147) times 4 μg/L] and 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 person’s dose (Lawrence 1992b). 

Uncertainty for this method was considered to be 96% ±6% at 5 – 26 pCi/L for enriched uranium and 

93% ±8% at 4 -17 μg/L for depleted uranium.

This method was eventually deemed operationally unacceptable because of longer-than-promised 

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 5-13. More guidance on 

interpreting the results will be available when examples of the uranium database are available. 

Radiochemical, Alpha Spectroscopy (RAS Technique)

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 

alpha pulse height analysis (RAS) and the activities of 232U, 234U, 235U, and 238U were determined. 

Several problems were initially encountered with this method. No chemical or synthetic urine blanks 

were used to establish 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 

analysis indicated isotopic ratios outside the realm of probability. The computer program converted all 

negative values to zero. All 1991 data yielded negative values for 235

U, rather than the expected 

negative and positive values. In early 1992, the computer program was still not modified to correct for 

chemical blanks.

Excretion of Environmental Levels of Uranium

Natural uranium from nonoccupational intakes (primarily food and water) is 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 depleted uranium from natural uranium. 

Table 5-14 lists activity ratios for natural uranium.
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Table 5-14. Natural uranium (Tuballoy).

Isotope 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 NIOSH Phase I database USDOE Version 1.0.42.

A 1992 study (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, using 0.01 as the factor of drinking water concentration to excretion concentration 

(Little, Miller, and Guilmette 2003a,b), it is reasonable to use the urinary excretion values of 

0.00015 μg/L for elemental analyses, 0.0001 pCi/L for 234U and 238U, and essentially anything 

detected for 235U, to distinguish between natural background and potential occupational exposure for 

uranium, unless the worker’s file indicates that the excretion was from natural sources

New Mexico is known for high levels of natural uranium in the soil and ground water. However, some 

LANL workers lived in areas of particularly high natural uranium concentrations ranging from 0 to 

4 Bq/liter (108 pCi/L) in 1992 and up to 6 Bq/liter (162 pCi/L) in 2001 (Little, Miller, and Guilmette

2003b). These areas of high concentration are primarily in the Espanola area. 

Beginning in 1992, all workers participating 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 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, the values should be 

multiplied by 0.01 to calculate the concentrations of uranium isotopes to be subtracted from 

the individual’s bioassay results prior to calculating an occupational intake. The factor is based 

on consuming 1.1 L of water a day (Little, Miller, and Guilmette 2003b). The concentrations were 

found to vary widely even for individuals. The concentration of the most closely associated water 

sample should be subtracted from the bioassay sample rather than an average value for all the 

samples listed for the individual. Other variables, such as drinking habits, dietary components, and 

individual physiological differences, can influence the individual’s baseline excretion rate of 

environmental uranium. However, the use of the factor is the best approximation available to the 

dose reconstruction.

However, when other environmental data are not available, the default environmental decision levels 

should apply to the entire history of LANL.

Background excretion of uranium in feces probably varies 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

The first gross beta urine count was devised in 1947 (LASL 1947). Fission product urinalysis was the 

method used to monitor for intakes of fission products until whole-body counting was fully 

implemented in 1970. A procedure dated 1958 lists gross beta urine analysis from 90Sr/Y, 140Ba/La, 

144Ce/Pr, 89Sr, and gross fission products (Milligan et al. 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 >200 dpm/L. A similar procedure of oxalate 

coprecipitation and beta counting, effective in 1974 (Gautier 1983), lists a sensitivity of 1-2 dps/L and 

Fis s ion and Activation Product Analys is
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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 located in office 

memoranda (LASL 1959b). 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 1962, 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 5 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/1500 mL). 

(LASL 1959b)

Gas leaks were detected with filter papers and charcoal. Filter papers and charcoal were analyzed by 

gamma spectroscopy. Personnel potentially exposed were analyzed in the human spectrometer [9.5-

in. x 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 1959b). Through the late 1940s and 

as late as 1961, prior to the use of charcoal cartridges, iodine was analyzed from the paper filter 

assuming a collection efficiency of 0.1%. Radioiodine and noble gases are released from facilities 

performing 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 <1 

mCi/y) from TA-48 operations, and the OWR Facility (TA-2) is not a source of radioiodine.

Interpretation of the fission/activation product urinalysis in a way that is meaningful, as representative 

of all the possible fission products 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 

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 interpreting 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, the environmental background radiation in Los Alamos was 

as high as 5 mR/hr from fresh fission products. These radionuclides were found in urine bioassays for 

gross fission products. The background levels, which were variable, provide a complicating factor 

when using 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.
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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/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.

Strontium

Records of 90Sr urinalyses, routine or special, are very sparse. The historical compilations of 

procedures do not list a specific 90Sr urinalysis procedure. It appears that any record of 90Sr analysis 

actually indicates that LANL performed a gross beta analysis or sent a sample to an outside 

laboratory. Strontium-90 dose currently can be reconstructed only when 90Sr results are actually 

listed for an individual. No information is available on MDA for 90Sr analysis. Section 5.3 discusses 

the early HUMCO whole-body counters as being capable of counting the bremsstrahlung from 90Sr. It 

appears that, at most, several hundred bioassay analyses might have occurred during the history of 

LANL.

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 Group 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 some time 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 curies 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 currently available for those 

nuclides. Savannah River Site lists an MDA of 9.3 nCi for that period.

5.2.6

Polonium was used in initiators in quantities of 100 Ci/month or more starting in February 1945 and 

increasing to 500 Ci/month by December 1945. Prior to that time, smaller quantities were handled at 

TA-1, H and Gamma Buildings (ENSR 2002). DP East, Buildings 151, 152, 153, and later (1949) 155 

were also used to process polonium for initiators.

Polonium

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. 

Work with 210Po was of a limited scope at LANL. Therefore, missed dose for polonium should be 

assessed only if actual polonium bioassay has been 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, LASL (1977a) states, “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 
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exposure. (Exposure was controlled from raw urine assay data.)” Dummer (1958) states that the 

possibility of exposure should be investigated at greater than 10 dpm/L in a urine bioassay sample.

Hemplemann (1945) states that, because of the ease of the analysis, polonium urine bioassay was 

performed routinely and frequently on all individuals working with polonium. However, records of 

routine polonium bioassays might not have been transferred to the current LANL database and might 

not be available in individual dose records. 

Initially the urine bioassay analysis procedure could not distinguish adequately 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. 

Detection limits for routine urinalysis are listed in Table 5-15. A procedure is listed for 210Po in urine in 

Gautier (1983). MDA continues to later periods unless changed. 

Table 5-15. Routine 210Po urinalysis detection levels. 

Period MDA Recheck Tolerance

1943-1952 440 dpm/24ha

1953 50 dpm/24ha

1954 10 dpm/L 100 dpm/L 500 dpm/L

1955 0.1 pCi/L

1958 100 dpm/L 500 dpm/L

a. Source: LASL (1979). 

Because 210Po is a naturally occurring radionuclide from the 238U decay chain, 210Po exists naturally in 

urine and feces. Nothing was found in the records indicating that a study on natural excretion levels 

for persons living around LANL had been conducted. ICRP Publication 23 indicates that excretion 

levels differ between smokers and nonsmokers (ICRP 1975). The following estimated excretion 

values are provided: urine, smokers - 0.065 pCi/d, nonsmokers - 0.011 pCi/d; feces, smokers - 3.3 

pCi/d, nonsmokers - 3.2 pCi/d. 

If there are person-specific baseline values for urine or fecal excretion of 210Po, those should be used 

to subtract from later results. If not, the ICRP (1975) values above can be used.

5.2.7

LANL has always been a center for research. As such, small-scale (in terms of either the number of 

persons or activity of the source) use of various radionuclides not addressed above has occurred 

throughout the history of LANL. Little or no documentation has been found on bioassay for these 

nuclides.

Other Limited-Expos ure Radionuclides

These nuclides should be assessed only when there is an indication that the worker had a potential 

for interaction with that nuclide.

Actinium-227

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 listed were 0 to 0.6 dpm/sample. 

Potential worker encounter with 227Ac occurred during the decommissioning of TA-21-153 in 1978. 

However, all workers on the decommissioning wore full-face particulate respirators and 
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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). 

No other information is currently available.

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 is found for 14C. 

Thorium 

A procedure for a coprecipitation technique and a colorimetric final determination is listed as effective 

January 1963. The MDA for this analysis is listed as 20 μg/L. Dummer (1958) lists the MDA for the 

colorimetric procedure as 0.01 μg/L. While plutonium can coprecipitate with thorium, the colorimetric 

determination prevented any interference from the plutonium. An earlier procedure titled “The 

Determination of Thorium230 in Urine” is listed in Milligan et al. (1958). This procedure uses low 

background alpha proportional counting and lists a sensitivity of 0.05 dpm (no volume units) of 230Th. 

The procedure states that natural thorium, plutonium, some americium, curium, actinium, and 

neptunium are carried over with the precipitate.

Only the results of radiometric analysis of 44 urine samples for 230Th performed in 1958 have been 

found thus far. A sensitivity of 0.2 dpm (no volume units) is listed for these analyses. Dummer (1958) 

suggests the use of alpha proportional counting for 230Th bioassay samples. Thorium-230 is also 

referred to as ionium.

LASL (1951) discusses monitoring of thorium daughters in air. Daughter products would be produced 

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 in 

environmental emissions from waste sites at LANL. After 1980, some work with thorium-impregnated 

cloth was performed. The thorium in the cloth 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, storing ingot and oxide 

forms, which would suggest 232Th and daughters. Therefore, the dose reconstructor should use best 

judgment when selecting the isotope or mixture. Thorium-230 is the maximizing conversion from 

mass units.

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 radiometric alpha spectroscopy analysis, 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 5-16. 
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Table 5-16. 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

Protactinium-231

A procedure for a coprecipitation technique and proportional alpha counting is documented in Milligan 

et al. (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.

5.3 IN VIVO 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-ray and gamma-ray-emitting radionuclides since 1970, and possibly as 

early as 1960. There is some indication that some of the counts recorded between the beginning of 

the program in 1955 and the 1960s were performed for development of the program rather than actual 

suspected intakes. Counts during this period should be evaluated as closely as possible for validity in 

the dose reconstruction. 

5.3.1

The first whole-body counter to be used at LASL was the HUMCO I. This “human counter” became 

operational in 1955. The counter consisted of a large double cylinder with a liquid scintillation fluid 

(possibly trichloroethylene) filling 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 (1-2 MeV) and one for 137Cs + 40K 

Compton counts between 0.5 and 0.8 MeV. The result was obtained by subtracting out 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) crystal were approximately the 

same as those for the HUMCO, except the count time was significantly longer. 

Whole-Body Counters

The HUMCO II became operational in 1958. The HUMCO II was housed in a count room (SB-16) 

made of 7 in. of pre-World War II steel. The resolution was improved, but it remained a screening 

counter. 

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 200- to 2,000-

keV photons (CsI) for a qualitative assessment of a variety of fission and activation nuclides. A planar 

Hyper Pure 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. Finally, a 

HPGe (formerly a GeLi) 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. 
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The twin Phoswich detectors were replaced by twin three-detector arrays of HPGe detectors in 1999. 

The Phoswich and germanium systems were operated concurrently during the 1998-1999 period. 

Two of the six HPGe detectors are used when a thyroid count is required. 

Minimum Detectable Activity and Decision Levels

Operationally an observed signal must exceed the Decision Level (LC), formerly referred to as the 

Maximum Significant Measured Activity (MSMA), to result in the decision “detected” (decision level) 

that there is some activity in the lung, body, or other organ (Vasilik et al. 1984). The MDA, formerly 

the Minimum Detectable True Activity (MDTA), is the smallest amount of activity required to be in the 

lungs or organ so that a measurement of an individual can be expected to imply, correctly, that 

presence of activity with a predetermined degree of confidence. MDA (or MDTA) and LC (or MSMA) 

values are listed in Table 5-17 for various years of operation. The MDA and LC values for lung 

counting are summarized in Attachment 5A, Table 5A-27. The values of MDA and LC are calculated 

according to the theoretical developments of Currie and of Altshuler and Pasternack (Vasilik et al.

1984). No information is currently available on MDAs for the thyroid detector or for 131I or 125I in the 

whole-body count. 

Whole-body and lung counting are the primary methods for determining intakes of fission and 

activation products. When both whole-body (in vivo) and in vitro results are available, whole-body

count results are generally considered more sensitive and dose reconstructors should use them to 

determine intake. 

Results found in bioassay records are generally reported in nCi unless otherwise indicated. Results 

listed as NULL indicate NDA (No Detectable Activity), not MDA. Results less than the LC are marked 

as NDA in the database. The actual decision-level values and counting errors might be available on 

the report of the individual but are not available in the database until 2003.

An in vivo count spectrum is not analyzed for a fission/activation product radionuclide unless a peak 

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 correlated to MDA or critical level calculations, 

especially with the broad peaks associated with scintillation-type 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.

Cesium-137 and Other Intakes from Fallout

Most workers in the early days of whole-body counting had detectable activities of 137Cs. Most of this 

was attributed to fallout. Some workers had even higher levels of 137Cs from consumption of wild 

game. A decision level used to establish the difference between occupational and non-occupational 

sources of 137Cs and other fallout radionuclide intakes has not been discovered in the records. In lieu 

of other information, the guidance in Table 5-18 can be used.
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Table 5-17. Routine whole-body counting detection levels.

a

Period Nuclide LC nCi MDAb,c nCi

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

Co-57 2.1 4.8

Co-60 0.78 1.8

1985 to 1998 Tl-202 0.5 0.9

C-11 (based on 511 keV)f 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 6.0

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-141 2.2 4.4

Ce-144 12.1 24.2

Cr-51 6.4 12.8

Co-57 1.4 2.8

Cu-67 1.5 3.0

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 to presentg 511 keV 1.15 2.3

Be-7 4.2 8.4

Ce-141 1.35 2.7

Ce-144 6.5 13.0

Co-56 0.55 1.1

Co-58 0.5 1.0

Co-60 0.45 0.9

Cr-51 4.45 8.9

Cs-134 0.5 1.0

Cs-137 0.6 1.2

Cu-67 1.5 3.0
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Table 5-17 (Continued). Routine whole-body counting detection levels.a

1999 to presentg 

(continued)

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 41.0

Sb-124 0.45 0.9

Sc-46 0.7 1.4

Se-75 0.85 1.7

Tl-202 0.5 1.0

U-235 1.35 2.7

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. 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.

• 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 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 

non-occupational 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 1987). Those values are 

listed in Table 5-18. If no other fission or activation products are linked to the intake and the 

137Cs result is less than the values in Table 5-18, the 137Cs result can be assumed to be due to 

fallout.
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Table 5-18 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 6.0 1975 1.1

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

a. Source: NCRP (1987).

