ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 42

Document Title:

Argonne National Laboratory – EastOccupational Internal Dose

Document Number: ORAUT-TKBS-0036-5 

Revision: 00

Effective Date: 03/01/2006

Type of Document: TBD

Supersedes: None

Subject Experts: Janet Johnson, Roger B. Falk, and Craig A. Little

Document Owner

Approval: Signature on File Approval Date: 03/01/2006

Norman D. Rohrig, TBD Team Leader

Approval: Signature on File Approval Date: 03/01/2006

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 03/01/2006

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 03/01/2006

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/01/2006

Stuart L. Hinnefeld, Health Science Administrator

 New Total Rewrite Revision Page Change

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

REVISION, DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 2 of 42

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

03/01/2006 00 New technical basis document for Argonne National Laboratory-East 

occupational internal dose. First approved issue. Initiated by 

Norman Rohrig and Robert Meyer.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 3 of 42

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

5.1 Introduction ..............................................................................................................................7

5.2 Summary of Activities and Processes ......................................................................................8

5.2.1 Site History and Facility Description..............................................................................8

5.2.2 Nuclear Materials on Site..............................................................................................9

5.3 Bioassay Programs..................................................................................................................9

5.3.1 Radionuclides of Concern.............................................................................................9

5.3.1.1 Solubility Classes.......................................................................................... 10

5.3.1.2 Route of Intake ............................................................................................. 10

5.3.1.3 Particle Size.................................................................................................. 11

5.3.2 Current Bioassay Program .........................................................................................11

5.3.2.1 Scheduling of Bioassay................................................................................. 11

5.3.2.2 Specific Radionuclide Bioassay Methods...................................................... 12

5.3.2.2.1 Tritium Urine Bioassay ................................................................. 12

 5.3.2.2.2 Cobalt-60 and Other Activation Products ..................................... 12

5.3.2.2.3 Strontium ..................................................................................... 12

5.3.2.2.4 Cesium ........................................................................................ 13

5.3.2.2.5 Iodine........................................................................................... 13

5.3.2.2.6 Plutonium and Americium Compounds ........................................ 13

5.3.2.2.7 Uranium ....................................................................................... 14

5.3.2.3 Bioassay Schedule ........................................................................................ 14

5.3.3 Bioassay Before 1995 ................................................................................................16

5.3.3.1 Bioassay from 1946 to 1972 .......................................................................... 16

5.3.3.2 Urine and Fecal Analysis from 1973 to 1995.................................................. 19

5.3.3.3 In Vivo Bioassay ............................................................................................ 21

5.3.4 Estimated Decision Levels and Minimum Detectable Amounts or Concentrations ...... 22

5.3.4.1 Minimum Decision Levels for 1946 to 1985................................................... 22

5.3.4.2 Estimated Minimum Detectable Activities for 1985-2001............................... 25

5.3.4.3 Current (2004-2005) Minimum Detectable Amounts or Concentrations......... 26

5.3.4.3.1 Unseparated Actinides in Urine ................................................... 27

5.3.4.3.2 Gamma-Emitting Isotopes in Bioassay Samples.......................... 27

5.3.4.3.3 Beta Liquid Scintillation Counting of Urine Samples..................... 28

5.3.4.3.4 Separated Actinides in Bioassay Samples ................................... 28

5.3.4.3.5 Strontium-90 in Urine ................................................................... 29

5.3.5 Estimation of Missed Dose for Unmonitored Individuals ............................................. 29

5.4 Air Sampling...........................................................................................................................29

5.5 Bioassay Reports................................................................................................................... 30

References ........................................................................................................................................36

Glossary ............................................................................................................................................39
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LIST OF TABLES

TABLE TITLE PAGE

5-1 Radionuclides of concern for internal dosimetry ..................................................................... 10

5-2 MDAs for activation product bioassay .................................................................................... 12

5-3 MDAs for iodine isotopes ....................................................................................................... 13

5-4 Scheduling of bioassay samples ............................................................................................ 15

5-5 Required bioassays by exposure ........................................................................................... 15

5-6 Analysis codes .......................................................................................................................15

5-7 MPLs, 1952 (adapted from Myers et al. 1952) and after 1959................................................ 17

5-8 Summary of bioassay analysis methods, 1970 to 1989.......................................................... 20

5-9 Summary of Detection Limits for Urine Bioassay Based on Reported Less Than 

Values....................................................................................................................................22

5-10 Summary of Reported Less Than Values for Fecal Bioassay................................................. 24

5-11 Estimated Decision Levels for urine bioassay by Year ........................................................... 25

5-12 Recommended Decision levels for Fecal Bioassay by Year................................................... 26

5-13 Estimated Bioassay MDAs for 1986 to 2000a ......................................................................... 26

5-14 2004-2005 Practical Dc’s and MDCs for unseparated actinides in urine ................................. 27

5-15 Dc’s and MDCs for gamma-emitting isotopes in urine and fecal samples ............................... 28

5-16 MDAs for activation product bioassay .................................................................................... 28

5-17 Dc’s and MDCs for beta liquid scintillation counting of urine samples ..................................... 28

5-18 Dc’s and MDAs for alpha spectrometry of separated actinides in urine and fecal 

samples ................................................................................................................................. 29

LIST OF FIGURES

FIGURE TITLE PAGE

5-1 Early bioassay data card........................................................................................................ 30

5-2 Early bioassay card with supplemental count data ................................................................. 31

5-3 Bioassay card from the 1960s................................................................................................ 32

5-4 Request for Bioassay Analysis card ....................................................................................... 33

5-5 Bioassay report from the mid-1970s....................................................................................... 34

5-6 Bioassay data report starting about 1986............................................................................... 35

5-7 Bioassay report with partial data ............................................................................................ 35
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AGHCF Alpha Gamma Hot Cell Facility

AMAD activity median aerodynamic diameter

ANL-E Argonne National Laboratory-East

APS Advanced Photon Source

ATLAS Argonne Tandem Linear Accelerator System

ATSR Argonne Thermal Source Reactor

BRC Body Radioactivity Counter

CEDE committed effective dose equivalent (50-year)

cm centimeter

CP-1 Chicago Pile 1

CP-2 Chicago Pile 2

CP-3 Chicago Pile 3

CP-3 ́ Chicago Pile 3 Prime

CP-5 Chicago Pile 5

d day

D&D decontamination and decommissioning

DL detection limit

DOE U.S. Department of Energy

dpm disintegrations per minute

EBR-II Experimental Breeder Reactor No. 2 

EBWR Experimental Boiling Water Reactor

EEOICPA Energy Employees Occupational Illness Compensation Program Act

hr hour

HVEM High-Voltage Electron Microscope

ICRP International Commission on Radiological Protection

in. inch

IPNS Intense Pulsed Neutron Source

IVEM Intermediate-Voltage Electron Microscope

keV kilo electron volt, 1,000 electron volts

kW kilowatt

L liter

LINAC linear accelerator

MARS Microcosm for Acid Rain Studies

MDA minimum detectable activity

MDC minimum detectable concentration

MeV mega electron volt, 1 million electron volts

min minute

mL milliliter

MPC maximum permissible concentration
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 6 of 42

MPL maximum permissible limit

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NRTS National Reactor Testing Station

SNM special nuclear materials

TBD technical basis document

U.S.C. United States Code

WBC whole-body counter

ZGS Zero Gradient Synchrotron

ZPR-I Zero Power Reactor No. 1 

ZPR-II Zero Power Reactor No. 2

ZPR-IV Zero Power Reactor No. 4

ZPR-IV ́ Zero Power Reactor No. 4 prime

ZPR-IX Zero Power Reactor No. 9

ZPR-V Zero Power Reactor No. 5

ZPR-VI Zero Power Reactor No. 6

ZPR-VII Zero Power Reactor No. 7

μCi microcurie

μg microgram

μm micrometer

§ section
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5.1 INTRODUCTION

Technical basis documents (TBDs) 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 [DOE] facility” as defined in the Energy Employees 

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the [probability of causation] guidelines established under subsection 

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

dose reconstruction regulation) define “performance of duty” for DOE employees with a covered 

cancer or restrict the “duty” to nuclear weapons work.

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

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

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

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

The purpose of this TBD is to provide information to assist in the evaluation of occupational internal 

radiation dose associated with operations at Argonne National Laboratory-East (ANL-E) from 1946 to 

the present. The information includes discussion of monitored and unmonitored worker exposure, as 

well as missed dose, using the methodology in Internal Dose Reconstruction implementation 

Guideline (NIOSH 2002b). This TBD provides guidance for assigning radionuclide intakes based on 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 8 of 42

individual dosimetry records and for situations where individual personnel monitoring data is 

unavailable or inadequate. 

Argonne National Laboratory was established on July 1, 1946 and this TBD covers activities since 

that date. Work done at ANL-E was a continuation of that done by the Metallurgical Laboratory (Met 

Lab) of the University of Chicago beginning in 1941. The Met Lab is classified as an Atomic Weapons 

Employer under EEOICPA. Job locations for workers who transferred from the Met Lab did not 

change until land and buildings were acquired for ANL-E. 

5.2 SUMMARY OF ACTIVITIES AND PROCESSES

The ANL-E site is described in detail in the Technical Basis Document for the Argonne National 

Laboratory East (ANL-E) – Site Description (ORAUT 2006a). The following paragraphs summarize 

that information to the extent necessary as background for the internal dosimetry programs that 

existed over time at ANL-E. 