5.3.2

Lung burdens of 239Pu, 238 Pu, 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 238 Pu can be 

determined from the 241Am. Otherwise, the plutonium is determined from the U-L X-ray region. When 

241Am, 239Pu, and 238 Pu are present, corrections for the contribution of the Np-L X-rays, from the 

decay of 241Am, to the 14- to 25-keV 239Pu and 238Pu region must be considered. The Phoswich 

detectors were eventually replaced by arrays of HPGe detectors, greatly improving the energy 

resolution. Improved energy resolution permits the system to distinguish between gamma and X-ray

lines that are closer together. However, because the U-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.

Lung Burdens 

Efficiency, and therefore the sensitivity level, varies for every count due to the effects of chest wall 

thickness on the attenuation of the 17-keV X-rays and the 59.5-keV gamma ray. Therefore, the MDA 

listed with the count, when available, should be used. The MDA and LC values listed in Table 5-19 

are nominal and are based on the calibration chest wall thickness (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 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. LASL (1977b) suggests that for large individuals, MDAs should be increased 50%. 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 

Attachment 5A, Table 5A-28.

The minimum intake detectable using in vivo counting for americium, plutonium, or uranium is much 

larger than the minimum intake detectable through in vitro methods. Results or detection limits for in 

vivo methods should typically be used only to bound intakes determined from in vitro results of 

detection levels and not as the primary source of determination of intake or missed dose for these 

nuclides. 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 
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Table 5-19. MDA LC values for lung counting (nCi). 

Period Radionuclide LC MDAa

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 to present 

(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 to 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 (LASL 1977b).

c. Might be applicable to the startup of the Phoswich system in 1970, no other 

information available.

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 

g. Lung counter has 10-300 keV and 80-3000 keV ranges, so a lower sensitivity 

for certain fission and activation products can be obtained.
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more than 50 years. None of these individuals has had activity above background 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 containing 241Am in 1957. 

Based on the above discussion, in-growth of 241Am from early intakes (1944-1945) should be 

expected to be negligible. Intakes in later years might result in 241Am in-growth or might have 

contained 241Am in the original intake.

Lung counts for follow-up on members of the UPPU Club were typically longer (60 minutes rather than 

15 or 20 minutes) 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. (Rule of 

thumb is quadruple the count time half the MDA.)

Lung counting can be useful in determining actual lung burdens when high-fired oxides of plutonium 

are involved. The extreme insolubility of these oxides has been described at several DOE sites 

including LANL (Hammond, Lagerquist, and Mann 1968). These oxides typically exhibit a lower than 

expected urine excretion rate, and the total lung burden might not be adequately predicted using 

standard excretion equations. The dose reconstructor should use best judgment on forcing the data 

to fit a type S model or decreasing the absorption rate constants in the lung model. High-fired oxides 

involving 238Pu from heat source production have been encountered at LANL (Miller et al. 1999). 

Results in bioassay records should be assumed to be reported in nCi 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 decision-level 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 recent years, routine 

235U and 234Th (as 238U) have been added to the routine in vivo analysis library.

5.3.3

In August 1959, the H-6 Group acquired a probe to be used to monitor wounds contaminated with 

plutonium. This probe was capable of detecting soft plutonium X-rays. The sensitivity of this probe 

was 1 × 10-9 Ci of plutonium unshielded. This was equivalent to detection of one-tenth of a 

permissible body burden of plutonium embedded in tissue to a depth of 1 cm (LASL 1959a). In 1977, 

a new NaI detector (12 mm x 2 mm) was being evaluated. This produced an MDA of 0.07 nCi based 

on WGPu (LASL 1977c).

Wound Monitoring

Wound counting was used primarily as a tool for surgeons to locate plutonium in the wound, not as 

results used to calculate internal dose. Wound monitoring continues to be performed. In most cases, 

intake and dose will not be assessed directly from the wound count but rather from the resultant urine 

bioassay data. Follow-up studies of wounds found that plutonium does not readily migrate from the 

wound site to uptake in the majority of the incidents (Voelz and Lawrence 1991). No other information 

on instrumentation or sensitivities is available.
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5.4 INTERFERENCES AND UNCERTAINTIES

5.4.1

Until the Health Pass Ward was established in 1945, the potential existed for sample contamination. 

It is likely that a contaminated sample will show up as an obvious outlier in the dataset for a given 

worker. The use of a result from a contaminated sample could result in an overestimate of the 

workers dose; however, the sample result should be considered real if other data do not demonstrate 

the sample to be a false positive result. Some variability is expected in any set of results. See 

Section 5A.9, Attachment 5A, for an example of a valid series of bioassay results for an individual 

whose autopsy results verified plutonium intake(s). 

Contamination 

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 

no detectable activity on the same day or within a few days, external contamination can be suspected. 

Radon progeny and medical diagnostic or therapeutic procedures involving radionuclides can cause 

interference to in vivo measurements, especially for sodium iodide (NaI) detectors. However, unless 

the count was invalidated or noted as being influenced by such interferences, the results should be 

used as recorded.

5.4.2

In Vivo Counting

Uncertainties

Uncertainties for bioassay measurements might be included with the results for excreta or in vivo

measurements. For in vivo results, uncertainties were not reported in the database until 2003; 

however, uncertainties might be listed on individual reports prior to 2003. The listed uncertainties are 

typically reported as one standard deviation. 

Uncertainties associated with chest counting are reduced by use of different calibrations for different 

chest wall thicknesses and use of ultrasound to measure chest wall thickness. A 1-sigma uncertainty 

of about 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 and/or liver. The uncertainty in lung activity estimates affected by 

contributions from activity in the liver and skeleton would likely 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%. The mathematical correction 

was made for routine counts. Special counts and counts with positive results were typically corrected 

using ultrasound chest wall thickness measurements.

Based on the above discussion, the assumption provided in the Internal Dose Reconstruction 

Implementation Guidelines (NIOSH 2002) is adequate and should be used; namely, the standard 

deviation is 0.3 times the MDA or reporting level, with the exception of chest counts, for which 0.5 

times the MDA should be used. For results greater than 3 times the MDA or reporting level, the 

standard deviation can be assumed to be 0.1 times the result, based on Currie’s quantification level 

(Currie 1968). Actual tests for in vivo counts of phantoms show even smaller uncertainty, but 0.1 is 

appropriate for broad applications. If actual standard deviations or other indications of uncertainty are 

reported with a bioassay measurement result, the reported value should be used. 

In Vitro Measurements

Uncertainties used for both normal and lognormal distributions are listed in the database for plutonium 

and americium samples. The measurement uncertainty at one standard deviation and the
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biovariability are listed with the results. 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. A study of the variations found in plutonium urinary data (Moss et al. 1969) 

determined that the LASL employee results exhibited as lognormal distribution with a geometric 

standard deviation of 1.9 calculated from the ratio of the 84th percentile to the 50th percentile. 

The tritium database lists the measurement uncertainty at one standard deviation with the bioassay 

results. It is likely that the uranium database will also list the uncertainty of each measurement at one 

standard deviation.

Chelation Therapy

Chelation therapy is used only in circumstances where significant doses are expected if the material is 

allowed to follow natural bioelimination kinetics. Fewer than 35 individuals have undergone chelation 

therapy at LANL for americium or plutonium intakes. The urine bioassay results directly associated 

with the chelation therapy period are coded with a “C” in the database (see Attachment 5A, 

Section 5A.1.1). The chelation process causes the rate of bioelimination of the plutonium or 

americium to increase. Therefore, the results of urine bioassay samples taken during treatment and 

for at least 100 days following the last treatment, which could last up to 100 days, should not be used 

to determine the intake. The intake should be evaluated based on later urine bioassay results, if 

available. Other considerations might have to be applied. The dose reconstructor should thoroughly 

investigate the dose and medical records when chelation therapy is noted.

5.5 AIR CONCENTRATION IN SELECTED BUILDINGS

Average Airborne Concentration Levels

Chronic intakes, or frequent intermittent intakes, can be modeled as chronic. Very low-level, frequent, 

intermittent intakes might have occurred for the highest risk workers, but the intakes were below 

detectability at the time. For radionuclides with long residence times in the body, chronic intakes lead 

to a slow buildup of activity in the body and a concomitant increase in urinary excretion. Studies have 

shown that room air concentration is not necessarily a reliable quantitative predictor of intake because 

of the variations in respiratory conditions and particle dispersion within an area. Even breathing zone

(BZA) samples are not always an accurate predictor of the amount of intake (Whicker 2004). 

However, maximum and average airborne concentrations can be helpful in establishing boundary 

conditions for intakes. Maximum and average airborne contamination levels in buildings with high 

exposure potential are listed in Table 5-20. Airborne concentrations were not available for all years of 

operation, especially for fission products. Average concentrations are calculated from general air 

samples as simple averages or averages obtained directly from LASL reports. Blanks in the table

indicate that no information is currently available for that period.

5.5.1

The first reference to the need for respiratory protection for LASL workers is in a memo requesting 

“assault-type gas masks for testing in connection with toxic effects of X10 (plutonium) dust” 

(Oppenheimer 1944). The May 31, 1944, “Health Safety Rules” for Buildings H and V-8 specify 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 1944c). Positive 

pressure masks were being introduced into the program at that time. Dry boxes and respirators were 

a satisfactory method of handling dry material; however, individuals working with solutions 

consistently received the highest exposure despite all precautions. 

Respiratory Protection Program 
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Table 5-20. Summary of airborne concentrations from selected buildings. 

Building Nuclidea,b Year

Maximum

airborne concentrationc

dpm/m3

Average

airborne concentrationc 

dpm/m3

D Alpha 1945 674.0

D Alpha U 1945 2,564.400

H Alpha Po 1945 48.00

DP East Alpha (EU) 1945 2,458.0 68.2

DP West Alpha 1945 254.0

D Alpha (Pu) 1945 480.0 151.0

D Alpha (Pu) 1946 2,590.0 46.86

DP East Alpha (EU) 1946 147,400.0 (Po) 271.27

TU Building Alpha 1946 1,958.400 43.400

DP West Alpha 1946 2,400.0

TU Building Alpha 1947 515.000

H Alpha Po 1947 8.80

Sigma Total alpha U 1947 187.20

D U 1947 55.00

D Pu 1947 134.00

D Cold lab alpha 1947 15.0

Sigma U 1948 1,393.80 124.60

U Pu 1948 379.40 66.67

H Po 1948 52.80 4.50

D U 1948 2,564.40 21.23

D Pu 1948 860.80 11.53

D Cold lab alpha 1948 882.4 (normally 35.8) 5.675

Am Lab Am alpha 1948 6.2 1.6

Sigma Total alpha U 1948 1,393.80 124.60

DP West Alpha 1949 1,694.0

Am Lab Am alpha 1949 18.0

Press Rm Alpha 1949 198.0

TU Bldg Furnace Alpha 1949 4,095.0

Waste Treatment Lab Alpha 1949 13.8

M Alpha 1949 580.8 84.0

DP East Polonium 1949 2,344.0

U Alpha 1949 37.2

D Pu 1952 2.4

CMR Pu 1952 1.9

M U 1952 4.6

D Pu 1953 600.4 3.2

CMR Pu 1953 388.7 1.0

U Pu 1953 4.5 1.4

M U 1953 210.6 2.4

Dd U 1953 1,969.3 4.3

CMR U 1953 21.1 1.2

CMR Pu 1954 2,851.3 1.3

CMR Pu 1955 527.2 0.7

CMR Pu 1956 162.7 0.7

CMR Pu 1957 351.0 0.8

CMR Pu 1958 1,370.2 1.0

CMR Pu 1959 6,712.9 1.0

CMR Pu 1960 518.1 0.9

CMR Pu 1961 426.0 1.0

CMR Pu 1962 4,680.0 1.0

CMR Pu 1963 166.0 1.0

CMR Pu 1963 4.0 0.0
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Table 5-20 (Continued). Summary of airborne concentrations from selected buildings.

Building Nuclidea,b Year

Maximum

airborne concentrationc

dpm/m3

Average

airborne concentrationc 

dpm/m3

CMR Pu 1967 285.0 1.0

CMR Pu 1968 5,370.0 1.0

CMR Pu 1968 11.0 0.0

CMR Pu 1969 685.0 1.0

CMR Pu 1969 11.0 0.3

CMR U 1954 77.4 0.8

CMR U 1955 112.1 0.7

CMR U 1956 67.8 1.0

CMR U 1957 2,231.1 1.4

CMR U 1958 51.3 1.3

CMR U 1959 24.0 1.2

CMR U 1960 94.1 1.2

CMR U 1961 35.0 1.0

CMR U 1962 43.0 1.0

CMR U 1963 65.0 1.0

CMR U 1963 11.0 0.0

CMR U 1967 80.0 1.0

CMR U 1968 53.0 1.0

CMR U 1968 14.0 0.6

CMR U 1969 65.0 1.0

CMR U 1969 258.0 2.4

Laundry Total alpha 1948 844 450.533

Laundry Total alpha 1949 2,268 583.067

Laundry Total alpha 1950 78 42.2857

TA-21e I-131 1961 1.4E4 NA

DP West I-131 1964 302 5.5

Sigma Beta -gammaf 1947 1.82E+06

H Beta -gammaf 1947 0.0

U Beta -gammaf 1947 5.95E+05

Sigma Beta -gammaf 1948 1.86E+09 8.21E+07

H Beta -gammaf 1948 3.33E+08 7.39E+07

U Beta -gammaf 1948 1.47E+09 1.95E+08

CMR Mixed fissionf 1961 78.0 12.8

CMR Mixed fissionf 1962 11,627.0g 19.7g

CMR Mixed fissionf 1963 518.0 15.0

CMR Mixed fissionf 1964 19,256.0h 39.0h

CMR Mixed fissionf 1965 13,404.0i 22.0i

CMR Mixed fissionf 1966 93,887.0j 366.7j

CMR Mixed fissionf 1967 19,256.0 39.0

CMR Mixed fissionf 1968 13,404.0 22.0

CMR Mixed fissionf 1969 93,887.0 366.7

CMR Mixed fissionf 1970 14,163

CMR Mixed fissionf 1971 18,104

a. Guidance on adjusting for different enrichments of uranium or mixtures of plutonium is in Section 5.2. 

b. Information on specific mixtures of fission products is not readily available from LANL Guidance.

c. Results prior to January 1, 1953, were reported as cpm/L. Listed results have been converted to dpm/m3 assuming 

a nominal counting efficiency of 50%.

d. D Building demolished in 1954.

e. Dummer (1961). 

f. Mixed Fission Products (assume gross beta/gamma) tolerance is 1 × 10-7 μCi/cc (2.22 × 105 dpm/m3 ) through 

1948; MAC is 6.7 × 103 dpm/m3 from 1961 to 1969. 

g. Highest concentration of 11,627 dpm/m3 is in "uranium cell corridor - no personnel exposure.” Next highest 

concentration is 207 dpm/m3

. Average concentration would be 11.8 dpm/m3. 

h. "No personnel exposure" to highest concentration of 19,256 dpm/m3

. Next highest concentration is 2,551dpm/m3. 
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Table 5-20 (Continued). Summary of airborne concentrations from selected buildings.

i. "No personnel exposure" to highest concentration of 13,404 dpm/m3. Next highest concentration is 2,233 dpm/m3. 

j. "No personnel exposure" to highest/next highest concentrations of 93,887 and 45,997 dpm/m3

. Next highest 

concentration is 21,553 dpm/m3

. 

Respiratory protective devices were provided during the Manhattan Project by the U.S. Army 

Chemical Warfare Laboratories. The M-9 mask with either the M-11 canister for gases and 

particulates or the M-14 canister for particulates was used beginning in 1946. The efficiency of the 

canister was 1 in 100,000 particles of dioctyl phthalate (DOP; AIHA 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 wherever 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. 

The Industrial Hygiene group at LASL continued to be at the forefront of development of respiratory 

devices and programs. Respiratory protection programs are documented beginning with Dummer 

(1958). The programs include training in the use of and fit testing for air-purifying respirators and selfcontained breathing apparatus. Protection factors assumed for respiratory protection equipment are 

listed in Table 5-21. The protection factors appear to remain consistent throughout the history of the 

program within the exceptions noted. 

Table 5-21. Respiratory protection apparatus

Equipment Protection factor Remarks

Half-mask Not more than 10x MPCa b, c

Full-face mask Not more than 50x MPC (Dummer 1958)d b

Full-face mask Not more than 100x MPC (Healy 1970) (LANL 1984)e b

Full-face mask with fit field tested Not more than 1000x 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 (IDLH)

c. Half-face respirators were not permitted to be used in airborne highly toxic dust such as plutonium (LASL 1964). LANL 

(2004a) states 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 98-99% efficient.

e. Powered air-purifying respirators not more than 3,000 times with high-efficiency 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 (CAM) in the area. 

Continuous Air Monitors

CAM alarms have been historically used as an indication that room air concentration had changed 

and that respiratory protection was required. Whicker et al. (1997) found that the selection of the 

location of the CAM was critical to the reliability of the response. The typical location of CAMs at the 

ventilation exhaust point of a room was not considered optimal. Results also suggest that when a 

worker causes the release and is at or near the release point, the worker could be exposed for a 

significant period before a radioactive cloud reaches the CAM. 
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The above limitations suggest that, without other interventions, the possibility exists that workers 

could be exposed to intakes that did not trigger alarms. However, after 1970 when the bioassay 

programs were well-established, the majority of workers with the potential for monitored and 

unmonitored intakes are expected to have participated in a bioassay program. 

5.5.2

Intake parameters can be derived from airborne contamination levels for buildings with the highest 

exposure potential or highest intakes for various periods. Maximum and average airborne 

contamination levels for selected buildings are listed in Table 5-20. Incidents and intakes are listed in 

Table 5-22. Average airborne contamination levels either are derived as simple averages or are 

reported as averages listed in LASL reports. Simple averaging is assumed for LASL reports, although 

no information on the methods used to obtain these reported averages is available.

Maximizing and Bes t Es timate Intake Parameters

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 5-22; Table 5-23 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.

Table 5-24 lists maximum airborne concentrations for selected buildings during various intervals. In 

practice, individuals in areas of maximum airborne concentration would most likely have been wearing 

appropriate respiratory protection.

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 maximum 

permissible body burden 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 exposure of personnel had to be capable of detecting from 2 to 14 dpm 

plutonium in a 24-hr 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. The alpha MAC values 

used for selected buildings are listed in Table 5-24. Assume that the period for the listed MAC 

continues until changed. The MACs are dpm for total gross alpha with the primary nuclide for the 

area listed, except as noted. MACs were eventually replaced by Maximum Permissible 

Concentrations (MPCs) and Derived Air Concentrations (DACs) as controlling values. Los Alamos 

Action Levels (ALs) are also listed where applicable. It appears that at some time prior to 1964 the 

MAC values were changed to Action Levels with larger, more specific MAC values established for 

nuclides, enrichments, and mixtures. Table 5-25 includes an example of the Action Levels and MACs 

for selected buildings in 1964.

In 1948, the tolerance level for beta-gamma emitters in air was 1 × 10-7 μCi/cc (2.22 × 105 dpm/m3

). In 

December 1961, the MAC for mixed fission products was established in CMR Building, Wing 9 at 6.7 
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Table 5-22. Reported exposure incidents and results . Date Incident

1944–1946 26 workers received intakes estimated to be 6 to 80 nCi while involved in Pu 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 Pu salts in 1945 (Hempelmann and Langham 1953).

1944 Periodic overexposure of individuals working on RaLa project occurred for 6 months at start of project until “bugs” 

were worked out of remote chemical handling procedures.

1944 Highest nasal swipe for May was 11,372 cpm. (assume alpha)

1945 August - Building 52 handled large amounts of Po. Note at bottom of 52 air concentration data table stated: "Note:

0.75 c/m/L is used as 2-year tolerance value assuming 100% retention by lungs."

1945 4 persons exceeded “safe” amount of Pu in body, 1 μg.

1945 Tolerance value for Po in urine samples 1,500 cpm/day (assume 50% efficiency for counter), exceeded by only 2 

individuals. All persons working with Po were monitored.

1945 At the Water Boiler at Omega several instances of mild to moderate overexposure when gas exhaust lines leaked or 

during decontamination of active material.

1948 11/2 - worker from GMX ingested sufficient amount of RaLa to give a reading on detector.

1948 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 1949). 

1949-1951 Number of airborne contamination results above Pu 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% of swipes 

in 1949 to 15% in 1951 for DP West and 60% to 15% for Tech Area (LASL 1951).

1951 1,876 Pu urine samples below MAC and 492 Po urine samples from DP East were below tolerance. All routine 

tritium urine samples were below MAC (LASL 1951).

1951 DP West dismantle 408 and 413, high airborne plutonium exposures were kept to minimum. 

1951 DP East experimental program with Ac. No overexposure; however, adequate urine analysis is not available, air 

samples difficult to evaluate.(LASL 1951)

1951 Pajarito Warehouse contaminated by leaking Po source. TA-33, Area 6 required cleanup from Po spill.

1951 Ruptured Pu slug at GT site vault. Contamination spread to main building. 

1951 Airborne alpha activity in Room 513 DP West was consistently above MAC. Source could not be localized; however, 

cleaning attic and sealing openings into room decreased airborne levels to satisfactory (LASL 1951). MAC total 

alpha plutonium as of April 1951was 0.0044 cpm/L.

1951 Kilocurie quantities of tritium handled in CMR areas. “Health program for this work has been satisfactory.”

1950’s late Radioiodine Experiments: In late 1950s, Los Alamos biomedical researchers orally administered I-125 and -131 to

group of 19 subjects.

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.” 

8/1959 Three employees > MPL, 11 > 50% MPL, 90% of people with ≤10% MPL for Pu (0.04 μCi) of 1,325 monitored 

(LASL 1959a) 

4/1959 Radioactive gas leak at LAPRE II resulted in thyroid uptakes by at least 5 workers of I-131, I-133, and I-135 resulting 

in thyroid burdens equivalent to approximately 3 μCi I-131 (Van Dilla 1959). 

1959 Laboratory air sample 3.8 dpm/, 16% of MPC for 1-week sample, highest value 2/1959 to 7/1959 (LASL 1959).

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 d. (ICRP 1979; Miller et al. 1999), half-time for ICRP 30 model is 10,000 

d, AMAD is 0.2μ. Intakes 2,150 to 210 nCi, found to fit other Wing 9 intakes.

1972 - 1973 Demolition of Bldg. 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 February - one employee received Pu intake estimated at 1 MPBB (0.04 μCi) (LASL 1977a,d).

January - oxide of Pu caught fire, airborne 4.5 to 11.65 dpm/m3, no positive nasal swipes, no bioassay (LASL 

1977b).

H-3 urine bioassay following 1/77 incident at LAMPF < 1 μCi/L (LASL 1977a)

February 10 incident resulted in <16 nCi lung burdens for five employees, one employee received DTPA chelation 

therapy. See Table 5-23 for details (LASL 1977b). Urine sample Pu intake estimates are 1 nCi (one worker), 4 nCi 

(two workers), 9 nCi (one worker).

U-235 airborne concentrations of >1,500 dpm/m3 and >25,000 dpm/m3 for DP Bldg 4, Room 412 5/77 (LASL 1977b).

June - Tritium in urine 101 μCi/L following airborne release DP, Bldg 5.

June - Alpha contamination (up to 100,000 cpm) from damaged uranium target found on several persons at LAMPF 

switchyard (LASL 1977d).

May - TA-35 and TA-41 alpha contamination escaped from pressure vessel, caused personnel contamination (LASL 

1977d).

20,000 dpm alpha airborne (U-235 and U-238 operations).

November - Tritium bioassay (confirmed) of P-9 employee 798 μCi/L. (LASL 1977c).

1978 UF6 tube ruptured (5-10 g) spreading HF fumes and depleted U-238. (LASL 1978a).

Operator replacing leaking hose at TA-33 stack had tritium bioassay of 4 μCi/L; 449 Ci HT released (LASL 1978b).
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Table 5-22 (Continued). Reported exposure incidents and results .

Date Incident

1979 Zia employee with nasal swipe of 553/70 dpm TA-21(LASL 1980a).

1980 TA-33 tritium exposure to employees.

54-μCi/L tritium bioassay TA-33 (LASL 1980b).

153-μCi/L tritium bioassay TA-33.(LASL 1980c).

1983 Pu spill in TA-55, 10 persons, no significant exposures (LASL 1983a).

High airborne contamination results in high nasal swipes TA-55 Pu (LANL 1983b).

21 Pu 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; 1 wound above 20 nCi, numerous positive (>50 dpm) nasal wipes (LANL 1983b)

Table 5-23. Summary of February 10, 1977, plutonium incident (LASL 1977d).

Worker

Total systemic burden (increase) Nasal swipe (L/R) dpm Chest burdena Pu nCi

Pu-239 nCi Pu-238 nCi Total alpha

Technician 1 8.2 (3.3) 1.7 (0.1) 12759/13385b 7.5

Technician 2 7.6 (5.8) 0.7 (0.5) 801/1849 13.

Supervisor 2 42.9 (7.5) 3.2 (0.6) 1091/1193 6.

Technician 3 21.8 (18.0) 1.1 (0.4) 611/1193 10.

HP Surveyor 4.2 (0.3) 0.7 (0.0) 226/577 0

a. No worker had chest counts above minimum detectable activity (16 nCi)

b. Received DTPA chelation therapy

× 103 dpm/m3. In January 1964, the MAC (DP West, Room 401) for 131I was 2.9 × 104 dpm/m3. In 

January 1968, a MAC for 131I was established at 6.7 × 103 dpm/m3

. 

5.6 UNMONITORED WORKERS

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

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

arrived at LANL in 1943 there was an awareness 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). 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. 

However, in 1944, bioassay techniques consisted of swipes of both nostrils at the end of the working 

day. Monitoring results were described as whether maximum permissible concentrations were 

exceeded, not whether there was an intake. 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 was constructed. 

Even hallways and other non-restricted areas had contamination. In 1944, shoe covers worn by 

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

cpm (LASL 1944a). 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 -1946) for any site worker. 

Bioassay was provided only for those workers who were most exposed. However, because the 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 49 of 96

Table 5-24. Summary of maximum allowable concentrations (MAC) for selected areas (dpm/m3).

Year

TA-1 TA-3 TA-21

(dpm/m3) (dpm/m3) (dpm/m3)

U Bldg 

U Labs

D Bldg 

EU Labs

D Bldg 

Pu/U Labs

TU Bldg 

normal U

Sigma Bldg 

U/EU Labs

M Bldg 

EU Labs

CMR

U and EU

CMR 

Pu Labs and Waste 

Treatment/ Laundry

DP West

EU

DP West

Pu and 

Am CMR-4

DP East

U Bldg 54

DP East

Po

1945 70 1,500 Po

1947 70 400

1950 70 400 400 400

8/1950 68 68 68 68

9/1950 70; Pu, Am, Po, U 68 70 68 1,400 Po

4/1951 8.8 8 66 66 66 8.8 8.8 8.8 1,400 Po

5/1951 8.8 8.8 8.8

10/1951 66 66 8.8

4/1952 66 U

8.8 Pu

66 8.8

1961 66 235Ua

1963 9.0

1964 66 (AL)

1967 4.0

a. Wing 9.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 50 of 96

Table 5-25. MAC and Action Levels for 1964 (Meyer 1964). 

Nuclide

Action level 

(dpm/m3)

DP West

(dpm/m3)

CMR 

(dpm/m3)

Sigma (TA-3)

(dpm/m3)

Shop 13

(dpm/m3)

Natural U (sol) 66 154.0 154.0 154.0

Natural U (insol) 66 132.0 132.0 132.0

EU (sol) 66 1,100.0

EU (insol) 66 220.0

U-233 (sol) 66 1,100.0

U-233 (insol) 66 220.0

I-131 NA 2 × 104

Fission products NA 6.7 × 103

Am-241 (sol) 8.8 13.2

Am-241 (insol) 8.8 220.0

Pu-239 (sol) 8.8 4.4

Pu-239 (insol) 8.8 88.0

Cm -244 (sol) 8.8 19.8

Cm-244 (insol) 8.8 220.0

excretion of plutonium is continuous following 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 received a large 

intake of radioactive material that was unmonitored and unnoticed is less after 1946. However, the 

probability of unmonitored small intakes is larger. Periodic reports from H Division of air samples, 

contamination incidents, and “hot spots” continue to report a significant number of “over tolerance” 

occurrences throughout the 1960s. Respiratory protection was available, but except in a few 

locations, was only donned when a CAM alarmed or airborne levels approached tolerance or action 

levels. Thus, a potential existed for an intake prior to the alarm. Review of a Summary of 

Radiological Incident Reports (RIR) through June 1998 indicates, that while the number of RIRs had 

decreased significantly, the potential for unmonitored small intakes continues to exist. (LANL 1998) 

In addition, instances might have occurred, especially in the early history, where a person not 

normally assigned to radiation work was asked to participate as a substitute in a task involving 

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

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

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

As late as 1977, it is reported that releasable and non-releasable (contaminated) salvage materials 

were accessible to “scavenger personnel” who frequented these bins and possibly mixed the types of 

scrap while it was waiting for pickup (LASL 1977d). A fence was installed to prevent uncontrolled 

access when an incident resulted in contaminated scrap leaving the area on a truck. 