5.2.1

The ANL-E was established as the first national laboratory on July 1, 1946; the University of Chicago 

has operated ANL-E since its creation. The research that ANL-E carried out in the early years as a 

national laboratory began under the university’s Metallurgical Laboratory, which built the first nuclear 

reactor, Chicago Pile 1 (CP-1), under the West Stands of the university’s Stagg Field. Before 1946, 

three reactors (CP-1, CP-2, and CP-3) were built, the latter two 25 miles southwest of Chicago at the 

Palos Park Site A in the Argonne Woods. The 7.9-hectare (19-acre) Site A became known as 

Argonne Laboratory in 1943. The CP-1 reactor was rebuilt at Site A and renamed CP-2. The natural

uranium fuel for the CP-3 reactor was replaced with enriched uranium during its period of operation, 

and the reactor was renamed CP-3’. The CP-2 and CP-3’ research reactors were operated until 1954 

(Wescott and O’Rourke 2001). The present site of ANL-E is about 5 miles west of Site A (Wescott 

and O’Rourke 2001). This site was acquired in 1947, and was called Site D (D for Dupage County). 

Table 2-1 in ORAUT-TKBS 0036-2, Technical Basis Document for the Argonne National Laboratory 

East (ANL-E) provides a summary of ANL-E buildings and operations (ORAUT 2006a). 

Site History and Facility Description

Military projects were of top priority (Wescott and O’Rourke 2001) during the early years at ANL-E, but 

the mission has encompassed research on power reactor design and development including breeder 

and boiling water reactors. Reactors for producing plutonium have been studied at ANL-E since its 

inception. Reactors were also used at ANL-E for basic research. The CP-5 reactor, which was the 

first research reactor at ANL-E Site D, went on line in 1954 and operated until 1979. Research and 

training for commercial power plants occurred at CP-5. The Zero Power Reactor No. 6 (ZPR-VI) and 

ZPR-IX were constructed in Building 315 in the 1960s to relieve the workload of the ZPR-III at the 

National Reactor Testing Station (NRTS) in Idaho. Argonne had a demonstration reactor during the 

1950s called the Argonaut (Argonne Nuclear Assembly for University Training). In 1962, the 

Juggernaut reactor was built in Building 335 based on the design of the Argonaut reactor. The Janus 

reactor, brought on line in 1964 and shut down in 1992, was built specifically for biological research 

into the effects of ionizing radiation.

During the 1960s, high-energy physics research requiring particle accelerators was one of the most 

important areas of investigation at ANL-E. Before the 1960s, the laboratory had four accelerators. 

The 60-Inch Cyclotron built in 1951 in Building 211 was used for nuclear solid-state physics, 

chemistry, isotope production, and biological studies. Two Van de Graaff accelerators in Buildings 

208 and 203 were used in basic research similar to that conducted with the Cyclotron. A Dynamitron 

accelerator in Building 203 was used in basic physics experiments. The first linear accelerator 
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(LINAC) was installed in 1962 in Building 211. The most elaborate accelerators are those associated 

with the Argonne Tandem Linear Accelerator System (ATLAS), which was the world’s first 

superconducting heavy-ion accelerator. The Zero Gradient Synchrotron (ZGS) operated from 1963 to 

1979. Together with bubble chambers, the ZGS put ANL-E at the forefront of subatomic structure and 

neutrino research in the 1960s and 1970s. The Intense Pulsed Neutron Source (IPNS) went on line in 

1981 and continues to be used in materials research. 

Basic research in the 1950s at ANL-E largely involved reactor projects but included research into 

other areas of radiation science including waste management, radiation sterilization, and the basic 

properties of the atom. Containment facilities allowed research involving hazardous radioactive 

materials to be performed safely. 

Biological studies on the effects of radiation on animals, plants, and humans were initiated in the 

1950s. Most of these studies focused on occupational or medical exposures to radium. The Biology 

Division focused on whole-body experiments using gamma irradiation facilities and the Janus reactor 

during the 1960s to early 1980s. Following that time, the focus shifted more to studies on the cellular 

and molecular level. 

Other activities at ANL-E have included research on superconductors and fuel fabrication and 

production. Environmental studies became an important mission beginning in the late 1960s. 

Decontamination and decommissioning (D&D) of ANL-E facilities has been ongoing since the 1950s. 

ORAUT (2006a) provides detailed information on these activities.

Over the course of its operations, ANL-E has measured radionuclide effluents released to the 

environment. ORAUT (2006a) summarizes the air and water effluent releases as well as incidents 

involving the release of radioactive materials or exposure to workers. ORAUT (2006b) details 

potential occupational environmental doses, which are not part of this internal dosimetry TBD. 

5.2.2

Descriptions of nuclear materials present at the ANLE Site can be found in ORAUT-TKBS-0036-2 

(ORAUT 2006a).

Nuclear Materials on Site

5.3 BIOASSAY PROGRAMS

Urine and fecal analyses were conducted at ANL-E for various radionuclides, as was whole-body 

counting (WBC) to evaluate radionuclide intakes. The ANL-E bioassay program is described in a 

collection of unrelated documents from 1946 to 1995, the most comprehensive of which is Technical 

Basis Document for Internal Dosimetry (Bertelli 2003). This document was first developed in 1995 

based on a similar Hanford document. A section for uranium dosimetry based on ANL-W procedures 

was added later in Revision 11

5.3.1

. Bertelli (2003) can be assumed to represent the bioassay program 

from 1995 to 2005.

The bioassay programs evolved over time. Table 5-1 lists the radionuclides of concern for internal 

dosimetry (Bertelli 2003) and the type of bioassay used to assess worker intakes. 

Radionuclides of Concern

 

1

 The title page of the version of Bertelli (2003) reviewed for this analysis indicates Revision 00, but the 

document is Revision 1 as noted in the Acknowledgments.
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Table 5-1. Radionuclides of concern for internal dosimetry (Bertelli 2003)

Primary Nuclides of Concern Type of Analysis

3

H Urine

Corrosion Products

(

58Co, 60Co, 54Mn and 59Fe)

In vivo measurements (whole body counting)

90Sr Urine and feces

In vivo 137Cs measurements as an indicator of 90Sr.

125I, 131I In vivo measurements (whole body counter or 

thyroid counter using NaI or phoswich detector)

134Cs, 137Cs In vivo measurements (whole body counting)

Urine

Uranium Urine

239Pu, 240Pu, and 241Am Urine and fecal sampling

In vivo measurements (lung counting) 7

Be, 14C, 22Na, 32P, 33P, 45Ca, 99Tc, 147Pr, 226Ra, 

230Th, 232Th, 

Other radionuclides, bioassay by urinalysis

147Pr, 226Ra, 22Na, Other radionuclides, whole body counting

5.3.1.1 Solubility Classes

The solubility classes for particular radionuclides on the site are generally not specified. Unless the 

solubility class is clearly defined for a particular radionuclide, the dose reconstructor should use the 

class that is most claimant-favorable for the intake indicated by the bioassay data.

Tritium can be assumed to be in the form of tritiated water or water vapor because that is most typical 

at ANL-E (Bertelli 2003), and it is claimant-favorable because the dose from elemental tritium is 

significantly lower than the dose from tritiated water or water vapor. While the dose from organic 

forms of tritium can be significantly higher than for tritium in water, organic tritium compounds were 

not in widespread use at ANL-E (Bertelli 2003).

The corrosion or activation products (e.g. 60Co) are generally in the oxide forms. The ANL-E 

Technical Basis Document for Internal Dosimetry recommends use of the least soluble form of the 

radionuclide established for the element in ICRP 30 unless sufficient data are available to derive 

individual-specific retention (Bertelli 2003). For the purpose of dose reconstruction, in the absence of 

specific data, the ICRP 66 inhalation type that is consistent with ICRP 30 Class Y is Type S (ICRP 

1994).

According to Bertelli (2003) strontium, cesium and iodine intakes at ANL-E are considered to be ICRP 

30 Class D, consistent with ICRP 66 Class F. Plutonium oxides are assumed to be ICRP Class Y 

(ICRP 66 Inhalation Type S). All other chemical forms are assigned ICRP 30 Class W, consistent 

with ICRP 66 Class M. Uranium compounds are assigned different inhalation types based on 

solubility with UF6, UO2F2, and UO2(NO3)2 assigned to ICRP 30 Class D (ICRP Type F); UO3, UF4, 

and UCl4, considered ICRP 30 Class W (ICRP 66 Type M); and UO2 and U3O8, ICRP 30 Class Y 

(ICRP 66 Type S).

5.3.1.2 Route of Intake

In general, the most reasonable routine route of intake for the radionuclides at ANL-E would have 

been inhalation. Under accident conditions, intake via ingestion or through breaks in the skin could 

have occurred. In such cases the accident or incident reports should provide the dose reconstructor 

with information on which to base a judgment as to route of intake.
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5.3.1.3 Particle Size

The dose reconstructor should assume that the route of intake for radionuclides was inhalation and a 

particle size distribution of 5 μm activity median aerodynamic diameter (AMAD), the default ICRP 66 

particle size. If specific information is available to the contrary, that information should be used.