The formal protocol for placing workers in bioassay programs was in place in 1970 with the Health 

Physics Checklist. Prior to the Health Physics Checklist, a person was placed in a bioassay program 

at the discretion of the health physics monitor in charge of the area. While these protocols ensured 

that most workers who required monitoring were monitored, some workers might not have been 

included. In addition, there is a possibility that the checklist might not have been updated when a 

worker’s job title or responsibilities changed. 

Over the history of the Laboratory, the health physics program has always been at the forefront of 

development of instruments, analysis procedures, and guidance for radiation protection.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 51 of 96

No bioassay data, except plutonium and americium urine and fecal analysis, have been submitted to 

date. Therefore, the experiences of workers with monitored intakes cannot be used to develop a 

scenario for workers who were unmonitored. Potential intakes will be derived from removable 

contamination levels, tolerance and MAC air concentration levels, and airborne concentrations of 

significant radionuclides. 

5.6.1

Plutonium, Uranium, and Polonium

Workers Pre-1947

Until the new DP laboratory facilities were built in 1946, contamination levels were significant. Stop 

work and decontamination of the area was required when alpha contamination exceeded 10,000 cpm 

(assume ~50% counting efficiency). Decontamination was recommended between 2,000 and 

10,000 cpm, but was not required. Bioassay was not available before late 1944 and not completely 

established until 1945. If a worker had potential for plutonium exposure from 1943 through 1946, with

no bioassay during that period but plutonium bioassay during later years, dose reconstructors could

use the later plutonium bioassay with claimant-favorable assumptions to bound intakes. However, if 

the worker did not participate in a plutonium bioassay program after this period, Table 5-26 contains 

plutonium intakes that can be assigned. Although the potential of contamination in 1943 was much 

less than in the following years, an intake equal to 10% of the 1945 intakes is being assigned. 

Table 5-26. Hypothetical chronic intakes for plutonium, 1943-1946. 

Period

Total Pu alpha

Comments

Worker

dpm/day

Casual

dpm/day

8/1943-12/31/1944 9.92 0.992 1% and 0.1% of 1945 intakesa

1/1/1945 – 12/31/1945 992.0 99.2

1/1/1946 – 12/31/1946 308.8 30.88

a. Based on comparison of total plutonium material at the lab in 1943 (micrograms), 1944 

(milligrams), and 1945 (kilograms). 

For unmonitored workers potentially exposed to plutonium, uranium, or polonium from 1943 through 

1946, hypothetical intakes were constructed based on the listed parameters. Nuclides should be 

selected according to available information on work locations or job descriptions. It is expected that 

most workers would not have worked with all nuclides simultaneously. Resultant dose should be 

reconstructed as a lognormal distribution with a GSD of 5. The GSD has been increased above the 

GSD of 3 attributed to biokinetics because of the additional uncertainty attributed to air sampling.

Intakes are derived assuming a breathing rate of 2,400 m3 per working year for persons performing 

work in the area and 10% of the rate for persons casually entering the areas. 

Plutonium

A limited number of air sample results were available for 1945 and 1946. These air sample data are 

consistent with normal resuspension factors (of 1 × 10-6/cm to 1 × 10-5/cm) if surfaces were slightly 

above or below the 0.004-nCi contamination criterion. Intakes are divided into three periods because 

of the amounts of plutonium present at the Laboratory. Plutonium mixtures during these years should 

be assumed as 3%, fresh. Intakes are based on average air sample results as listed in Table 5-20 for 

Building D, which was a Pu building. No respiratory protection is assumed, although respiratory

protection was available for these years. Table 5-26 lists the hypothetical intakes for workers and 

casual encounters with plutonium 1943 through 1946.

For hypothetical acute intakes, a 4-hour exposure at the maximum air concentration for the applicable 

building can be assumed from Table 5-20. 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 52 of 96

Polonium

Average air sample data from DP East for 1945 and 1946 are used to estimate potential intakes for 

210Po. No average concentration data are available for H Building. The average values are

representative of the potential exposure if the individual was not involved in an incident. If the 

information indicates that the person might have been involved in an area of high airborne 

contamination, the maximum air concentrations listed in Table 5-20 for 1945 and 1946 would result in 

hypothetical acute intakes of 210Po for a hypothetical 4-hr period of exposure, without respiratory 

protection, of 12,000 dpm and 710,000 dpm, respectively.

For general application for unmonitored individuals the intakes listed in Table 5-27 can be assumed. 

Intakes of polonium do not exhibit a long-term excretion pattern compared to plutonium intakes; 

therefore, monitoring for these early intakes at a later time is not possible.

Table 5-27. Hypothetical intakes for polonium and uranium, 1943-1946. 

Total alpha activitya

Comments

Worker

dpm/day

Casual

dpm/day

1943-12/31/1944 74.25 7.425 Assumed from 1945 concentrations

1/1/1945 – 12/31/1945 74.25 7.425

1/1/1946 – 12/31/1946 149.16 14.92

1/1/1947 – 12/31/1949 13.95 1.40 For uranium only

a. Can be interpreted as either Po-210 or U-234 as benefits the claimant, with the exception of 1947-

1949. 

Polonium was used in initiators, utilizing the (α,n) reaction of 210Po and 9Be. In February 1945, the 

schedule for polonium delivery from Monsanto to TA-1 was increased from a few curies to 100 Ci per 

month by June and 500 Ci per month by December. At TA-1, polonium and radium were handled in 

H Building and Gamma Building. H Building was used for preparation of neutron and alpha sources 

for initiators and isotopic experiments. Workers involved in these operations were part of Group CM15. Operations in H Building were carried out between 1943 to July 1945 when operations were 

moved to the DP Site. Polonium (210Po) 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.

Absorption type F or M might be appropriate.

Uranium

Plutonium, uranium, and polonium were the most significant nuclides encountered during the 1943-

1946 period. No bioassay was available until at least 1949; therefore, a hypothetical intake rate has 

been calculated for uranium until 1949. The average 1945 and 1946 air sample data from the DP

East site, which was an enriched uranium facility as well as a polonium facility and D building, was 

used to construct the hypothetical intake for enriched uranium listed in Table 5-27. The average 

values are representative of the potential exposure if the individual was not involved in an incident. If 

the claimant interview indicated that the person might have been involved in an area of high airborne 

contamination, the maximum air concentrations listed in Table 5-20 for 1945 and 1948 would result in 

a hypothetical intake of 234U for a hypothetical 4-hour period of exposure, without respiratory 

protection, of 12,000 dpm and 670 dpm, respectively. 

Absorption type F or M is appropriate for depleted or enriched uranium unless other specific 

information is available.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 53 of 96

5.6.2

These considerations lead to the following reasonable assumptions for unmonitored workers at LANL: 

General Guidance

• 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 

depleted or enriched, although natural uranium was used extensively in conventional weapons 

testing. 

• Claimant interviews can indicate that the worker might have participated in radiological 

assignments not typically part of the normal work functions without radiological records or 

monitoring. In the early years of operation substitution of non-radiological workers on 

radiological tasks was possible. 

• Workers with potential exposure to fission or activation products prior to 1955 (possibly 1958) 

were not monitored. This includes exposures or potential exposures to 131I and 125I.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 54 of 96

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NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guidelines, OCAS-IG-002, Revision 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

Oppenheimer, J. R., 1944, “Request for gas masks,” record of telegraph transmission, Los Alamos, 

New Mexico, February 3. 

Rich, B. L., S. L. Hinnefeld, C. R. Lagerquist, W. G. Mansfield, L. H. Munson, and E. R. Wagner, 

1988, Health Physics Manual of Good Practices for Uranium Facilities, EGG-2530, UC-41, 

Idaho National Engineering Laboratory, Idaho Falls, ID, June. 

Schulte, H. F., and D. D. Meyer, 1957, Control of Health Hazards in Handling Plutonium, Results of 14 

Years Experience, LADC 3318 revised, Los Alamos Scientific Laboratory, Los Alamos, New 

Mexico.

Vasilik, D. G., and I. C. Aikin, 1983, In Vivo Assessment of Whole Body Radioisotope Burdens at the 

Los Alamos National Laboratory, LA-9858-MS, Los Alamos National Laboratory, Los Alamos, 

New Mexico.

Vasilik, D. G., I. C. Aikin, K. L. Coop, and C. J. Umbarger, C. J., 1984, In Vivo Assessment of Lung 

Burdens at the Los Alamos National Laboratory, LA-9979-MS, Los Alamos National 

Laboratory, Los Alamos, New Mexico.

Voelz, G. L., L. H. Hempelmann, J. N. P. Lawrence, and W. D. Moss, 1979, “A 32-Year Medical 

Follow-up of Manhattan Project Plutonium Workers”, Health Physics, volume 37 (October), pp. 

445-485.

Voelz, G. L., R. S. Grier, and L. H. Hempelmann, 1985, “A 37-Year Medical Follow-up of Manhattan 

Project Pu Workers”, Health Physics, volume 48, number 3, pp. 249-259.

Voelz, G. L., and J. N. P. Lawrence, 1991, “A 42-y Medical Follow-up of Manhattan Project Plutonium 

Workers”, Health Physics, volume 61, number 2, (August) pp. 181-190.

Whicker, J. J., 2004, “Relationship of Air Sample Measurements to Internal Dose: A Review,” Health 

Physics Society Midyear Topical Meeting.

Whicker, J. J., J. C. Rodgers, C. I. Fairchild, R. C. Scripsick, and R. C. Lopez, 1997, “Evaluation of 

Continuous Air Monitor Placement in a Plutonium Facility”, Health Physics, Volume 72 (May), 

pp. 734-743.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 60 of 96

Uncited References

Campbell, E. E., M. F. Milligan, W. D. Moss, and H. F. Schulte, 1972, History of the Plutonium 

Bioassay Program at the Los Alamos Scientific Laboratory, 1944-1972, LA-5008, Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico.

DOE (U.S. Department of Energy), 2000, DOE Standard: Health Physics Manual of Good Practices 

for Uranium Facilities, DOE-STD-1136-2000, Change Notice No. 3, Washington, D.C. 

Inkret, W. C., D. Lewis, G. Miller, and M. E. Schillaci, 1998, Los Alamos National Laboratory 

Radiological Dose Assessment – Uranium Internal Dosimetry and Bioassay Programs, 

LA-UR-99-837, Los Alamos National Laboratory, Los Alamos, New Mexico.

Langham, W. H., 1947, Determination of Plutonium in Human Urine, LAMS-603, Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico.

LANL (Los Alamos National Laboratory), 1998, Summary of Radiological Incident Reports through 

June 1998, ESH-12 ALARA Metrics, Los Alamos, New Mexico.

LASL (Los Alamos Scientific Laboratory), 1944, Health and Safety Report, Chemical and Metallurgical 

Division, May, 1944, LAMS-00099, Los Alamos, New Mexico.

LASL (Los Alamos Scientific Laboratory), 1977, Photodosimetry Evaluation Book, “BIBLE”, Vol. 3, 

Procedures, 1970-1977, Los Alamos, New Mexico.

LASL (Los Alamos Scientific Laboratory), 1954, Analytical Procedures of the Industrial Hygiene 

Group, 1954 Edition, Action Levels, Los Alamos, New Mexico.

LASL (Los Alamos Scientific Laboratory), 1958, Analytical Procedures of the Industrial Hygiene 

Group, 1958 Edition, Action Levels, Los Alamos, New Mexico.

LASL (Los Alamos Scientific Laboratory), 1977, Quarterly Progress Report, October – December 

1977, Group H-1, Health Physics, Los Alamos, New Mexico.

Lawrence, J. N. P., 1957, Estimation of Whole Body Dose (rem) from Tritium in Body Water, LA-2163 

Health and Safety, Los Alamos Scientific Laboratory, Los Alamos, New Mexico.

Leggett, R. W., K. F. Eckerman, V. F. Khokhryakov, K. G. Suslova, M. P. Krahenbuhl, and S. C. 

Miller, 2003, Mayak Worker Study: An Improved biokinetic model for reconstructing doses 

from internally Deposited Plutonium, Oak Ridge National Laboratory, Oak Ridge, Tennessee; 

to be published 2004, Applications of Dosimetry in Epidemiology, currently in peer review .

Schubert, J., 1951a, “Estimating Radioelement in Exposed Individuals –I, Radioelement Metabolism”, 

Nucleonics, volume 8, number 2, pp. 13-28.

Schubert, J., 1951b, “Estimating Radioelement in Exposed Individuals –II, Radiation Dosage and 

Permissible Levels”, Nucleonics, volume 8, number 3, pp. 66-78.

Schubert, J., 1951c, “Estimating Radioelement in Exposed Individuals –III, Bioassay Operations and 

Procedures”, Nucleonics, volume 8, number 4, pp. 59-67.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 61 of 96

Van Dilla, M.A., 1959, “LAPRE II Gas Leak, Radioactive Contamination of Personnel (April 22, 1959),” 

Memo to D. Meyer, H-1, Los Alamos Scientific Laboratory, Los Alamos, New Mexico

Whicker, J. J., 1993, 1993 Radiation Protection Workshop Proceedings, CAM Correlation to Fixed Air 

Sample Data at Los Alamos National Laboratory, Department of Energy Office of Health and 

Nevada Field Office.
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GLOSSARY

acceptable MDA (AMDA)

That MDA listed by nationally recognized standards as the minimum sensitivity that a 

laboratory should achieve. Actual detection levels are often lower than this AMDA.

aging

In the context of reactor fuel and mixtures of plutonium isotopes, the time since 241Am was 

separated from the plutonium mixture.

activation

The induction of radioactivity in material by irradiation with neutrons.

activity fraction

The fraction of the total activity represented by a particular radionuclide. 

BiPO4 era 

The period when plutonium samples were analyzed by the BiPO4 technique; October 1949 to 

January 1957. 

concentration guide

The average concentration of a radionuclide in air or water to which a worker can be 

continuously exposed without exceeding acceptable radiation dose standards. 

cooling

In the context of reactor fuel, the time since the fuel was removed from the reactor core. 

cupferron era

The period when plutonium samples were analyzed using the cupferron analytical technique; 

March 1944 to October 1949. 

DTPA

Chelating agents in the form of calcium salts (CaDTPA) or zinc salts (ZnDTPA) of 

diethylenetriaminepentaacetic acid.

dry box

The predecessor to the modern-day glovebox, the dry box was made of wood with ports for 

attached rubber gloves. The atmosphere often included inert gas to further contain the dust. 

High-efficiency filters on dry box exhaust and pressure differentials were used to control the 

spread of contamination to the outside. The operator could perform tasks completely separate 

from the material without exposure to the dusts, etc.

fabrication

Manufacturing

hydrodynamic test

High-explosive non-nuclear experiment to investigate hydrodynamic aspects of primary 

function up to mid to late stages of pit implosion.

implosion

A sudden inward compression and reduction in volume.
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Kiva

A remotely controlled critical assembly building associated with the Critical Experiment Facility.

nCi-years

Nanocurie-years – The 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 nCi-years can be summed to provide total nCi-years

NTA era

The period when plutonium bioassay samples were analyzed by the NTA technique; January 

1957 to April 1966. 

outfall

The discharge point of a drain, sewer, or pipe as it empties into a body of water. 

rad

unit of absorbed dose

simulated

In the context of urine sampling, collection of urine from about one-half hour before retiring to 

bed, through the sleep period, and for about one-half hour after rising for two consecutive 

nights, or other similar protocol, to simulate a 24-hour sample.