5.3.2

The Technical Basis Document for Internal Dosimetry (Bertelli 2003) describes the existing bioassay 

program in detail. As of 2003, 400 of the 5,000 ANL-E workers participated in special and routine 

bioassay programs. The bioassay programs were developed in accordance with 10 C.F.R pt. 835, 

“Occupational Radiation Protection,” which was based on the concepts in ICRP Publications 26 and 

30 (ICRP 1977, 1979). ANL-E also used the DOE Standard Internal Dosimetry (DOE 1998) and 

Internal Dosimetry Program Guide (DOE 1999) as guidance for the bioassay programs. Current 

policy requires internal radiation monitoring for all workers exposed to surface or airborne radioactive 

contamination where the worker could receive 100 mrem committed effective dose equivalent (CEDE) 

or a committed organ dose in excess of 5 rem.

Current Bioassay Program

The bioassay program includes urine and fecal analysis as well as in vivo measurements. The type of 

bioassay depends on the physical and biological characteristics of the radionuclides to which a worker 

could have been exposed.

In addition to those requirements for bioassay, there are other objectives of the program including 

providing assurance for the radiation protection program. Bertelli (2003) notes that workplace 

monitoring is the primary means of controlling and limiting worker exposure. However, the internal 

dose to an individual worker was assessed only if in vivo or in vitro measurements confirmed intakes 

resulting in a CEDE of 1 mrem or greater. 

5.3.2.1 Scheduling of Bioassay

The current policy for selection and frequency of bioassay is stated in an appendix to Bertelli (2003). 

In general, individuals have been selected based on several specific conditions. The primary criterion 

that dictated scheduling and type of routine bioassay samples was the goal of maintaining potential 

missed dose at less than 100 mrem in a year. The following conditions required bioassay 

measurements:

• Unanticipated radiological incidents

•

 – Bioassay was required for unanticipated incidents in

which radioactive materials had the potential to be inhaled, ingested, absorbed through the 

skin, or introduced into the body by a puncture or wound. For such incidents, a sufficient 

number of samples were taken to permit a reasonable estimate of the intake, and therefore the 

CEDE, for each radionuclide that could have entered the body. 

Elevated Air Samples

•

 – Work in areas where exposures could exceed 4 derived air 

concentration-hr in any 1-week period required bioassay. 

Routine Respirator Use

•

 – Routine respirator use required bioassay sampling. 

New Operations with Dispersible Radioactive Materials – New operations for which there were 

no previous monitoring data required bioassay in addition to air monitoring.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 12 of 42

• Confirmatory Bioassay Samples

•

 – Bioassay was required for approximately 10% of individuals 

who did not meet the other criteria as a check to ensure that these individuals did not need 

routine bioassay. 

Initial Work With Dispersible Radioactive Materials

•

 – Bioassay was required for initial work 

with dispersible radioactive materials as baseline samples.

Termination Samples

5.3.2.2 Specific Radionuclide Bioassay Methods

 – All individuals who worked with dispersible radioactive materials were 

required to submit termination samples.

5.3.2.2.1 Tritium Urine Bioassay

Because tritiated water is uniformly distributed in the body, tritium bioassay is relatively simple. A spot 

sample is sufficient to determine tritium activity as long as the tritium has had sufficient time to 

equilibrate in the body. Bertelli (2003) recommends sample collection at home or through multiple 

voiding. The samples were analyzed in the liquid scintillation counter. The minimum detectable 

activity (MDA) for the procedure was 1,000 dpm/L. 

5.3.2.2.2 Cobalt-60 and Other Activation Products

Neutron activation products like 60Co are present in the reactor areas and are most likely to become 

airborne during decontamination. Shorter-lived activation products such as 54Mn, 59Fe, and 58Co are 

also found in workers in and around nuclear reactors. Because they are generally gamma emitters, 

body burdens can be evaluated through WBC and with fecal and urine bioassay. The primary 

methods for bioassay for activation products were urine and fecal analyses, but in vivo analyses were 

also employed. Table 5-2 lists the MDAs for urine and fecal bioassay for activation products. 

Table 5-2. MDAs for activation product bioassay (Bertelli 

2003). 

Nuclide

WBC 

(nCi)

Urine sample

(dpm/L) Fecal sample (dpm/d)

Co-58 1 20 10

Co-60 1 20 10

Mn-54 1 20 10

Fe-59 2 40 20

5.3.2.2.3 Strontium

Potential intakes of 90Sr are often assessed using 137Cs as an indicator because both nuclides have 

comparable yields from the fission of 235U, but Bertelli (2003) describes specific bioassay methods for 

90Sr. 89Sr is initially present after release of fission products, but it is not likely to be a significant factor 

in the dose to workers because of its relatively short half-life (50.5 d). 

Bioassay techniques for strontium were primarily urine analyses. For urine bioassay, a stable 

strontium carrier is added to the sample, and the strontium is separated chemically. The samples are 

counted overnight, and the results are corrected for the ingrowth of 90Y. Bertelli (2003) lists the MDA 

for urine as approximately 0.3 dpm/L. Bertelli (2003) indicates that fecal analysis for 90Sr was not 

routinely used.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 13 of 42

5.3.2.2.4 Cesium

Cesium-137 is present at ANL-E as one of the most significant fission products from a worker dose 

perspective. Because it is more dispersible than 90Sr, 137Cs is the major component of fission product 

intake. It is easily detected by both in vivo and in vitro bioassay methods, and it is a good indicator of 

fission product exposures. 

In vivo measurements were made using the 8- by 4-in. NaI(Tl) detector and counting for 20 min. The 

WBC technique is capable of detecting 1 nCi of 137Cs. Annual in vivo measurements were 

recommended for workers exposed to 137Cs. 

The MDA for urine bioassay is not specified, but Bertelli (2003) notes that it is comparable to the MDA 

for in vivo measurements. Urine bioassay was recommended for instances where the worker could 

have been exposed to 90Sr as well as 137Cs.

5.3.2.2.5 Iodine

The iodine isotopes of primary interest at ANL-E are 131I and 125I. Iodine-131 is a fission product and 

125I was used in biological research at the site. Other iodine isotopes that could have been present as 

a result of fission or produced in accelerators by activation have short half-lives, except 129I. However, 

129I may have been present but unless concentrated, in “negligibly small quantities” (Bertelli 2003). 

(Iodine-129 is included in Table 5-3 for completeness although there is no indication in the available 

documents that bioassay for this isotope was routinely performed.) Iodine exposures are assumed to 

occur primarily as a result of accident situations because use of containment, ventilation, and 

respiratory protection should prevent routine inhalation exposures.

Bioassay measurements are primarily in vivo methods. Iodine-131 can be readily detected using a 

NaI or phoswich detector system. Table 5-3 lists the MDAs for specific types of detection systems. 

Table 5-3. MDAs for iodine isotopes (adapted from Bertelli 2003). 

Type of measurement

MDA (nCi)

I-125 I-131 I-129

WBC Log counter (8” x 4” NaI 

system)

Not detectable 1.0 (based on a 20-min count over 

the chest region)

Not detectable

5-in.-diameter phoswich detector 

( for thyroid counts)

0.03 0.03 0.03

5.3.2.2.6 Plutonium and Americium Compounds

Transuranic nuclides are present in research areas, standards laboratory, reactor fuel areas, and 

reactor D&D areas. The plutonium is a mixture of isotopes with 239Pu and 240Pu as the primary 

nuclides of concern. The composition of the plutonium mixture “depends on the length of time the fuel 

was irradiated, the time since irradiation, and the time since processing of the fuel or purification of the 

plutonium” (Bertelli 2003). Plutonium-241 is also present at ANL-E along with its decay product 

241Am. Some facilities employ 236Pu and 238Pu as tracers. 

The ANL “Draft Environmental Impact Statement” (ANL 1979) presents the isotopic compositions of 

two plutonium mixtures. For the scenario of a glovebox explosion in the New Brunswick Laboratory, 

the isotopic composition considered to be “representative of the most hazardous material expected to 

be received (i.e., from LWR recycle) and is not typical of present operations” was 2% 238Pu, 58% 

239Pu, 23% 240Pu, 12% 241Pu, and 5% 242Pu by mass fractions. For the scenario of a plutonium fire in 

one of the Building 350 vaults, the plutonium isotopic considered to “bound all similar risks at ANL” 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 14 of 42

was 0.1% 238Pu, 70.9% 239Pu, 24% 240Pu, 4% 241Pu, and 1% 242Pu by mass fractions. Values for the 

241Am component were not given. Dose assessments were performed either by assuming that all of 

the 241Pu had converted to 241Am or none of the 241Pu had converted to 241Am, and the higher of the 

two assessments was used. 

Unless case-specific information is available, the DR may consider the LWR recycle isotopic 

composition for maximizing assessments and the second scenario isotopic composition for best 

estimates. In lieu of 241Am data, the DR may consider the approach used by ANL in the draft impact 

statement.

Bioassay methods at ANL-E include in vivo measurements using phoswich detectors, urinalysis, and 

fecal sampling. The nominal MDA for 239Pu, for a chest-wall thickness of 2.5 cm is approximately 

80 nCi for a 20 min count, but it is dependent on chest-wall thickness. This MDA does not meet the 

criterion that the bioassay be able to detect an intake that would result in an annual dose of 

100 mrem. Therefore, in vivo bioassay under these conditions is not adequate for routine 

measurements.

Urine and fecal bioassay are necessary for determining intakes of plutonium. Urine bioassay detects 

soluble and moderately soluble forms of plutonium, but fecal bioassay is necessary to assess intakes 

of insoluble plutonium. Urine bioassay requires an overnight or 16-hr urine sample. The document 

does not describe how the 16-hour sample volume is normalized to represent a 24-hour excretion. All 

of the actinides are analyzed as a group from a single urine sample. Fecal samples are ashed, 

dissolved, and electrodeposited onto planchets. Presumably, the planchets are counted in an alpha 

spectrometry system that allows separation of individual nuclides. The MDAs for fecal analysis are 

approximately 0.05 dpm/sample. 

Workers with known internal depositions of plutonium can be required to participate in a long-term 

bioassay monitoring schedule. The urine bioassay detection limits (DL) for 239,240Pu and 241Am are 

approximately 0.02 dpm/L (Bertelli 2003).

5.3.2.2.7 Uranium

Uranium is present at the ANL-E site in hot cells, hoods, glove boxes, radioactive waste, irradiated 

reactor fuel, and in research areas generally mixed with other radionuclides. Uranium bioassay is not 

currently part of the routine bioassay program (Bertelli 2003). However, should it be necessary, urine 

bioassay and in vivo methods could be used. 

5.3.2.3 Bioassay Schedule

The scheduling of bioassay measurements, described in a 1997 Health Physics procedure, was 

based on frequency of exposure with the general protocol given in Table 5-4 (Marchetti, Holtzman, 

Keane 1997). Table 5-5 lists the types of bioassay required for specific types of exposures, and Table 

5-6 lists the codes for those analysis types.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 15 of 42

Table 5-4. Scheduling of bioassay samples. 

Class Work assignment Sample frequency (d)

1 Persons who work directly with radioactive materials with a frequency of at 

least:

Daily, in situations such as plutonium glove box decontamination and 

volume reduction or similar work

30

4 d/month 91

1–2 d/month 182

2 Persons who work in the same area as Class 1 but do not work with 

radioactive materials

182

3 Persons who regularly provide a service to radioactive material work areas 182

4 Persons who provide a service as needed or visit radioactive material 

work areas 1 to 2 hr per week

182

Nonscheduled personnel Only when they have been in or 

near an area where a known or 

suspected release has occurred

Table 5-5. Required bioassays by exposure. 

Radionuclides Types of bioassay required Types of analyses

Irradiated reactor fuel (including mixed 

fission and activation products)

Urine, WBC Gamma spectrometry, actinides, Sr-90, beta 

liquid spectrometry

Iodine Urine, WBC, thyroid Gamma spectrometry, beta liquid spectrometry

Tritium (HTO or organic) Urine beta liquid spectrometry

Uranium Urine, WBC Gamma spectrometry, actinides

C-14, S-35, P-32, P-33 Urine Beta liquid spectrometry, S-35, P-32

Sr-90 Urine Sr-90

Actinides – transportable Urine Gamma spectrometry, actinides

Actinides – nontransportable Fecal Gamma spectrometry, plutonium, uranium, 

americium, thorium 

Lung count Gamma spectrometry, X-rays

Table 5-6. Analysis codes. Analysis code Description

ACT All alpha-emitting actinides (actinium, thorium, protactinium, uranium, neptunium, 

plutonium, americium, curium, berkelium, californium, einsteinium)

ALP Alpha counting – heavy elements precipitated from urine

AM Americium, alpha-emitting isotopes

BLS Beta emitters by liquid scintillation counting

CA Ca-45

GAM Gamma spectrometry

NI Ni-63

P P-32

PM Pm-147 and beta-emitting actinides

PU Plutonium, alpha-emitting isotopes of Pu and Np

RA Ra-226

S S-35

SR Sr-89, Sr-90

T Tritiated water

TC Tc-99

U Uranium, alpha-emitting isotopes

TH Thorium, alpha-emitting isotopes

WBC Whole-body count, organ count, lung count

Lung Lung count, low-energy photon spectrometry in vivo

THY Thyroid count, low-energy photon spectrometry in vivo
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 16 of 42

5.3.3

The bioassay methods and frequency before the development of Bertelli (2003) are described in 

various monthly bioassay reports and other documents, including memoranda, reports, and 

responses to inquiries.

Bioassay Before 1995

5.3.3.1 Bioassay from 1946 to 1972

Nicksen (1946) cites maximum permissible levels (MPLs) of radionuclides in urine and refers to fecal 

and sputum analyses but notes that no MPL had been established. The MPL for plutonium in urine 

was given as 13.3 dpm/24-hr sample based on a body burden of 1 μg. This implies that bioassay 

programs were carried out but does not specifically describe these programs.

A 1947 letter from the Associate Director of ANL to the U.S. Atomic Energy Commission (AEC), notes 

that blood, urine, and fecal analyses were performed to form a record of the body content of toxic 

materials such as plutonium, uranium and beryllium for each individual exposed to radiation or toxic 

materials (Hilberry 1947). No details of the program were provided.

The Procedures of the Bioassay Group at the Argonne National Laboratory (Myers et al. 1952) 

describes the scheduling, recordkeeping, and analysis procedures for internal dose assessment in 

1952. The document notes that urine and feces were analyzed on a regular schedule. The samples 

could be analyzed for tritium, 14C, gross beta (nonvolatiles in boiling HNO3 with beta activities of 

energies greater than 0.3 MeV), radium, protactinium, polonium, and gross alpha (thorium, plutonium, 

neptunium, americium, and curium).

The scheduling of bioassay was dependent on the amount of radioactive material handled, frequency 

of exposure, chemical and physical nature of the material, apparent biological dangers of the material, 

working conditions, recommendations from the supervisors, and other factors. For example, a worker

with sufficient activity to take in more than the MPL in a single event would have monthly analyses at 

a minimum. If a worker was handling tracer levels of a low-energy beta emitter so that intake of the 

entire amount used over the year would not exceed the MPL, that worker would be scheduled for 

annual bioassay. Individuals falling between these two extremes would be scheduled for quarterly or 

semiannual bioassay. Employees who worked in technical areas were requested to submit both urine 

and fecal samples at termination. Urine samples were collected in 1-quart glass bottles and fecal 

samples in 1-pint cardboard cartons.

The laboratory kept three types of notebook records: A workbook listed the types of analyses 

required, volume of the sample and volume analyzed, date of analysis, and analyst identification; a 

counter notebook listed the results of the analysis; and a record book contained a permanent record 

of all of the information about the analysis. If an individual bioassay showed concentrations greater 

than the pre-set MPLs, an immediate resample was requested. Table 5-7 lists the 1952 MPLs, which 

were somewhat different from Bioassay Operational Levels from an unidentified document and 

included in the materials used in an epidemiologic study of workers who received doses in excess of 

5 rem in any year (Strom 1982). According to interviews at ANL-E in 1982, the revised values were in 

effect around 1960. Table 5-7 lists the later values in parentheses. The table for 1960 included an 

entry for beta activity with a normal range of 2,000 to 10,000 dpm/1,500 mL with a resample activity 

level of 12,000 dpm/1,500 mL. The reported beta activities for routine urine samples typically 

included naturally occurring 40K.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 17 of 42

Table 5-7. MPLs, 1952 (adapted from Myers et al. 1952) and after 1959.

a

Nuclide 

units

MPL in body MPL in urine (1960 and 

later) 

[dpm/24-hr or /1,500 mL]

MPL in feces 

(dpm/24-hr sample)

Amount in urine requiring 

resample (1960 and later) 

μCi μg [dpm/1,500 mL]

Ac-227 0.04 8.8 175 0.7

Am-241 0.04 0.01 4.4 (4) 65 0.7 (0.4)

Cf-242 0.04 8E-06 350 3,500 40

Pa-231 0.04 0.8 8.8 0.9 0.7

Po-210 0.005 N/A 300 (25) 4,500 40 (2.5)

Pu-239 0.04 0.6 4.4 (4) 4.4 0.7 (0.4)

Ra-226 0.1 0.1 22 (6) 220 12 (1.0)

Th-230 0.04 1.4 8.8 90 0.7

Th-234 0.8 Not reported Not reported Not reported

Np-237 0.04 57 8.8 90 0.7

H-3 1E04 N/A 6E08 (8E07) 6E07 1.5E07 (7.5E06)

Sr-90/Y-90 1 N/A 1000 (400) 2000 1000 (40)

a. N/A = not applicable. 

Monthly bioassay reports were submitted to Industrial Hygiene and Safety (Pingel, 1953; Robinson 

1955) detailing the detectable amounts of radioactivity in urine and feces. The 1953, 1954, 1961, and 

1962 monthly reports were reviewed (Pingel 1953 Robinson 1955, 1962, 1963).

The 1953 reports for the first 10 months of the year listed all results whether the activity was 

detectable or not. The vast majority of gross alpha results in urine were listed as zero dpm/mL. 

However, the DL was not defined in these reports. Repeat samples were requested for individuals

whose gross alpha activity was 1.0 dpm/mL or greater. No repeat samples were requested for 

concentrations less than 1.0 dpm/mL. Carbon-14 and 3

H in urine were reported; they were nearly all 

shown as “L 25” dpm/mg for 14C and “L1000” dpm/cm3 for 3

H. Presumably these values were the 

DLs.

The minimum positive fecal sample gross alpha concentration for 1953 was 0.2 dpm for a 15.4-g 

sample. In contrast to the urine samples, more than half of the fecal samples were positive.

In 1954 the lowest reported activity in urine for a group of radionuclides presumably analyzed together 

(plutonium, neptunium, thorium, actinium, americium, and curium) was 0.3 dpm/1500 mL. The lowest 

reported activity for those radionuclides in feces was 0.3 dpm for a total sample (in that case, 28.4 g). 