Sv

sievert -- the SI unit for dose equivalent. (1 Sv = 100 rem). 

tolerance values 

Refers to the concentration of a radionuclide in a bioassay sample, above which would 

indicate that an unacceptable intake had occurred or an unacceptable body burden existed in 

that individual. 

Tuballoy

Natural uranium
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1 

ATTACHMENT 5A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

TABLE OF CONTENTS

Section

5A.1 Scope.....................................................................................................................................

Page

66

5A.2 Codes Used in Bioassay and Internal Dose Records ............................................................. 66

5A.2.1 Los Alamos Bioassay Data Repository Codes.................................................................. 66

5A.2.2 Other Database and Report Codes .................................................................................. 68

5A.3 Locations and Typical Radionuclides ..................................................................................... 68

5A.4 Computer Code for Validating Plutonium Bioassay Samples ................................................. 69

5A.5 Summary of Computer Codes................................................................................................ 81

5A.6 Current and Historical In Vitro Bioassay Programs................................................................. 82

5A.7 Estimation of Date of Intake ................................................................................................... 83

5A.8 Routine Sampling Procedures................................................................................................ 83

5A.9 Summary of Analytical Techniques for Plutonium .................................................................. 83

5A.10 Example Bioassay Results..................................................................................................... 83

5A.11 Routine Sampling Frequency ................................................................................................. 86

5A.12 Sampling Protocol for Accidental Exposures.......................................................................... 86

5A.13 Summary of In Vitro and In Vivo Sensitivities ......................................................................... 89

5A.14 Inhalation Absorption types .................................................................................................... 96

5A.15 In Vivo Bioassay Routine Frequency...................................................................................... 96
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LIST OF TABLES

Table

5A-1 Pathway codes for current in vitro bioassay reports..............................................................

Page

66

5A-2 Assessment method codes for current bioassay summary reports ....................................... 66

5A-3 In vitro bioassay type and incident verification codes from current dose reports ................... 66

5A-4 Incident line ..........................................................................................................................67

5A-5 Example urine data header information ................................................................................ 67

5A-6 Example bioassay data line .................................................................................................. 67

5A-7 Tritium bioassay in urine (edited databases)......................................................................... 68

5A-8 Uranium bioassay in urine .................................................................................................... 68

5A-9 In vivo report codes .............................................................................................................. 69

5A-10 Current and historical locations and default respiratory absorption type................................ 70

5A-11 Zia employee access to plutonium areas.............................................................................. 82

5A-12 Historical in vitro bioassay .................................................................................................... 82

5A-13 Current in vitro bioassay performed 1997 ............................................................................. 82

5A-14 Routine sampling procedure................................................................................................. 84

5A-15 Plutonium urinalysis—summary of sensitivities and analytical techniques ............................ 84

5A-16 Example urine bioassay results ............................................................................................ 85

5A-17 Routine plutonium sampling frequency ................................................................................. 87

5A-18 Routine sampling frequency 241Am exposures ...................................................................... 89

5A-19 Sampling protocol for tritium ................................................................................................. 89

5A-20 Uranium routine sampling protocol ....................................................................................... 89

5A-21 Sampling protocol for accidental plutonium exposures ......................................................... 90

5A-22 Sampling protocol for accidental plutonium exposures ......................................................... 90

5A-23 Sampling protocol for accidental plutonium exposures ......................................................... 91

5A-24 Sampling protocol for accidental 241Am exposures................................................................ 92

5A-25 Uranium nonroutine sampling protocol ................................................................................. 92

5A-26 Summary of in vitro bioassay (except plutonium and americium) sensitivity ......................... 93

5A-27 Summary of whole-body counting detection levels ............................................................... 93

5A-28 MDA, LC values for lung counting ......................................................................................... 95

5A-29 Solubility class and absorption type assigned to uranium compounds .................................. 96
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5A.1 SCOPE

The purpose of Attachment 5A is to summarize the data available in Section 5 and to provide 

supporting information to discussions in the text.

5A.2 CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS

5A.2.1

The codes in Tables 5A-1 to 5A-6 are in text files (.txt) from the Los Alamos Bioassay Data 

Repository. These files currently contain plutonium and americium bioassay data, both current and 

historical. Eventually, they are expected to contain uranium and thorium bioassay data. These codes 

and formats might not be applicable to bioassay data other than plutonium or americium. Autopsy 

results might be included in the database for certain individuals. This would be indicated by results 

after the date of death. Results coded as lung or liver would be for a tissue sample, not the in situ

organ count. Autopsy results will not typically used for dose reconstruction. 

Los Alamos Bioassay Data Repository Codes

The statistical approach currently in use at Los Alamos is the Bayesian Method. Therefore, certain 

parameters in the electronic text files are related to Bayesian statistics and are not applicable to dose 

reconstruction. These data are noted as such in the tables. 

Table 5A-1. Pathway codes for current in vitro bioassay reports.

Code Pathway Lung class/retention timea Particle size, if applicablea

ICRP 66/68 

retention type

iys Inhalation Y (D, W) 0.2 μ AMAD (small) S (F,M)

iym Inhalation Y (D, W) 1.0 μ AMAD (medium) S (F,M)

iyl Inhalation Y (D, W) 5.0 μ AMAD (large) S (F,M)

wnd Woundb 1/3 = half time 7d,2/3 =500 days NA NA

wdt Woundb 1/3 = half time 7d,2/3 =30 days NA NA

win Woundb Prompt, complete injection NA NA

wta Woundb Absorption NA NA

a. Might not be specified in current Los Alamos Bioassay Data Repository; only pathway listed.

b. Use wound model in IMBA, if available. Otherwise, use injection model.

Table 5A-2. Assessment method codes for current bioassay summary reports. 

Code Description Code Description

IVML In vivo measurement laboratory UF3.8 Dose calculation method

UF3.5 Dose calculation method EVU15 Enriched uranium urinalysis

ID1.3e Dose calculation method EVU18 Depleted uranium

UF3.7e Dose calculation method Urine Uranium urinalysis after 1992

Table 5A-3. In vitro bioassay type and incident verification codes from current dose reports. 

Code Description Comments

U Urine Assume 24-hr sample.

C Urine affected by chelation Ignore urine sample results for first 100 days following end of therapy.

L Lung burden In vivo, however, this can indicate lung tissue from autopsy.

F Fecal Assume per sample.

V Liver burden In vivo, however, this can indicate liver tissue from autopsy.

B This code is listed in manual.pdf file but not explained in software help file;

might indicate bone from autopsy.

T Total body burden In vivo.
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6

7 of 96

Table 5A-4. Incident line. 

i and/or p 02/28/1993a i or iw or iwmb 1.e1 b 2.e0 1.00a

Description of 

the incident

“i” -Incident related – indicates 

incident description line follows

Date of incident Models beginning with i or w 

(inhalation or wound models), iw 

= models inhalation class W, 

etc.

Median of 

lognormal 

prior 

Geometric 

standard 

deviation

Prior probability that intake is nonzero 

based on Bayesian statistics – not 

required for reconstruction

Optional

“p” – Date to modify prior 

probability (Bayesian statistical 

terminology) follows – not used 

in reconstruction.

Date to modify prior 

probability and 

(multiplier).

Primary codes are i or w. Other 

codes, e.g., Xwd, inhalation 

class (w), and particle 

diameter(d) – these parameters 

should be inferred from fitting of 

data.

a. Prior probability is Bayesian Methodology

b. See Table 5A-1 

Table 5A-5. Example urine data headera information. 

Data Description

123456M IDsex – ID and gender

12/31/1999 ENDCALC – end of calculation period – used for dose calculations

3.70E-02 Bioassay unit (Bq) – factor; 3.70E-02 indicates that listed results are pCi/sample for urine and fecal, 3.70E+01 indicates nCi 

burden for chest and whole-body counts

3.70E+01 Intake unit (Bq) –factor 37 indicates that intake is expressed in nanocuries (not used for reconstruction)

1.00E-2 Dose unit (Sv) –factor .01 indicates that dose listed is in rem. (not used for reconstruction)

a. Bioassay data is preceded by the following data.

Table 5A-6. Example bioassay data line. 03/01/1993 1174 0.0035 N or Ia 0.30 u Kit number Sim24h RAS

Date of 

sample

Results in 

bioassay units 

listed above

1 standard 

deviation in 

bioassay units

– used with 

normal 

distributions

N= Not used by LANL in 

LANL dose assignment

I= Invalid

Blank = valid

C= chelation

Biovariability 

factor uncertainty 

used for 

lognormal 

distributions 

Bioassay 

typeb

-1, -2, -3 if not assigned in 

earlier years. Indicates 

separate collections

Kit type. Sim24h. Jnpl 

normalized to /24h.

Methodb,c,d,e

See note

a. See Section 5A.3 for a description of plutonium sample results validation protocol. 

b. See Table 5A-3. 

Table 5A-6 (Continued).a

Counts Background counts Calibration Factor Sigma Tb/T Sigma Additional Background Sigma Prior Background Prior Teff/Tb

a. Remainder of line contains raw data that might not be necessary to complete entries to IMBA.
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5A.2.2

The codes in Tables 5A-7 to 5A-9 are in the Safety First Data base, the In Vivo Database, and other 

records of bioassay at LANL.

Other Database and Report Codes

A significant amount of editing and corrections were made to the original results in the tritium 

database by J.N.P. Lawrence. However, both databases still exist. Data from the original database, 

rather than the edited database, are expected to be transferred to the final database. In the interim, 

results can be submitted from either database.

Table 5A-7. Tritium bioassay in urine (edited database) . 

Z number Collected Nuclide Result Unit Uncertainty Unit

Might change if employment

status changed.

Sample 

date

H-3 “0” = < MDA

or

value

Bq/ 24 hr 

sample

1 sigma (standard 

deviation)

Bq/24 hr 

sample

As an alternative, if the unedited tritium results are supplied the format will be as follows.

Name, last 

first

Employee 

number Org. Unk

Sample 

date

Sample 

type Result Units

Lab 

book Pg Initials

Analysis 

type

Energy E W 00000 AFB LA 72059 U1 4.0 UL 9264 70 OCR T3R

Unk This column has not been identified. However, contents of the database indicate that it might 

refer to a location code. This column is blank in the database for workers from Zia. 

Sample Type; U1 is the only code currently identified in the results; others might apply. 

Results can be positive or listed as L1.0 for less than 1.0. Zeros (0) should be interpreted as less 

than the current reporting limit. 

Units: UL = μCi/L

Analysis type: T3R appears to be tritium, radiometric.

Table 5A-8. Uranium bioassay in urine. [Pending receipt of examples from LANL]

[Placeholder]

The tabulated results of in vivo counts will be accompanied by a White Card for the individual. This 

card will contain a handwritten explanation of the evaluation of the count. In addition, information on 

chest wall thickness (TM) and serial number is listed. The serial number is not relevant for the dose 

reconstruction at this time. The chest wall thickness might be useful in relating the thickness of the 

person’s chest wall to a potential change in counting efficiency and, therefore, the MDA. The group 

with which the person was associated might be listed in the upper right-hand corner of the card. This 

can be helpful in determining the areas worked. Be aware that some groups worked site-wide. One 

day a machinist might be working with D-38 and the next day with Or-93. The record might also 

include a “Beige Card” that contains pertinent personal information, including chest wall thickness.

5A.3 LOCATIONS AND TYPICAL RADIONUCLIDES 

The historical and current typical locations for radionuclides are listed in Table 5A-10 (Inkret et al. 

1998c; ENSR 2002). When the compound of uranium is known, use Table 5A-29 in Section 5A-13 

rather than the absorption type listed in Table 5A-10.
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Table 5A-9. In vivo report codes. 

Field Code Description

Count type Request Request was sent by HRS-1 or equivalent to count individual. This usually 

occurs after potential intake.

Routine

Special Nonroutine count that was not initiated by incident. Example: follow-up count 

of retired employees who have had intake in past. Note in comment field 

should clarify result. Might not have been used this way in past.

Termination Termination count, performed for persons who have ever been monitored for 

specific radionuclide. Might not indicate potential exposure in that year.

Recount To confirm results of previous count or because of failed analysis.

Comments See file Check worker’s file for notes

MDTA Results below MDTA 

Recount needed Recommendation; look for recount in records.

MDA Results below MDA

Results Null When result is NDA = No detectable activity rather than MDA. Results are 

below reporting limit (probably decision level or MSMA)

5A.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 data sets.

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. 

Review of studies listing the bioassay results of the original 26 workers with intakes of Pu show 

considerable variability between samples both in the long and short term (Voelz et al. 1979). Section 

5A.9 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.

Validations Techniques PUQFUA2 (Lawrence 1978)

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.

4. LANL recognizes that unique situations can and do occur when: 

a. An individual urine result should be retained despite its being invalidated by the test 

procedures, and
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Table 5A-10. Current and historical locations and default respiratory absorption type. 

Location Start

Demolished or 

decommissioned Nuclidea Comment

Inhalation class/type

ICRP 30 ICRP 68

TA-0 None Original town site

TA-1, Original Main Technical Area 1945 1965 active

1975 

decommissioned

All See footnote See footnote

TA-1, Bldg. D, Pu chemistry and 

metallurgy

1943 1954 Pu-239, Pu-238,

U-238, DU,

Am-241, 

Po-210, 

Ba-140,

La-140

Absorption depends on matrix or 

pure

May indicate Sr-90

Y- W 

Y 

W, Y (Pu matrix)

D –W 

D 

D-W 

M-S 

M 

M, S

F-M 

F 

F-M 

TA-1, Bldg. D-2, Contaminated 

laundry

1953 Pu-239; Pu-240; Pu238; U-236; DU;

Po-210;

Ac-227;

Ra-226

Depends on the compound W - Y 

W 

D 

Y 

M

M - S 

M 

F-M;

M-S 

M

TA-1, Bldg. D-5, Sigma vault -

storage

1965 Pu-239, 

U-238

W

W

M

W

TA-1, Bldg. ML Bldg., Medical 

laboratory

Cm

Am

Processing W

W

M

M

TA-1, Bldg. C, Uranium machining 1943 1964 Uranium W M

TA-1, Bldg. G, Uranium and graphite 

sigma pile

1959 Uranium 

Ra-226

W

W

M

M

TA-1, Bldg. H, and Gamma Bldg. 1945 1957, 1959 Po-210

MFP

Initiators

Cs-137 contamination incident 

occurred

D

Various 

F - M

Variable 

TA-1, Bldg. HT, Heat treatment and 

machining

1965 Unat and EU W M

TA-1, Bldg. HT Barrel House, 

Storage

1964 Pu-239,

U-238

D-W-Y

W

F-M – S

M

TA-1, Bldg. M, Processing and 

recovery EU

EU Processing, metallurgy and 

recovery

W M

TA-1, Bldg. M-1, Machining U-238 W M

TA-1, Bldg. O, 1956 Radium

Radon

Radon cooked off sources on a 

hot plate, Ra /RaBe Calibration 

Sources

TA-1, Bldg. Q, 1959 Radium

Radon

A spill occurred 

Ra Calibration Sources

TA-1, Sigma Bldg, 1965 Unat, EU, 

Th

Casting, machining, powder 

metallurgy

W

W-Y

M

M-S

TA-1, Bldg. TU, Machining Tuballoy 1964 Unat W M

TA-1, Bldg. TU-1, Recovery of EU 1964 EU Furnace for burning rags W M

TA-1, Bldg. V, Machine Shop 1943 1959 Uranium Unusual assignments W M
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Table 5A-10 (Continued). Current and historical locations and default respiratory absorption type.