Several of the monthly reports listed fluoride-insoluble alpha activity. “Fluoride insoluble” is another 

term for gross alpha analysis and reported as “Alpha.” According to Procedure ANL-4509 (1951) the 

process was co-precipitation with bismuth phosphate and lanthanum fluoride. In 1975 the gross alpha 

procedure used cerium fluoride instead of lanthanum fluoride. During the first part of the year, gross 

alpha in urine and feces was given in the report. The values above are the lowest values given either 

for the specific radionuclides or for gross alpha (fluoride-insoluble) activity (or both). 

Radium-226 and 235U were reported separately. One positive tritium urine sample was reported in the 

January 1954 Bioassay Report (collection date 12/2/53) (Sedlet 1954). This indicates that either 

tritium analysis in urine was not routine or that there were very few positive samples.

Blood and sputum analyses were reported for several months during the first half of 1954. 

Several individual bioassay records for 1951 to 1957 were reviewed to determine the recordkeeping 

method. The records were kept on cards that included the date, bioassay type, element for analysis, 

excretion per 1,500 mL (not explained on the card but which might be the ratio of 1,500 cm3 to the 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 18 of 42

actual volume of the sample), type of exposure, hours worked, air analysis of the area, and remarks. 

The column for type of exposure recorded the pH of the urine in many cases rather than type of 

exposure. A bioassay monthly report from 1958 (author unknown) for Great Lakes Carbon Co. 

reported results as less than 5 mg per 1,500 cm3 (This is most likely a typographical error with “m” 

indicating the prefix “micro” instead of “milli”.).

A 1959 bioassay report for Great Lakes Carbon Company showed worker concentrations for enriched 

uranium down to 0.2 dpm per 1,500 cm3 (Sedlet 1959). The MPL for enriched uranium was given as 

50 dpm per 1,500 cm3

. 

The bioassay reports for 1961 and 1962 were more descriptive than for 1953 and 1954 (Pingel 1953, 

Robinson 1955); however, they included only the results for nonroutine and special analyses. The 

analyses performed during that period included alpha (plutonium, neptunium, thorium, actinium, 

americium, and curium), beta activity, iodine, plutonium, polonium, radium, tritium, uranium, and other 

miscellaneous nuclides (e.g., 99Tc, 65Zn). Gross alpha data were reported in some cases. The 

reports did not generally specify the type of bioassay but, because the data were reported as 

disintegrations per minute per specified volume, it can be assumed that they refer to urine bioassay. 

Several reports from that period indicated activity in fecal samples and nasal swabs. In one case, a 

blood sample was indicated with a value of 0 dpm/1,500 mL. Because it is unlikely that a 1,500-mL

blood sample was analyzed, this probably was a typographical error in the report. In another case, 

skin scrapings were analyzed. Because only special and nonroutine sample data were included in the 

monthly reports, it is unclear whether routine fecal analyses occurred during that period.

For most radionuclides, the concentrations were reported in disintegrations per minute per 1,500 cm3

. 

The 3

H concentration was reported in disintegrations per minute per cm3. The DL for 3H was 1,000 

dpm/cm3 in the 1950s and early 1960s. Bioassay program reports for 1967 indicated a DL of 100 

dpm/mL for 3

H (Sedlet 1968). The DLs for other radionuclides were not easily discernable from the 

annual reports. Beta concentrations were often reported as “normal range” rather than a specific 

value, but the normal range was not defined. The lowest alpha emitter data was given as zero with no 

minimum detectable concentration (MDC) value provided. Uranium concentrations were reported in 

disintegrations per minute per 1,500 cm3 without reference to isotopic ratios or an MDC. 

At least one of the monthly reports referred to requests for resampling and follow-up bioassay for 

individuals until the tritium concentration was less than 10,000 dpm/cm3

. There is no statement in the 

report that this was standard practice. 

A DL for polonium was given as 2 dpm/1,500 cm3 for a special analysis (volume of 100 mL) and 4 

dpm/1,500 mL for a routine volume of 50 mL in one of the monthly reports (Sedlet 1962). 

Monthly reports for 1968 were reviewed (Sedlet 1969). These reports contained information similar to 

that included in the 1961 to 1962 reports. It is unlikely that significant changes were made in the 

program in the intervening years. Results were given only for special and nonroutine samples. The 

list of analytes remained generally the same with the addition of gamma activity. The gamma 

activities specified radionuclides and listed a result of “normal” for individuals whose spectra were 

presumably in the range of background. A DL for plutonium of 0.2 dpm was given for a fecal sample.

A sample page of bioassay data from 1969, presumably a page from a bioassay log, showed 

quarterly normal (sic) uranium bioassay data. The log provided the data by location and worker. 

Quarterly results were recorded for most of the individuals with only one or two records for 1969 for 

some individuals. All of the results for those sample pages were less than 5 with no units indicated. 

There was no indication on the log page regarding who made the log entries. 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 19 of 42

Estimated decision levels, discussed in Section 5.3.4 were derived from the “less than” values in the 

Bioassay Program Reports and, in a few cases, the lowest reported positive value. 

5.3.3.2 Urine and Fecal Analysis from 1973 to 1995 

ANL-E developed a Health and Safety Manual before 1973 (ANL 1973). That document indicates that 

industrial safety and health physics policies had previously been issued through the Laboratory’s 

Policy and Practice Guide. The Health and Safety Manual was subject to revisions over time. 

Chapter V-13 of the Health and Safety Manual dated March 1, 1975, describes bioassay procedures 

(ANL 1975). The document notes that urine samples were requested from all new employees, 

returning employees, and temporary employees with previous occupational exposure. Urine samples 

were requested from all terminating employees. The document did not exclude employees who did 

not work in areas with dispersible radionuclides, which indicates that the program likely applied to all 

employees.

Specific requirements for periodic urine and fecal analyses or in vivo measurements included the 

potential to receive a dose commitment in any calendar quarter in excess of 10% of the standard. 

The document does not specify bioassay methods or DLs, and it does not provide information about 

selection of employees for in vivo bioassay. 

The 1976 monthly Bioassay Program Reports include the results of special and nonroutine bioassay 

samples (Fairman 1976). The records show less-than values for plutonium in fecal samples at 

0.05 dpm/sample and 0.05 dpm/1,500 mL for urine bioassay. The basis for this presumed detection 

limit is not provided in the documents. The less-than values for tritium in urine was 25 dpm/mL. A 

reported value of 0.25-dpm/mL was a typographical error because all other less-than values for tritium 

for 1976 were given as 25 dpm/mL. The minimum value given for uranium in urine was simply zero, 

so no MDA could be discerned from the data. The lowest positive value for uranium was 0.1 

dpm/1,500 mL. 

Gamma activities were reported as “normal,” again with no specified MDA. Values for beta oxalate 

analysis were listed simply as “below detection limit.” Strontium and 14C minimum values were also 

listed as “below detection limit.” The apparent DL for 249Cf in feces was 0.1 dpm. The minimum 

values for radium and polonium and +3 and +4 actinides in urine were 0 dpm/1,500 mL. 

Table 5-8 lists various draft and final laboratory procedures used from 1970 to 1989 for urine and 

fecal analysis of analytes of concern. In most cases the specific nuclide procedures described in 

Table 5-8 were from a laboratory manual identified in the header on the procedure as IDOP:IDCL 

(presumably Internal Dosimetry Operational Procedures: Internal Dosimetry Chemistry Laboratory) 

Revision 00. The headers on the procedures were generally undated. However, several of the 

individual procedures had handwritten dates on the first page. In some cases, the procedures had 

identifying headers, e.g. ALP-1 for the gross alpha procedure. It is reasonable to assume that the 

urine and fecal bioassay requirements did not change significantly during that period.

The ANL-E internal dosimetry program was reviewed in 1982 in conjunction with an epidemiologic 

study of workers who had received doses greater than 5 rem in any year (the 5-rem cohort) (Strom

1982). The investigators noted that all but one of the 54 individuals in the 5-rem cohort had internal 

dose monitoring records in comparison to the ANL-E average of 10% of workers. This indicates that 

most of the individuals likely to be exposed to high levels of radionuclide concentrations and high 

external exposure rates were monitored. 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 20 of 42

Table 5-8. Summary of bioassay analysis methods, 1970 to 1989. 

Analyte Medium Description of method Date

S-35 Urine Low-background beta counting – Sulfate from ashed urine sample precipitated as 

barium sulfate and counted.

Undated

Ra-226 Urine and 

feces

Alpha counting – Coprecipitation of radium with lead sulfate, barium chloride, and 

barium sulfate separates radium from other activities and provides a mount 

sufficiently thin for alpha counting.

4/1/75 and 

12/1/88

Sr and Ba Urine and 

feces

Beta counting – Carriers added and carbonates precipitated from solution made 

basic with NaOH. Sample scavenged with ferric hydroxide after a series of 

nitrate precipitations from concentrated nitric acid solutions. Ba separated from 

Sr as the chromate and Sr is precipitated and weighed as the oxalate. Ba 

precipitated and weighed as the chloride. Precipitates mounted on stainlesssteel planchets for beta counting.

6/12/79 and 

undated

IDOP:IDCL* 

Rev. 00

*(See text) 

Ni-63 Urine Liquid scintillation counting – Aliquot wet-ashed and residue dissolved in dilute 

nitric acid. Ni precipitated as the hydroxide, dissolved in HCl and passed through 

anion exchange column to remove Fe. Ammonium citrate is added if rare earths 

or Zr-Nb are potential interferences. Ni precipitated with dimethylglyoxime. 

Precipitate dissolved in concentrated nitric acid, converted to the oxide then to 

the chloride. Chloride dissolved in water and incorporated into a gel scintillation 

cocktail for liquid scintillation counting. Recoveries are >95% and the counting 

efficiency is approximately 40%. MDA at the 95% probability level is 

<33 dpm/1000 mL.

12/1/88

Tritium Urine Liquid scintillation counting – Urine vacuum-distilled and scintillation solution 

added to an aliquot of the distillate. Sample, blanks, and spikes counted.

4/2/75, 

8/17/87 and 

IDOP: IDCL 

Rev. 00

Tc-99 Urine Low-background Geiger-Mueller or proportional counter – Technetium (IV) 

carried on ferric hydroxide precipitate, then oxidized to Tc (VII) with concentrated 

nitric acid. Technetium coprecipitated with cupric sulfide, filtered, and mounted 

for counting.

12/1/88

Gamma 

emitters

Urine Gamma spectrometer – Samples counted by gamma spectrometry using a Ge 

detector.

8/17/87 and 

Undated 

IDOP:IDCL 

Rev 00

C-14 Urine Liquid scintillation counting – Aliquot of urine sample added directly to gel 

scintillation cocktail.

7/1/68, 

4/3/75, and 

11/30/88

Pu, Np, 

Am, Cm, U, 

Th

Feces Alpha spectrometry – Sample ashed, yield monitors added, and leached with 

nitric acid in microwave oven. Pu (+Np) and Th purified using anion exchange 

column. Am (+Cm) and U purified using carbamoyl methylphosphine oxide 

dissolved in tributyl phosphate. Counting plates made for Pu and U fractions 

using hexone extraction; for AM and Th fractions by electrodeposition. 

4/8/75, 

5/27/87 and 

IDOP:IDCL 

Rev 00 

8/04/89

Po Urine Low-background alpha spectrometer – Po electrochemically plated onto Ag 

planchet from dilute HCl solution of ashed urine salts.

Undated 

IDOP:IDCL 

Rev 00

Ca-45 Urine Liquid scintillation counting - Aliquot of raw or ashed urine sample mixed directly 

with gel scintillation cocktail. 

8/25/75

Beta 

activity

Urine Low-background beta counter – Coprecipitation with strontium oxalate from urine. 4/9/75 and 

12/1/88

Actinide 

and 

lanthanides 

Urine Alpha spectrometer and low-background beta counter – Actinides concentrated 

from urine using alkaline phosphate precipitation, separated from interfering 

elements using a column loaded with octyl(phenyl)-N-N-diisobutyl CMPO/TBP. 

Electrodeposited and counted in an alpha spectrometer then low-background 

beta counter for determination of Pr-147 or other lanthanides.

IDOP:IDCL* 

Rev 00 

6/12/89

*(See text)

Gross 

alpha

Urine Alpha counter – Heavy elements precipitated from the urine in alkaline phosphate 

then coprecipitated with bismuth phosphate then cerium fluoride. Cerium fluoride 

mounted for counting. Uranium does not come through the procedure. 

ALP-1 4/1/75 

and 11/16/84 

Rev.

Strom (1982) reports the investigators used the Bioassay Laboratory Routine Sampling Scheduling 

Criteria dated 1972 to extract dose information. Scheduling criteria were based on the amount and 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 21 of 42

frequency of radionuclides handled. The schedules ranged from bimonthly to annual. The schedule 

did not specify the type of bioassay, but most of the routine bioassay measurements during this period 

were urine analysis. 

5.3.3.3 In Vivo Bioassay

None of the bioassay monthly reports for the period provided data from lung counting or WBC. The 

1971 Bioassay Program Reports described the Radioactivity Body Counter (RBC), later termed the

Whole Body Counter (WBC), as it was constructed and noted that the instrument would need to be 

calibrated. No calibration records were found.

ANL (1990) described the WBC facility. The procedure did not indicate the date this facility went into 

service, but it was built in 1962 (Wescott & O-Rourke 2001). The facility consisted of three lowbackground counting rooms housed in an underground concrete vault designed to minimize 

background. The shielded counting rooms housed three different types of detection systems: 

• Room 1 housed a system with two 11.5- by 4-in. NaI(Tl) detectors mounted above and below 

a bed that could be moved. The system simultaneously scanned a supine subject.

• Room 2 had a single 6-in.-diameter, 8-in.-long NaI(Tl) detector that was termed the “log” 

detector. The subject sat in a chair or reclined on a curved bed with the detector above the 

trunk of the subject.

• Room 3 was designed for counting low-energy photons. The system employed a pair of 

Phoswich detectors that consisted of thin 5-in.-diameter NaI(Tl)/CsI(Tl) detectors. The system 

was used for measuring actinides in the lung, and the phoswich detectors were also used for 

measuring 125I in the thyroid and actinides in other organs. A gas proportional counter in this 

room distinguished between internally deposited actinides and external contamination.

The counting efficiencies for these systems were determined for various configurations using three 

principal methods:

• Mathematical analysis of counting data obtained from measurements of an internally 

contaminated subject and point source measurements of the specific gamma emitter of 

interest

• Measurements using an anthropomorphic phantom containing a known quantity of the 

radionuclide of interest

• Measurements of human volunteers after administration of tracer quantities of radionuclides

ANL (1990) did not provide the efficiencies and DLs. 

Toohey (1980) described the detectors used in the WBC facility, initially called the Body Radioactivity 

Counter (BRC), for assessment of transuranic elements. Two detectors were used: A phoswich 

consisting of a thin NaI(Tl) crystal backed by a CsI(Tl) crystal for detection of low-energy gamma 

photons or X-rays, and a xenon-filled proportional counter for detecting external contamination. For 

mixed plutonium and americium contamination, the 241Am 60-keV gamma ray was used to quantify 

the amount of 241Am. The contribution from the 241Am to the X-ray region (12 to 24 keV) was 

subtracted to determine the plutonium contribution. The alpha activity could be determined if the 

plutonium isotopic ratios were known. Toohey (1980) did not provide the DL for the method. 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 22 of 42

An April 2005 memorandum stated that the Phoswich detectors were no longer in service and that the 

MDAs for the instrument “could be taken as” three times the detection level (DL), which was not 

specified (Keane 2005). ANL-E documentation sometimes uses decision level and the code Dc to 

indicate DLs.

5.3.4

5.3.4.1 Minimum Decision Levels for 1946 to 1985 

Estimated Decision Levels and Minimum Detectable Amounts or Concentrations

There has been no evidence found that MDAs were assessed or recorded before 1995. Instead, 10% 

of maximum permissible excretion levels, action (repeat sampling) levels, and reporting levels have 

been used. These levels, if found or discerned, are reported in previous sections. The information is 

significantly incomplete for certain periods and radionuclides. 

The decision levels for years prior to 1985 can be inferred from the “less than” values for individuals 

reported in the Bioassay Program Monthly Reports. These values, where a positive value is assigned 

for an analyte result, can also be determined from the smallest positive measurement result. Tables 

5-9 and 5-10 provide those values for urine bioassay and fecal bioassay, respectively (When similar 

“less than” values are given in the bioassay Monthly Reports only the earliest value is included in the 

table.)

Table 5-9. Summary of Detection Limits for Urine Bioassay Based on Reported “Less Than”Values . 

Nuclide “Less than” value

Earliest 

Year Source of Information Comments

Actinides 0.5 dpm/1500 mL 1958 Memo. J. J. Robinson to Novak. 

Oct 29, 1958.

Less than value, specified as 

Pu, Np, Th, Ac, Am and Cm

Actinides 0.1 dpm/1500 mL 1967 Sedlet. Bioassay Report. 1967 Unspecified nuclides

Alpha 0.4 dpm/1500 ml 1948 Individual bioassay record

Alpha 0.5 dpm/1500 ml 1951 Individual radiation exposure 

record

Alpha 0.6 dpm/1500 ml 1952 Sample log entry

Alpha 0.3 dpm/1500 ml 1959 Individual bioassay record

Alpha 0.05 dpm/L 1960 Memo. Keane to Dolacek. July 7, 

1993

Denoted as minimum 

detectable urine alpha activity 

concentration using “fluoride 

insoluble method”

Alpha 0.1 1967 Memo Sedlet to Novak (11/3/67) Reported as fluoride insoluble 

alpha (mostly due to Pu)

Alpha 0.133 dpm/1500 

mL

1985 Individual Radiation Exposure 

Record

Ca-45 0.5 nCi/1500 mL 1975 Fairman, Bioassay Program 

Reports- 1975

C-14 25 dpm/mg C 1948 Individual bioassay record

C-14 2 dpm/mL 1969 Sedlet. Bioassay Report. 1969

Californium 0.1 dpm/1500 mL 1975 Fairman. Bioassay Report 1975 Unspecified isotope

H-3 100 dpm/cc 1948 Individual Bioassay Record This value appears to be in 

error given the later much 

higher “less than” values

H-3 1000 dpm/cc 1953 Pringle 1953 Bioassay Report

H-3 2000 dpm/cc 1960 Robinson 1960. Bioassay Report “Less than” values prior to and 

after 1960 were 1000 dpm/mL

H-3 100 dpm/mL 1967 Sedlet. 1967-1968. Bioassay 

Report
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 23 of 42

Nuclide “Less than” value

Earliest 

Year Source of Information Comments

H-3 25 dpm/mL 1969 Sedlet. 1969 - 1973. Bioassay 

Reports, Fairman. 1976. Bioassay 

Report

H-3 2.5E4 dpm/sample 1988 Individual bioassay record

Iodine 25 dpm/1500 mL 1967 Sedlet. Bioassay Report. 1967 Unspecified isotope

Ni-65 33dpm/100 mL 1988 Bioassay Procedures 12/1/88

Plutonium 0.1 dpm/1500 mL 1967 Bioassay Monthly Reports – 1967 

– Sedlet

Unspecified isotope

Plutonium 0.05 dpm/1500 mL 1976 Fairman. Bioassay Reports 1976 Unspecified isotope

Plutonium 0.033 dpm/1500mL 1980 Individual Radiation Exposure 

Records

Pu-238 0.05 dpm/1500 mL 1974 Bioassay Monthly Reports – 1974 

– Fairman

Pu-239 0.05 dpm/1500 mL 1974 Bioassay Monthly Reports – 1974 

– Fairman 

Pm-147 300 dpm/1500 cc 1962 Robinson. Bioassay Report. 1962

Pm-147 100 dpm/1500 cc 1967 Sedlet. Bioassay Report. 1967. 