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-1, Bldg. W, Van de Graaff 

accelerator

Uranium

Po-210

Tritium

Th-228 “Mesiothorium”

W 

D 

D 

--

M, 

F M

FSR-2 

M – S

TA-1, Bldg. X, Cyclotron Be, uranium, lithium, 

tritium, strontium

targets. 

Zn-65

Targets had induced beta activity. Various Various 

TA-1, Bldg. Y, Physics Laboratory Tritium,

Uranium

F 

M

TA-1, Bldg. 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 U fuel W 

D 

D 

NG

M 

SR-1,F

F

SR-D 

TA-2, Clementine 1946 1952 Unat, Pu,

I-131, I-125,

Rb-88, Cs-137,

Xe-131, Ar-41

Ruptured Pu fuel rod, U reflectors W

D 

D 

M

SR-1,F

F

SR-D

TA-2, Omega West Reactor (OWR) 1956 1995 U-235

I-131, I-125,

Rb-88, Cs-137,

Xe-131, Ar-41

Cr-51, 

Na-24

Tc-99m

Enriched U fuel

I-125 production Loop Schedule 

– at times operated “around the 

clock”

M

SR-1,F

F 

SR-D 

F,M,S

F 

F,M

TA-3, South Mesa Site, Technical 

Facilities

1953 All Pu processing See footnote See footnote

TA-3, Bldg. 29, CMR – Chemical 

and Metallurgical Research 

(SM-29)

1951 Pu-238 Wing 9 handled irradiated U and 

Pu In hot cells. Small quantities 

of uranium and plutonium, mixed 

fission products including iodines, 

Pu-238

Y S–M

TA-3, Bldg. FE-19 Plutonium W M

TA-3, Bldg. 34, Cryogenics Tritium 3,000 Ci HTO released in 1979 D SR-2

TA-3, Bldg. 16, Van de Graaff 

accelerator

Tritium 800 Ci HTO released in 1977 D SR-2 

TA-3, Bldg. 35, Press building U-235 W M
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 72 of 96

Table 5A-10 (Continued). Current and historical locations and default respiratory absorption type.

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-3, Bldg. 39, Tech Shop DU W M

TA-3 SM-40, Physics All Incident contaminated large 

portion of building with Po-210 

through ventilation (late 1950s or 

early 1960s)

See footnote See footnote

TA-3, Bldg. 66, Sigma Complex, 

metallurgy and fabrication

EU, DU powders, 

Pu

Pu processing W M 

TA-3, Bldg. 102, Tech Shops Uranium W M

Pu Y S

TA-3, Bldg. 141, Rolling Mill DU W M

Pu Y S

TA-3 Chemistry and Metallurgy 

Bldg. (CMR), Wings 3,5,7

H-3 HTO, HT D SR-2, SR-D

TA-3 Chemistry and Metallurgy 

Bldg. (CMR), Wing 9

1961 Cs-137 Potential for low-level chronic 

intake in hot cell work

D F

MFP including I-131 D SR-1

F

Pu-238, -239, -240 0.1-10μ AMAD, oxide, nitrate, 

fluoride and metal. Oxide is most 

common.

Usually Y, can 

be W

S -M 

TA-3, Ion Beam Facility (IBFa), 

SM-16

I-125 Iodide, labeled organics D SR-1, F

H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

D SR-2(HTO, 

OBT), 

SR-0 (HT)

P-32 Labeled organics, phosphates W F-M

TA-3 Tritium Instrument Calibration 

Facility, SM-40

H-3 HTO, HT D SR-2, SR-0 

TA-3-184, Occupational Health Pu Y S-M

TA-3-216 Weapons Test Support Pu Y S-M

TA-3-700, 

SM-700

Acid Neutralization and Pump Bldg.

Pu Y S-M

TA-4, Alpha Site 1956 DU Firing site until 1956, Materials 

disposal site C.

W M

TA-6, Two-Mile Mesa Site 1944 1950 DU Detonator manufacturing W M

TA-8, Nondestructive testing 1984 Pu-239; Pu-238

U-235; DU

Gun firing site. W-Y

D-W-Y

M – S

Co-60; Ir-192 W M

Cs-137 Y S

D F
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 73 of 96

Table 5A-10 (Continued). Current and historical locations and default respiratory absorption type.

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-10, Bldg. CMR-10, Bayo Canyon 

Site 

1944 1950 Sr-90, RaLa radiochemistry D or Y F or S

DU, Unat W W

La-140 D or W F or M

Ba-140 D F

TA-11, K Site 1947 Ra-226-Be 20-MeV betatron

TA-15, Electron accelerator 1962 Present Pu-239 PHERMEX W M

DU W M

H-3 D SR-2

TA-16, S Site, Weapons 

Engineering Tritium Facility

(WETF) 

1989 Present Pu-239 Explosive casting and machining W M

DU W M

H-3 D SR-2

TA-16 WETF, Bldg 205 Present H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

D SR-2 (OBT, 

HTO)

SR-0 (HT)

TA-18 U-235; U-233

Pu-239; Pu-240;

MFP; 

I-131;

polonium

0.1-10: AMAD, oxide, nitrate, 

fluoride and metal. Oxide is most 

common.

Ruptured Po source, 1953

W 

Usually Y, can 

be W

D-Y 

D 

D

M 

M 

F-S 

SR-1, F

F

TA-18, Pajarito Laboratory, Rover 

reactor, criticality experiments

1946 1973 MFP Betatron used from 1951 to 1954, 

Enriched uranium metal sphere 

1952

Pu core added 1 year later. 

1954 unreflected, delta phase Pu.

TA-19, East Gate Laboratory 1962 None

TA-21, DP-West, Plutonium facility 11/45 WGPu Y M-S

TA-21, CMR, Heat sources Pu-238 Accident with glovebox breached 

1971

Y M-S

TA-21, Bldg. 2 and 3, Wet 

Chemistry

1982 Pu 1958 accident , separated 

phases in Pu process tank, 

unshielded tank

Y M-S

TA-21, Bldg. 4 and 5, Dry chemistry 1981 Pu Y M-S

TA-21, Bldg. 12, Filter building 1975 Pu Ac contaminated Y M-S

TA-21, Bldg. 3, Oxalate precipitation 

operations

Pu-239

Pu-238

U-235

Y 

Y 

W

M-S 

M-S 

M

TA-21, Bldg. 4 1945 1948 EU hydride

TA-21, Bldg. 4 1960 Pu-239 Hot cell examine irradiated Pu 

and EU fuel elements

Y M-S

TA-21 1965 Pu-238 and

Pu-239

Y M-S
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 74 of 96

Table 5A-10 (Continued). Current and historical locations and default respiratory absorption type.

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-21, Bldg. 5, Pu fabrication Limited use in 1975 Pu-239, Pu-238 Fire contaminated exhaust filter, 

1959

Y M-S 

TA-21, Bldg. 150, Pu fuels 

development, heat sources 

development

1963 Pu-238, Pu-239 Sealed capillary broke, 2800 x 

MPC

10/1970

Y M-S

TA-21, Bldg. 210, Pu research Plutonium Y M–S

TA-21, Bldg. DP-East EU W M

TA-21, Bldg. 155, Tritium Systems 

Test Assembly (TSTA), deuterium 

and tritium fuels.

1984 1990 HT and HTO >10 billion Ci. Equipment failure 

-- H-3, 13.8 Ci released

D SR-2, SR-0

TA-21, Bldg. 151, 152, Experimental 

program

1945 Po-210 Produced initiators D,W F-M

Ac-227 Produced initiators Y S

TA-21, Bldg. 155 1949 1984 Po-210 Produced initiators D,W F-M

Ac-227 Produced initiators Y S

TA-21, Bldg. 153 1945 In service until 1970-

1973

Po-210 Produced initiators D,W F-M

Ac-227 Produced initiators Y S

TA-21 DP East TSTA Tritium Test 

Assembly Facility, Bldg 155 and 

the Salt Laboratory, Bldg 209 

H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

D SR-2, SR-0 

(HT)

TA-21, Liquid Waste Reprocessing, 

Bldg 35 and 257

Late 1940s 1986 All Pu and transuranic liquid wastes See footnote See footnote

TA-21 DP West Pu-238, -239, -240 0.1-10μ AMAD, oxide, nitrate, 

fluoride and metal. Oxide is most 

common.

Usually Y, can 

be W

TA-22, TD Site DU Detonator development W M

TA-23 NU Site 1945 1950 Unknown Firing site

TA-24 T Site 1944 DU Facilities transfer to TA-16 W M

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 W 

D

M 

SR-2

TA-27 Gamma Site (Far Point) 1945 1947 Pu-239

DU

Thorium

Pu gun assembly Usually Y, can 

be W

W

M-S 

M 

M-S

TA-28 Magazine A 1979 Present DU Firing site W M

TA-29 Magazine B 1957 DU Explosives storage area W M

TA-30 through TA-32 unknown

TA-33, HP Site, High Pressure 

Tritium Laboratory, Bldg 86

1950s 1980s late H-3 Tritium oxide (HTO), tritium gas 

(HT)

D SR-2, SR-0

TA-35, LAMPRE 1955 1967 MFP, Molten Pu fuel 

Sr-90, F,S

Co-60 M,S

VFP

MAP
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 75 of 96

Table 5A-10 (Continued). Current and historical locations and default respiratory absorption type.

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-35, LAPRE I, LAPRE II test 

reactors

1955 1960 MFP Highly enriched U fuel

Sr-90 F,S

Co-60 M,S

VFP

MAP

TA-35, Target Fabrication Facility 

(TFF), TSL-213

H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

D SR-2, SR-0

TA-35, Ten Site, CMR-10 1950 1963 La-140, Sr-90 contamination suspected 

(F,S)

D –W F-M

Ba-140 D F

TA-35 1981 U-235 General site W M

DU W M

Np-237 M

Pu M-S

Po F-M

TA-35, Laser Fusion Research 1974 unknown

TA-36, Kappa Site 1950 Present DU W M

TA-37, Magazine Area C DU W M

TA-39, Ancho Canyon Site 1955 1960 Unat, DU

Thorium

Firing points W M 

TA-40, Detonator Firing (DF) Site H-3 D SR-2, SR-0

TA-41, W Site, Weapons Group WX H-3 

Pu 

U 

Am

Engineering of nuclear 

components

Fabrication of test materials

D 

Y-D 

W 

W-Y

SR-2, SR-0 

F-S 

M 

M-S

TA-41, Ice House, Bldg 4 H-3 Tritium oxide (HTO), tritium gas 

(HT)

SR-2, SR-0 

TA-42, Incinerator Site 1970 All Reduced low-level Pucontaminated wasteSee footnote See footnote

TA-43 HRLa Health Research 

Laboratory 

1953 1970 I-125 Iodide, labeled organics D SR-1, F

H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

D SR-2, SR-0

C-14 Labeled DNA precursors, D SR-2

P-32 Labeled organics, phosphates W F,M

TA-43 1953 1970 All See footnote See footnote

TA-45, Radioactive Liquid Waste 

Treatment Plant, WD Site

1951 1964- operations 

ended

All Removed Pu before discharging 

effluents 

See footnote See footnote

1960 1963 MFP

1967– decom.

TA-46, WA Site 1950 1974 Rover Batteries

TA-46, WA Site 1976 1980s U-235, 238 U isotope separation W M

Th M-S
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Table 5A-10 (Continued). Current and historical locations and default respiratory absorption type.

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-48, Radiochemistry Site 1950s present All Actinide chemistry and hot cell 

isotope production

See footnote See footnote

MAP

MFP

TA-48, Nuclear Chemistry 1950s Present Se-75 Spallation product, seen in hot 

chemistry on targets

M or F

H-3 Tritium oxide (HTO), tritium gas 

(HT)

D SR-2, SR-0 

Cd-109 CL-/NO3

- mixture loaded in 

SnPO4 resin. Cd phosphate is 

most probable material of intake, 

very soluble, 1: AMAD

D F 

I-131 Fission product chemistry F

TA-49, Frijoles Mesa Site 1960 1961 H-3

Pu 

U

D

Y-D 

W

SR-2, SR-0

F-S 

M

TA-50 Waste Management Site 1963 Pu-238, -239, -240

All 

0.1-10: AMAD, oxide, nitrate, 

fluoride and metal. Oxide is most 

common.

Usually Y, can 

be W

S 

TA-51, Environmental Research 

Facility

1962 Present Co-60

Sr-90

Animal exposure facility

Presently environmental impact 

research.

Y, possibly W

D, Y

S, possibly M

F, S

TA-52, Reactor Development Site Late 1960s 1970 U-238,

Pu-238

H-3 

VFP 

Kr, Xe

High-temperature, gas-cooled, 

graphite reactor, UHTREX (Ultrahigh Temperature Reactor 

Experiment)

W

Y 

D 

M

M-S 

SR-2, SR-0 

TA-53, Los Alamos Neutron Science 

Center, largest accelerator facility, 

LAMPF 

1972 C-11, N-13, O-15, 

Ar-41

I-131

Short-lived air activation

Medical isotope production

SR-2

SR-D 

F

Induced activity in U 

targets, corrosion 

products

M,S

TA-54 Waste Storage Facility Pu-238, -239 -240

All 

0.1-10: AMAD, oxide, nitrate, 

fluoride and metal. Oxide is most 

common.

Usually Y, can 

be W

M,S

TA-55 Plutonium Facility (PF-4) 1969 H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

D SR-2, F, SR-0 

HT

TA-55 Plutonium Facility (PF-4) 1969 Pu-238, -239, -240 0.1-10: AMAD, oxide, nitrate, 

fluoride and metal. Oxide is most 

common.

Usually Y, can 

be W

M, S
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 77 of 96

Table 5A-10 (Continued). Current and historical locations and default respiratory absorption type.