Polonium 2 dpm/1500 mL 1961 Robinson, Bioassay Report.. 

1961, Sedlet.

Unspecified isotope

Pr-147 50 dpm/1500 mL 1970 Sedlet. Bioassay Report. 1971

Radium 0.5 –1 dpm/1500 mL 1967 Sedlet. Bioassay Report. 1967 Unspecified isotope

Radium 1.333E-1 dpm/mL 1985 Individual Radiation Exposure 

Record

“<”

Strontium 25 dpm/1500 mL 1970 Sedlet. Bioassay Report. 1970 Unspecified isotope

Uranium 1.65 (no units) 1951-57 Bioassay logs The <1.65 in the logs was 

defined as <1.65 alpha/1500 

ml ifor enriched uranium in a 

memo from DP ONeil to J. F. 

Ege Jr. (10/22/55)

Uranium 10 μg/1500 mL 1957 Individual bioassay cards Unspecified isotopes

Uranium 5 μg/1500 mL 1958 Memo. Duffy to Novak. Oct 29, 

1958

Unspecified isotopes

Uranium 0.2 dpm/1500 mL 1959 Memo Sedlet to Novak – 

Bioassay Results for Great Lakes 

Carbon

Uranium 0.6 dpm/1500 mL 1961 Individual bioassay record

Uranium 0.2 dpm/1500 mL 1967 Bioassay Monthly Reports from 

Sedlet

Unspecified isotopes

Uranium 0.2 dpm/mL 1985 Individual Radiation Exposure 

Record

Zn-65 200 dpm/1500 mL 1975 Fairman. Bioassay Report. 1975

Zr-95 200 dpm/1500 mL 1975 Fairman, Bioassay Program 

Report - 1975
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 24 of 42

Table 5-10. Summary of Reported “Less Than” Values for Fecal Bioassay. 

Nuclide “Less than” Value First 

Years

Source of Information Comments

Actinides 1 dpm/sample 

(31.2 g)

1967 Sedlet. Bioassay Report. 1967 Unspecified nuclides

Actinides 0.1 dpm/sample

(12 g)

1968 Sedlet. Bioassay Report. 1968 Unspecified isotope

C-14 25 dpm/mg 1960 Robinson. Bioassay Report. 

1960

Cf-249 10 dpm/sample 1971 Sedlet - Bioassay Report 1971

Cf-249 0.5 dpm/sample 1971 Sedlet. Bioassay Report. 1972

Cf-249 0.1 dpm/sample 

(12 g)

1976 Fairman. Bioassay Report. 

1976

Cm 0.1 dpm/12 g 1968 Sedlet. Bioassay Report. 1968 Unspecified isotope

Cs-137 10 dpm/sample (29 

g)

1978 Fairman. Bioassay Report. 

1978

Presumably 10 dpm per 

sample

Cs-134 10 dpm/sample 1978 Robinson – Bioassay 

Radioactivity Report -1978

Co-57 10 dpm/sample 1978 Robinson – Bioassay 

Radioactivity Report – 1978

Cr-51 50 dpm/sample 1978 Robinson – Bioassay 

Radioactivity Report - 1978

Fe-59 400 dpm/sample 1968 Bioassay Monthly Reports 

1967-1968 - Sedlet

Fe-59 10 dpm/sample 1978 Robinson - Bioassay 

Radioactivity Report - 1978

Np 0.1 dpm/12 g 1968 Sedlet. Bioassay Report. 1968 Unspecified isotope

Pu 0.1 dpm/sample 

(12g)

1968 Sedlet. Bioassay Report. 1968 Unspecified isotope

Pu 0.05 dpm/sample

(20 g)

1978 Fairman. Bioassay Report. 

1978

Presumably 0.05 dpm per 

sample

Pu-238 0.05 dpm/sample 

(67 g)

1975 Fairman. Bioassay Report. 

1975

Presumably 0.05 dpm per 

sample

Pu-239 0.05 dpm/sample 

(33g)

1975 Fairman. Bioassay Report. 

1975

Presumably 0.05 dpm per 

sample

Th 0.1 dpm/sample 

(12 g)

1968 Sedlet. Bioassay Report. 

1968

Unspecified isotope

Zn-65 20 dpm/sample 1972 Sedlet - Bioassay Program 

Report – 1972 

In lieu of definitive statements regarding minimum detectable activities, and as a guide for dose 

reconstructors, Tables 5-11 and 5-12 for urine and fecal bioassay have been created from data 

displayed in Tables 5-9 and 5-10. These data represent the best available information regarding a 

suggested value for detection level in lieu of definitive information. 

The radionuclides represented in Tables 5-11 and 5-12 are those that were present more or less 

consistently throughout the history of the ANL-E facility. Other radionuclides (berkelium, californium, 

cobalt, iodine, mercury, polonium, promethium, protactinium, radium, strontium, technetium, and zinc) 

were present only sporadically depending on the particular research programs that were being 

conducted during a specific year.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 25 of 42

Table 5-11.Estimated decision levels for urine bioassay by year. 

Year H-3 C-14 Gross Alpha Actinides† Pu U* Pu-238 Pu-239 Enr. U**

Units

dpm/cc

dpm/mL

dpm/mg 

C dpm/1500mL

dpm/

1500 mL

dpm/

1500mL

dpm/1500mL

(μg/L) dpm/1500mL

dpm/1500 

mL

1946 1000 25

1947 1000 25

1948 1000 25 0.4

1949 1000 25 0.4

1950 1000 25 0.4

1951- 1000 25 0.5 1.65 

1956 1000 25 0.6 1.65

1957 1000 25 0.6 (10)

1958 1000 25 0.6 0.5 (5)

1959 1000 25 0.3 0.5 0.2 0.2

1960 1000 25 0.05 0.5 0.2 0.2

1961- 1000 25 0.05 0.5 0.6 0.2

1966 1000 25 0.05 0.5 0.6 0.2

1967 100 25 0.1 0.1 0.1 0.2 0.2

1968 100 25 0.1 0.1 0.1 0.2 0.2

1969- 25 25 0.1 0.1 0.1 0.2 0.2

1973 25 25 0.1 0.1 0.1 0.2 0.2

1974 25 25 0.1 0.1 0.1 0.2 0.05 0.2

1975 25 25 0.1 0.1 0.1 0.2 0.05 0.2

1976 25 25 0.1 0.1 0.05 0.1 0.05 0.2

1977 25 25 0.1 0.1 0.05 0.1 0.05 0.2

1978 25 25 0.1 0.1 0.05 0.1 0.05

1979 25 25 0.1 0.1 0.05 0.1 0.05

1980- 25 25 0.1 0.1 0.033 0.1 0.05

1984 25 25 0.1 0.1 0.033 0.1 0.05

 † Sometimes reported as "Pu, Np, Th, Ac, Am, Cm" 

 *Values for 1976-1984 are based on lowest reported positive level (Robinson 1976)

 ** Based on lowest reported level (Sedlet 1959)

5.3.4.2 Estimated Minimum Detectable Activities for 1985-2001

The estimated MDAs for the years 1986 to 2000 were calculated based on reported 

standard deviations for an individual worker. There were 42 routine urine bioassay 

samples and 35 fecal samples analyzed for 13 radionuclides during that period. The 

reported standard deviation for each measurement was multiplied by a factor of 4.65 to 

obtain a MDA. The MDAs for each nuclide were averaged and the standard deviation 

calculated. The recommended MDA for each nuclide for the period is the average of the 

MDAs plus two standard deviations. The calculated values, given in Table 5-13, are 

consistent with current reported MDAs. The data showed no significant trends over time 

so the calculated values are applicable to the entire 1986-2000 time period. This method 

provides a reasonable but claimant favorable basis for estimating the MDA.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 26 of 42

Table 5-12. Recommended Decision Levels for Fecal Bioassay by year (dpm) based on 

“Less Than” Values (All units are dpm per sample)

Year

Gross 

alphaa Pu U-235a Ua Actinidesb Pu-238 Pu-239

1946-

1952

1953 0.1

1954 0.1 0.4 0.3

1955- 0.1 0.03 0.3 

1957 0.1 0.03 0.3

1958 0.1 0.03 16.3 0.3

1959 0.1 0.03 16.3 0.3

1960- 0.1 0.03 2.3 0.3

1966 0.1 0.03 2.3 0.3

1967 0.1 0.03 0.3 1.0

1968 0.1 0.2 0.03 0.1 0.1

1969- 0.1 0.1 0.03 0.1 0.1

1974 0.1 0.1 0.03 0.1 0.1

1975- 0.1 0.1 0.03 0.1 0.1 0.05 0.05

1977 0.1 0.1 0.03 0.1 0.1 0.05 0.05

1978- 0.1 0.05 0.03 0.1 0.1 0.05 0.05

1984 0.1 0.05 0.03 0.1 0.1 0.05 0.05 a Based on Lowest Reported Value where there was an insufficient number of “less than” values b

Based on Lowest Reported Value from 1953 to 1966.