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

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 

Y S 

C-11 Byproduct at LANSCE, seen in 

workers during beam cycle

D SR-2, unless 

CO; SR-1

N-13 511 keV during beam cycle D F

a. Intakes of labeled compounds do not follow the default ICRP 68 models. (ICRP 1994b) 

b. See Table 5A-3. 

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); Ra-226 (Type S)

d. MFP – mixed fission products; Cs-137

e. MAP – mixed activation products; C-11, N-13, O-15, Ar-41, Be-7, Na-22, Na-24, Co-58, Co-57, Mn-54, Mn-52, V-48

f. Delta phase plutonium = Pu-240 = 4.5%; also used sphere of 20.1% Pu-240 and 98.1% U-233

g. VFP – Volatile Fission Products
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 78 of 96

b. An individual urine result should be invalidated despite its being validated by the test 

procedures.

5. 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.

2. Select the four closest positive samples, ±1 year. 

a. Find the average of the four samples

b. Find the average of the standard deviation

3. 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. 

RF = [I/(E + I)]0.74

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

PS = RF × U2S

where: 

CL = Calculated urine level

UR = result

PS = pseudo standard deviation

U2S = standard deviation of the result
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 79 of 96

5. 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.

6. 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 precede 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 < 0.075 

pCi, the sample is rejected as too small.

NTA-ZnS or PHA Eras

1. If there at least two samples > 0.075 pCi in the set and the standard deviation is ≤ 0, the 

sample is rejected as too small.

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 exceeding 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. 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.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 80 of 96

4. If there are not four retained samples greater than LEAST, the sample being tested is kept by 

default.

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

and

σc = σe[(De- E1)/(Dl – E1)]0.74

where:

Ue = earlier dated urine result

Ul = later dated urine result

σe = standard deviation of Ue

σl = standard deviation of Ul

De = integer date of Ue

Dl = integer date of Ul

E1 = estimated integer date of intake

Uc = calculated urine result expected from Ue and 

σc = calculated standard deviation of Uc, and

(1.282)Z1-α = standard normal variable for (1-α) one-sided confidence interval; α = 0.1

(Uc - Ul) > (1.282)Z1-α [(σc

2

/nc) + (σl

2

/nl )]1/2

Hypothesis

1. Does the calculated urine result Uc exceed the measured result Ul? 

2. 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 the 

equations developed in the late 1980s or early 1990s, which take into account very-long-term 

elimination 20 or more years after uptake. Again, these are discussed fully in the description 

of PUQFUA4.)

3. Where nc = nl = 1, we conclude that Uc does exceed Ul at the chosen level of significance, and 

sample Ue on date De is rejected as invalid. 
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5A.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 of the code 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.

Internal monitoring of individuals with a potential for exposure to plutonium was controlled using a 

computer code. 

The Z1YRPUU code provided a method of marking 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 5A-11 lists the areas and exempt areas for access.

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 various 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

7. Other accident if next year’s urines show obvious increases
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 82 of 96

Table 5A-11. Zia employee access to plutonium areas. 

Area

Urine sample 

within 425 

days of entry Exempt

Job requiring respiratory protection X

Modifications or repairs on dry boxes or other highly contaminated X

Replacement of Pu-contaminated filters at all sites. X

Janitorial (long-term) work in Pu operation areas X

Long-term operations (weeks) in areas of low levels of Pu 

contamination (>1,000 dpm-60 cm2 and <10,000 dpm-60 cm2)

X

Decontamination of Pu spills with more than >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 days) whenever sizable quantities of Pu 

(grams of Pu-238 or kg 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 days) in areas of CMR Building, Ten Site, 

TA-50, TA-55, TA-54, TA-18, TA-48 or TA-21, where there is little 

Pu contamination (<1,000 dpm-60 cm2)

X

Jobs in other minimum exposure potential areas when respiratory 

protection is not required and possibility of Pu contamination is 

minimal.

X

5A.6 CURRENT AND HISTORICAL IN VITRO BIOASSAY PROGRAMS

Historically the In Vitro Bioassay Program included the nuclides listed in Table 5A-12. The current in 

vitro bioassay program includes the nuclides and techniques listed in Table 5A-13. The in vitro

bioassay samples have been taken at LANL throughout the history of the bioassay program. 

Sensitivities vary with technique and period. 

Table 5A-12. 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 5A-13. Current in vitro bioassay performed 1997 (Inkret et al. 1998b).

Material Analytical technique Number monitored

Tritium Liquid scintillation 139

Uranium α-spectroscopy 66

Plutonium α-spectroscopy 1,467

Plutonium TIMS 408

Americium α-spectroscopy 75

Strontium Liquid scintillation (gross beta counting – LSC sent to contract laboratory) 4
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 83 of 96

5A.7 ESTIMATION OF DATE OF INTAKE

The following guidance from Lawrence (1978) may be relevant to estimating the date of plutonium 

intake, based on routine and special sampling protocol, when no intake date is recorded or to 

understand LANL’s rationale for selecting 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 determining the intake date 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 

later sample. 

3. The earliest potential accident date, if several occur between the paired samples. 

4. One-half day before initial sample if potential accident is recorded on date of initial sample. 

5. The earliest potential accident date before the initial sample if any potential accidents occurred 

before initial sample. 

6. Fifteen days before the initial sample if no potential accidents occurred before initial sample.

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.

5A.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 5A-14 lists the 

sampling protocols.

5A.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 5A-15.

5A.10 EXAMPLE BIOASSAY RESULTS 

Table 5A-16 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 of the nasal swipes greater than 50 dpm were prior to 

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 are 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. 
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 84 of 96

Table 5A-14. Routine sampling procedure. 

Years Nuclide Protocol

1944 Pu Collected on 24-hr urine sample in clean areas following decontamination. 

Working in Pu areas operations – weekly, daily nasal swabs. Blood counts 

every 6 weeks (LASL 1944b).

1945 – 1952 Pu Collected 24-hr urine sample away from working environment to ensure 

contamination-free urine. 

1953 – 1957 Pu Metal kit for collection of last voiding of day and first in morning for 2 consecutive 

days. 

1958 – 1967 Pu Four 1-pt bottle disposable kit. More closely approached volume of true 24-hr 

urine sample. Reduced potential contamination.

1944 Po Collected on 24-hr urine sample in clean areas following decontamination. 

Working in Po areas operations – weekly, daily nasal swabs. Blood counts 

every 6 weeks (LASL 1944b).

1945 –1952 Po Collected 24-hr urine sample away from working environment to ensure 

contamination-free urine. 

1970 – 1975 Tritium >150 μCi/L – 1 per day; 150-75 μCi/L – every 2 weeks; 74 -24 μCi/L – 1 per 

week (Healy 1970).

1975 – present Tritium Routine – every 2 wks; or 2 hr after expected exposure then if <1- no more 

samples, 1-10 μCi/L – 1 w/in next month, 10-100 μCi/L – weekly samples, >100 

μCi/L –daily samples.

1944 –– 1967 Uranium Weekly urinalysis (LASL 1944b).

1967 – present Uranium Annual, including drinking water sample since 1992.

Table 5A-15. Plutonium urinalysis—summary of sensitivities and analytical techniques. 

Period Urine analysis method

1944 – 1952 Pu extracted by cupferron chloroform, gas flow counter (1944), Simpson alpha counter 

efficiency = 50%, background = 1 cpm to 0.1 cpm (1945+) 

Could not differentiate Po and Pu. 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, Pu plated on 

stainless-steel disk. NTA counting method, 1,000-min exposure to emulsion, background 

0.007 dpm.

1963 – 1965 Ion exchange/ Pu separated on anion-exchange resin. Electroplated for NTA counting, 

84% ±14% recovery

1966 ZnS counter

1967 – 1971 Either ZnS or alpha pulse height analysis RAS (PHA) permit measurement of Pu-238;

background= 0.003 ± 0.003 cpm 

1971 to present All alpha PHA, computerized spectrometry 

9/18/80 Began Pu-242 analysis

1982 – 1986 Coprecipitation (alkaline earth PO4 or oxalate) 

Alpha spectrometry (60,000-second count time) or rapid Alpha Phosphor Scintillation

Counting (3,600 second)

1997 Alpha spectroscopy, Based on Class Y, 1 μm, Pu-239 (Inkret et al. 1999a) 

1997 to 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. 1999a).
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 85 of 96

Table 5A-16. Example urine bioassay results. 

Date (L/R nasal count)

Average room 

air conc. dpm/m3 Dpm/24 h

8/9/46 (189/320 dpm 7/29/46) 6 – 188 1.2

9/19/46 (149/19 9/5/46) 4.7

12/18/46 (57/68 12/10/46) 1.7

4/18/47 (164/106 dpm 12/30/46); (102/61 

dpm 1/21/47); (91/135 dpm 4/1/47)

11 – 98 0.7

5/23/47 0.7

6/26/47 0.7

7/30/47 (144/40 dpm 7/7/47) 0.0

8/27/47 1.0

10/2/47 1.5

11/7/47 (120/78 dpm 10/3/47) 0.8

12/8/47 4.0

1/13/48 24 – 69 1.0

2/13/48 (0/59 dpm 2/10/48) 0.0

3/19/48 0.0

4/22/48 (86/3 dpm 4/26/48) 2.0

6/23/48 (244/72 dpm 6/10/48) 3.7

7/22/48 (72/1 dpm 7/2/48) 0.0

8/19/48 (65/0 dpm 8/2/48) 0.8

9/20/48 3.0

10/26/48 2.0

11/22/48 2.5

12/21/48 (50/38 dpm 12/1/48) 0.0

1/24/49 19 – 72 (Moved to uranium work)

1/31/49 2.0

7/14/49 1.2

2/8/50 0.8

9/1/50 0.1

2/28/51 0.8

9/4/51 0.7

5/19/52 0.3

12/14/52 0.0

9/4/53 0.0

6/4/54 Off uranium

6/18/54 0.0

6/8/55 (Moved back to plutonium)

8/1/55 3 1.2

8/12/55 (22/28 dpm 8/9/55) 0.6

8/19/55 0.0

9/30/55 0.7

11/14/55 0.7

12/27/55 0.6

2/9/56 3 0.0

4/5/56 0.7

4/30/56 0.0

6/8/56 0.4

7/20/56 0.6

8/23/56 0.5

9/25/56 0.0

10/24/56 0.0
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Table 5A-16 (Continued). Example urine bioassay results.

Date (L/R nasal count)

Average room 

air conc. dpm/m3 Dpm/24 h

11/23/56 0.0

12/17/56 0.1

1/31/57 4 0.23

2/28/57 0.68

4/12/57 0.22

5/14/57 0.12

6/14/57 0.11

7/15/57 0.03

8/19/57 0.00

9/20/57 0.20

10/22/57 0.21

10/31/57 0.39

11/14/57 0.51

1/10/58 4 0.00

2/21/58 0.65

3/25/58 0.51

5/7/58 0.025

6/19/58 0.55

7/30/58 0.49

9/15/58 0.79

11/28/58 0.47

A body burden of plutonium was potentially maintained by this individual since 1946. The bioassay 

results varied as listed. 

Average room concentrations for each year are listed. When more than one room was involved, the 

range of average concentrations is listed. The 1946-1949 period involved plutonium nitrate, plutonium 

oxalate, and plutonium fluorination. The 1955-1958 period involved primary liquid-liquid extraction of 

plutonium under greatly improved exposure conditions.

5A.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 5A-17 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 

currently available before Lawrence (1967). Table 5A-18 lists routine sampling frequencies for 241Am 

exposures; routine samples were not performed for 241Am prior to 1998. Routine and postexposure 

protocols are combined in Table 5A-19 for tritium. Table 5A-20 lists frequencies for uranium. Routine 

work or frequent entry in an area with beta/gamma emitting radionuclides currently requires annual 

whole-body count. (LANL 2004b)

5A.12 SAMPLING PROTOCOL FOR ACCIDENTAL EXPOSURES

Awareness of the sampling protocol associated with exposures provides guidance for associating 

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 prior to Lawrence 

(1967). It should be assumed that the protocol continued until the year of the next dated reference 

unless otherwise stated. Tables 5A-21 to 5A-25 list protocols for accidental exposures.
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Table 5A-17. Routine plutonium sampling frequency.

Frequency 1944 (Kolodney 1946) 1967 (Lawrence 1967) 1973 (Lawrence 1973)

1998 (Inkret et al. 1998a,c;

LANL 2004b)

Monthly Great exposure

Quarterly Moderate exposure 1. Persons working with ≥ 10 g 

Pu-239 in chemical or 

metallurgical operations, inside 

or outside glovebox 

2. Persons with 50% body burden. 

(MPBB= 0.04 μCi Pu-239)

1. Persons working w/ > 10g Pu239 or > 0.04g Pu-238 

(~0.6-0.7 Ci of either) inside 

or outside glovebox, or 

2. Persons w/ body burden 

>20,000 pCi

Semiannual 

(biannual)

Slight exposure 1. Persons working with <10g 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 w/ Pu but

 < 10 g Pu-239 or < 0.04 g 

Pu-238 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 (0.7Ci) of 

Pu-238, analyzed by RAS 

2. Performing chemical or 

metallurgical operations or 

maintenance on systems 

containing ≥10g 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 Pu 

3. Other persons with casual 

encounters with plutonium who 

regularly work in areas where 

Pu 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 Pu-239 or 0.08 mg of 

Pu-238 (~1.3 mCi of either)

1. Routine work with operations of 

<10g (0.6Ci) of Pu-239 or Pu242 (0.6-0.8 Ci) 

2. Performing maintenance on 

systems containing ≥10g 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. TRU glovebox, bag outs, etc.

6. ESH-1 RCTs who frequently 

enter work areas 

7. All personnel with confirmed, 

measurable intakes of Pu-238 

and/or Pu-239.
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Table 5A-17 (Continued). Routine plutonium sampling frequency.

1944 (Kolodney 1946) 1967 (Lawrence 1967) 1973 (Lawrence 1973)

1998 (Inkret et al. 1998a,c;

LANL 2004b)

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 Pu 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 Pu 

areas, but never sampled

Workers who submitted routine 

samples

UPPU Club

(see Section 

5.1)

Complete bioassay and physical 

examination at 5-year intervals. 

Bioassay continues after termination 

of employment.

Complete bioassay and physical 

examination at 5-year intervals. 

Bioassay continues after termination 

of employment.

Complete bioassay and physical 

examination at 5-year intervals. 

Bioassay continues after termination 

of employment.

a. Samples analyzed by TIMS – thermal ionization mass spectroscopy.

b. Samples analyzed by RAS – radiometric, alpha-spectroscopy.

c. The frequencies above do not apply to working with sealed sources in TA-15 where bioassay is as needed.
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Table 5A-18. Routine sampling frequency 241Am exposures. 

Frequency 1998 (Inkret et al. 1998; LANL 2004b)

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 – alpha-spectroscopy. 

Table 5A-19. Sampling protocol for tritium. 

Period Program Protocol

1950 – 1970 Biweekly for significant quantities or more often if exposure 

confirmed, removed from tritium work at 124 μCi/L.

1970 – 1975 >150 μCi/L – 1 per day; 150-75 μCi/L – every 2 weeks; 

74 -24 μCi/L – 1 per week (Healy 1970)

1975 – 1998 Routine – every 2 wks; or 2 hours after expected exposure 

then if <1 μCi/L - no more samples, 1-10 μCi/L – 1 w/in next 

month, 10.1-100 μCi/L – weekly samples, >100 μCi/L –daily 

samples (Inkret et al. 1998b)

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 metal tritide or 0.1 Ci of 

organic tritium

Every 2 weeks

Table 5A-20. Uranium routine sampling protocol. 

Period Protocol

Pre-1983 Biweekly samples collected.

Collected on Fridays, minimum annual

8/1983-6/1993 Spot sample collected on Monday mornings before entering work area

7/1993 Last voiding on Sunday night and second on Monday morning, first voiding.

1998 Spot samples every 2 weeks/persons performing hands-on work/ potential for 100 mrem; 

machining operations, polishing operations, foundry work, chemistry operations in which >10g 

of U, work with oxidized metal, >100g of bulk powder.a

a. Source: Inkret et al. (1998c). 

5A.13 SUMMARY OF IN VITRO AND IN VIVO SENSITIVITIES

Table 5A-26 lists sensitivity data for in vitro bioassay.

Tables 5A-27 and 5A-28 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 (LASL 1977b).
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Table 5A-21. Sampling protocol for accidental plutonium exposures (Lawrence 1967).

Severity class Description Sampling protocol

PA (Prompt Action) Most serious accidents

1. Injection detectable by plutonium wound monitor 

(~0.01 μg) 

2. Chemical burns from plutonium solutions 

3. Facial contamination > 20,000 cpm 

4. Nose swipes > 500 dpm

1. Evening following accident 

2. 10 days later 

3. 1 month later 

4. 14-monthly thereafter.

DA (Delayed Action) 1. No known equipment failure, 10x MPC for 1 

week or 50x MPC for 1 day 

2. Skin contamination in excess of limits 

3. Superficially contaminated cuts that are positive 

on surface monitoring only. 

4. Equipment failure causes exposure for 

indeterminate period.

1. 1 month 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)

1. Small wounds occurring in major plutonium 

areas (DP-West, Ten Site, CMR Bldg) 

2. Exposures without respirators at 10x MPC for 

period less than 1 week. 

3. Nose counts > 50 dpm 

4. Skin contamination > 500 cpm/60 cm2 

1. 1 urine sample collected 1 month after 

intake. 

Automatic Rescheduling 1. Detect unexpected high exposure and verify its 

existence 

2. Detect contaminated urine sample 

3. Previous sample is 1 dpm/sample

1. Request another sample 

2. Persons routinely excreting plutonium 

are exempt from special sampling but 

are maintained on routine sampling

Table 5A-22. Sampling protocol for accidental plutonium exposures (Lawrence 1973).

Severity class Description Sampling protocol

PA (Prompt Action) Most serious accidents

1. Injection detectable by plutonium wound monitor 

>2000 pCi (dl 600 pCi Pu-239 and 200 pCi Pu238) 

2. Chemical burns from plutonium solutions 

3. Facial contamination > 40,000 dpm 

4. Nose swipes > 500 dpm

1. Evening following accident 

2. 10 days later 

3. 1 month later 

4. 12 monthly thereafter.

Fecal samples might be collected 

DA (Delayed Action) 1. No respiratory protection, 10x MPC for 1 week 

or 50x MPC for 1 day 

2. Skin contamination after decontamination in 

excess of limits 

3. Superficially contaminated cuts that are positive 

(500 dpm) on surface monitoring only. 

1. Fecal collected on day 2 for inhalation 

2. 1 month after accident fecal and urine

3. 11 more urine samples at monthly 

intervals. Possible fecal 

Potential accident entry is made in PUQFUA

NRS (Nonroutinely 

Sampled)

1. Small wounds occurring in major plutonium 

areas (DP-West, Ten Site, CMR Bldg) no 

activity detected. < 500 dpm

2. Exposures without respirators at 10x MPC for a 

period less than 1 week. 

3. Nose counts >50 dpm 

4. Skin contamination >1,000 dpm/ 60 cm2 

5. Potential accident date is noted in PUQFUA and 

no sample has ever been submitted or routine is 

not scheduled for at least 3 months.

1. 1 urine sample collected 1 month after 

intake. 

2. Fecal samples for types 2 and 3.

3. Memorandum sent to H-1; potential 

accident entry in PUQFUA might be 

required

Automatic Rescheduling 1. Detect unexpected high exposure and verify its 

existence 

2. Detect contaminated urine sample 

3. Previous sample is 1 dpm/sample

1. Request another sample 

2. Persons routinely excreting plutonium 

are exempt from special sampling but 

are maintained on routine sampling
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 91 of 96

Table 5A-23. Sampling protocol for accidental plutonium exposures (Inkret et al. 1998c).

Severity class Description Sampling protocol

PA (Prompt Action) Most serious accidents 

1. Injection detectable by plutonium wound 

monitor > 0.2 nCi 

2. Chemical burns from plutonium solutions, skin 

contamination > 500 dpm, α by 60 cm2 probe 

or >0.2 nCi in wound area 

3. Facial contamination > 10,000 dpm α 

4. Nasal > 100 dpm 

5. Airborne > 200 DAC-h, without respiratory 

protection.

6. Skin after decontamination is > 1,000 dpm α 

by 60 cm2 probe

7. As requested 

Pu-239, Pu-240

1. Urine samples days 1, 3, 5 after 

intake

2. Fecal samples days 1, 3, 5–optional

3. In vivo chest count days 3, 5–optional

Pu-238

1. Urine samples on days, 1, 3, 5, 8, 16, 

30, 60, 120, 240 after intake

2. Fecal samples on days 1, 3, 5–

optional

3. In vivo chest count days on 3, 5–

optional

Urine analyzed by RAS and TIMS

DA (Delayed Action) EDE > 100 mrem; CDE >1 rem any tissue or organ

1. Present in room when CAM alarms

2. Other individuals on Prompt Action (PA)

3. Positive wound count

4. Chemical burns; after decon skin >100 dpm

5. Facial contamination, before decon, >1,000 

dpm

6. Nasal swipes; 1 nostril >15 dpm, both sum = 

35 dpm α

7. Average ≥ 40 DAC-hr, without respiratory

8. Skin contaminate >100 dpm after 

decontamination.

9. Wounds contaminated with >250 dpm (by α 

probe)

10. Request by manager, group leader, or medical

Typically 1 urine, by RAS and TIMS, if Pu239, 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 ESH-12 Dose 

Assessment.

Collection for duration of chelation 

therapy, up to 60-100 days after therapy.

Prompt Action schedule followed by 

monthly urine samplesa

. 

Might include blood and fecal.

Or as directed.

Follow-up Any positive routine urine As directed.

a. Samples taken for up to 100 days after termination of the therapy should not be used in the calculation of intake.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 92 of 96

Table 5A-24. Sampling protocol for accidental 241Am exposures (Inkret et al. 1998d).

Severity class Description Sampling protocol

Pa (Prompt Action) Most serious accidents 

1. Injection detectable by Am wound monitor > 0.2

nCi 

2. Chemical burns from Am-bearing solutions, skin 

contamination > 500 dpm, α by 60 cm2 probe or 

> 0.2 nCi in wound area 

3. Facial contamination > 10,000 dpm α 

4. Nasal > 100 dpm 

5. Airborne > 200 DAC-h, without respiratory 

protection.

6. Skin after decontamination is > 1,000 dpm α by 

60-cm2 probe

7. As requested 

1. Urine samples days 1,3,5, 

8,16,30,60,120,240 after intake

2. Fecal samples days 1,3,5 – 

optional

3. In vivo chest count days 3,5 – 

optional

DA (Delayed Action) EDE >100 mrem; CDE >1 rem any tissue or organ

1. Present in room when CAM alarms

2. Other individuals on Prompt Action (PA)

3. Positive wound count

4. Chemical burns; after decon skin > 100 dpm

5. Facial contamination, before decon, > 1,000 

dpm

Nasal swipes; 1 nostril >15 dpm, both sum= 35 

dpm α

6. Average ≥ 40 DAC-hr, without respiratory

7. Skin contaminated > 100 dpm after 

decontamination.

8. Wounds contaminated with > 250 dpm (by α 

probe)

9. 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.

Prompt Action schedule followed by 

monthly urine samples.

Might include blood and fecal

or as directed.

Table 5A-25. Uranium nonroutine sampling protocol. 

Years Nuclide Protocol

Pending input from LANL

1998 Uranium Days 1, 4, 8 following possible incidents [high airborne, high alpha skin contamination 

with >10,000 dpm or nasal swipes over 100 dpm (summed)].a

a. Source: Inkret et al. (1998a).
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Table 5A-26. Summary of in vitro bioassay (except plutonium and americium) sensitivity. 

Radionuclide Period Typea MDA Reporting level

Plutonium U/F See Table 5-5

Americium U/F See Table 5-8

H-3 1950 -- 1968 U 5.18E4 Bq/24 hc (1 μCi/L)

1969 d – 1987 U 1.04E3 Bq/24 h (0.02 μCi/L) c 5.18E4 Bq/24 h (1 μCi/L)

1988 – 1998 U 5.18E2 Bq/24 h (0.01 μCi/L) 5.18E3 Bq/24 h (0.1 μCi/L)

1999 – present U 2.59E2 Bq/24 h(0.005 μCi/L)

Uranium See Table 5-13

Isotopic uraniume See Table 5-13

Fission product 1950–1970 U 50 to 100 dpm/L

Sr-90 No information availableb.

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.

Th-230 as thorium 1958–1963 U 0.01 μg/

1963–? - 

present

U 

20 μg/L

Pa-231 1985 U 0.88 dpm/24 hour sample

a. U = urine; F = fecal.

b. Currently sent to an offsite laboratory.

c. Source: Gautier (1983).

Table 5A-27. Summary of whole-body counting 

detection levels (nCi).

a

Period Nuclide LC MDAb,c 

1955-1958d Cs-137 8

Sr-90e 30

1958-1970 d Cs-137 4

Sr-90e 30

1970-1984 Be-7 0.9 1.8

Cs-134 0.9 1.8

Cs-137 0.9 2.1

Co-57 2.1 4.8

Co-60 0.78 1.8

1984 to 1998 Tl-202 0.5 0.9

C-11 (based on 511 keV)f 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 6.0

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
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 94 of 96

Table 5A-27 (Continued). Summary of whole-body 

counting detection levels (nCi)a

Period Nuclide LC MDAb,c 

1984 to 1998 

(continued)

Sb-124 0.4 0.8

Ce-141 2.2 4.4

Ce-144 12.1 24.2

Cr-51 6.4 12.8

Co-57 1.4 2.8

Cu-67 1.5 3.0

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

1998 to presentg 511 keV 1.15 2.3

Be-7 4.2 8.4

Ce-141 1.35 2.7

Ce-144 6.5 13.0

Co-56 0.55 1.1

Co-58 0.5 1.0

Co-60 0.45 0.9

Cr-51 4.45 8.9

Cs-134 0.5 1.0

Cs-137 0.6 1.2

Cu-67 1.5 3.0

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 41.0

Sb-124 0.45 0.9

Sc-46 0.7 1.4

Se-75 0.85 1.7

Tl-202 0.5 1.0

U-235 1.35 2.7

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 given radionuclide does not necessarily mean 

that the radionuclide was frequently encountered. If smaller MDAs 

are listed in the results for a given count, use those.

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- × 4–in. NaI detector, which had the same 

sensitivities with an extended count time.

e. By bremsstrahlung. 

f. 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 1022 keV.

g. Lower sensitivities might be available using the lung counter for 

certain nuclides if lung counting is appropriate to the dose 

reconstruction.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 95 of 96

Table 5A-28. MDA, LC values for lung counting. 

Period Radionuclide LC MDAa

1977 – 60 min count special UPPU group

(LASL 1977b) 

Am-241 0.3

Pu-238 10

Pu-239 21

1980

(Ennis 2003)b

Am-241 0.155 0.31

Pu-238 11 22

Pu-239 24 48

1984 

(Vasilik et al. 1984)c Am-241 0.16 0.32

Pu-238 14 28

Pu-239 30 60

1998d to present 

(Ennis 2003)

Am-241 0.1 0.2

Am-243 0.1 0.2

Pu-238 10 20.0

Pu-239 31 62.0

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 to 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; MDA should be increased by at least 50% for very large 

individuals (LASL 1977b).

b. Assume chest wall thickness of 2.5 cm. 

c. MDA = LC * 2 

d. Lung counter has 10-300 keV and 80-3,000 keV ranges, therefore a lower sensitivity for certain 

fission and activation products can be obtained.
Effective Date: 12/21/2004 Revision No. 00 Document No. ORAUT-TKBS-0010-5 Page 96 of 96

5A.14 INHALATION ABSORPTION TYPES 

Table 5A-29. Solubility class and absorption type assigned to uranium compounds (Rich et al. 1988). 

Class D/Type F Class W/Type M Class Y/Type S

Uranium hexafluoride UF6 Uranium tetrafluoride UF4 Uranium aluminide UAlx

Uranyl fluoride UO2F2 Uranium oxide U3O8 Uranium carbide UC2

Uranyl nitrate UO2 (NO3)2 Uranium dioxide UO2 Uranium-zirconium alloy UZr

Uranyl acetate UO2(C2H3O2)2 Uranium tetroxide UO4 High-fired uranium dioxide UO2

Uranyl chloride UO2Cl2 Ammonium diuranate (NH4)2+ U2O7

Uranyl sulfate UO2SO4

Uranium trioxide UO3

5A.15 IN VIVO BIOASSAY ROUTINE FREQUENCY

Table 5A-30 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. 

Table 5A-30 In vivo bioassay routine frequency

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 < 10g of Pu-239 or -242 or 0.04g of Pu-238. 

TRU glove box/fume hood operations

TRU hot jobs

TRU bag outs

Glovebox operations

Fume hood operations

Hot jobs

Bag outs

Maintenance operations on process systems, and 

RCT duties.

Pu Semiannually Special request

Uranium Annual Weekly access to areas where depleted uranium is machined or polished, casting 

and cleaning crucibles outside dry boxes, chemical operations, including 

purification and recovery. Handling > 100g metal or bulk powder outside 

gloveboxes, or operations with uranium-hexafluoride in uncontained systems.

MFP Semiannually Special request

MAP/MFP Annual Works in high contamination areas and or works in airborne radioactivity area.

Th Annual Works with thorium