Table 5-13. Estimated Bioassay MDAs for 1986 to 2000a

Nuclide Recommended MDA

Urine Analysis

(dpm/liter)

Fecal Analysis

(dpm/sample)

Am-241 0.036 1.77

Cm-242 0.059 0.55

Cm-244 0.030 NA

Cf-252 NA 0.10

Np-237 0.075 0.21

Pu-238 0.035 0.16

Pu-239 0.033 0.19

Pu-242 0.026 0.07

Th-228 0.052 2.82

Th-230 0.044 1.96

Th-232 0.031 1.00

U-234 0.068 1.82

U-235 0.032 0.60

U-238 0.059 1.49 a NA means not analyzed 

5.3.4.3 Current (2004-2005) Minimum Detectable Amounts or Concentrations 

In contrast to the previous time periods, there is a considerable amount of information available with 

regard to current MDAs and methods for their determination. A set of intra-laboratory memos (Zhang 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 27 of 42

2005) lists the MDAs or MDCs for bioassay analyses for the analytes currently included in the routine 

monitoring program at ANL-E. The general equation used to calculate the MDA or MDC is:

MDA = (4.65 SB + 3)/(K0.05 x T), (1)

where 

K0.05 = Lower limit of the 90% confidence level of the calibration factor K 

 SB = Standard deviation of the net background count

T = Count time

This method is consistent with the methodology recommended in the American National Standard 

Performance Criteria for Radiobioassay, especially the use of the lower limit of the 90% confidence 

level of the calibration factor (HPS 1996, Section A.5.4). 

5.3.4.3.1 Unseparated Actinides in Urine

Zhang (2005) describes in detail the methods used to analyze urine bioassay samples for actinides. 

Unseparated actinides include plutonium, neptunium, americium, uranium, curium, and thorium. Table 

5-14 lists the practical Dc’s and the a priori MDCs for calendar years 2004 and 2005. Results for this 

analysis might include test or analysis code “ACT.”

Table 5-14. 2004-2005 Practical Dc’s and MDCs for 

unseparated actinides in urine (Zhang 2005). 

Analyte Dc (dpm/L) MDC (dpm/L)

Pu-239,240 0.0070 0.044

Pu-238 0.0028 0.045

Pu-242 0.0021 0.040

Np-237 0.0103 0.049

Am-241 0.0028 0.045

U-234 0.0134 0.056

U-235 0.0036 0.048

U-238 0.0110 0.049

Cm-242 0.0254 0.095

Cm-244 0.0032 0.039

Th-232 0.0042 0.040

Th-230 0.0072 0.037

Th-228 0.0034 0.055

5.3.4.3.2 Gamma-Emitting Isotopes in Bioassay Samples 

Gamma-emitting isotopes include 7

Be, 22Na, 54Mn, 58Co, 60Co, 134Cs, 137Cs, and 40K. Table 5-15 lists 

the Dc’s and MDCs for urine and fecal bioassay samples. Results for this analysis might include test 

or analysis codes “GAM,” “GAMMA,” or a variation of these codes.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 28 of 42

Table 5-15. Dc’s and MDCs for gamma-emitting isotopes in urine 

and fecal samples (2004-2005). 

Analyte

Urine samples Fecal samplesa

Dc (dpm/L) MDC (dpm/L) Dc (dpm) MDC (dpm)

Be-7 87 199 na Na

Na-22 13 30 na Na

Mn-54 12 27 na Na

Co-58 9 21 na Na

Co-60 11.2 26 14 62

Cs-134 11 27 na Na

Cs-137 11 25 15 64

K-40 208 473 294 1,244

a. na = not analyzed.

Bertelli (2003) provided the following MDAs for activation product bioassay.

Table 5-16. MDAs for activation product bioassay (Bertelli 2003). 

Nuclide

WBC 

(nCi)

Urine sample

(dpm/L) Fecal sample (dpm/d)

Co-58 1 20 10

Co-60 1 20 10

Mn-54 1 20 10

Fe-59 2 40 20

5.3.4.3.3 Beta Liquid Scintillation Counting of Urine Samples

Liquid scintillation counting of urine samples involves counting a 0.003-L aliquot of the urine sample 

for 60 min. The beta energy spectrum is divided into three regions: Region A (Lowbeta) for 

low-energy beta emitters such as tritium; Region B (Midbeta) for medium-energy beta emitters such as 

14C, 35S, and 45Ca; and Region C (Topbeta) for unspecified high-energy beta emitters other than 40K, 

(Zhang, 2005). Table 5-17 lists the Dc’s and MDCs. Results for this analysis might include test or 

analysis code “BLS.”

Table 5-17. Dc’s and MDCs for beta liquid scintillation 

counting of urine samples (2004-2005). 

Dc (dpm/L) MDC (dpm/L)

Energy Region A: Lowbeta 1,324 3,242

Energy Region B: Midbeta 731 1,726

Energy Region C: Topbeta 743 2,120

5.3.4.3.4 Separated Actinides in Bioassay Samples

Separated actinides in bioassay samples include isotopes of uranium, plutonium, thorium, neptunium, 

americium, curium, and californium. Table 5-18 lists the Dc’s and MDAs. Results for this analysis 

might include a test or analysis code of the element symbol in capital letters (e.g., PU for plutonium 

isotopes).
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 29 of 42

Table 5-18. Dc’s and MDAs for alpha spectrometry of separated 

actinides in urine and fecal samples (2004-2005). 

Analyte

Urine samples Fecal samples

Dc (dpm) MDA (dpm) Dc (dpm) MDA (dpm)

U-234 0.017 0.045 0.022 0.082

U-235 0.010 0.042 0.016 0.081

U-238 0.016 0.045 0.021 0.079

Th-228 0.293 0.44 0.620 1.48

Th-230 0.273 0.31 0.324 0.55

h-232 0.019 0.055 0.066 0.28

Pu-238 0.032 0.10 0.039 0.17

Pu-239, 240 0.008 0.067 0.014 0.13

Pu-242 0.007 0.058 0.011 0.11

Np-237 0.005 0.059 0.010 0.12

Am-241 0.012 0.032 0.017 0.058

Am-243 0.009 0.033 0.016 0.063

Cm-244 0.004 0.023 0.008 0.046

Cf-252 0.006 0.023 0.009 0.043

5.3.4.3.5 Strontium-90 in Urine

The Dc is 0.23 dpm/L. The MDC is 0.66 dpm/L. Bertelli (2003) lists the MDC for urine as 

approximately 0.3 dpm/L and notes that fecal bioassay was not routinely used for strontium.

5.3.5

The routine bioassay monitoring program at ANL-E has historically been conducted for workers 

(identified by self-reporting, by supervision, and by radiation protection staff) in areas involving 

possible internal exposure. Most workers at ANL-E were not identified to be at risk and were not 

monitored.

Estimation of Missed Dose for Unmonitored Individuals 

5.4 AIR SAMPLING

ANL-E monitored air in work areas with risk of airborne radionuclides from the onset of operations, 

and there are many archived boxes of air-sampling records. The Laboratory did not use air-sampling 

data to assign individual doses but only to provide a means for assessing the adequacy of control 

methods. Elevated air sample results might have been used to trigger requests for special bioassay 

(urine or fecal) samples from potentially affected workers. The extent of this practice was not readily 

discernable from the available documentation. In general, workers in areas with an active airmonitoring program would have been on a routine bioassay monitoring schedule.

ANL-E controlled the airborne radionuclides based on the maximum permissible concentration (MPC), 

also called “tolerance.” (Nickson 1946) stated that, for alpha emitters, the MPC in air was 70 dpm/m3

(3 x 10-11 μCi/mL). No maximum permissible concentration was stated for beta emitters in 1946. 

Annotations on air sample worksheets in 1956 and 1959 indicate that the MPC for alpha emitters was 

still 70 dpm/m3

, and the MPC value for beta emitters was 2200 dpm/m3 (1 x 10-9 μCi/mL). Analysis of 

air sampling work sheets from 1954 also indicate a beta MPC of 2200 dpm/m3

. 

The usefulness of air-monitoring data for dose reconstructions would be very limited because there is 

insufficient information available to correlate area air sample data with the location of a worker over 

time.
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5.5 BIOASSAY REPORTS

This section shows and discusses examples of reports of bioassay data that the dose reconstructor 

might encounter in worker files. Personal identifiers have been removed from these examples. 

Figure 5-1 is an example of the card used to record bioassay data starting in 1946 for monitored 

workers. The header information on the card is as follows. The F and M boxes indicated the gender 

of the worker. H and MO indicated the salary classification of the worker (hourly or monthly). A and T 

referred to the type of work (administrative or technical). The worker’s job title, division, and location 

were recorded. When any of that information changed the old data was lined or crossed out and the 

new data were typed in, but the date of the change was not recorded. The Schedule codes W, Q, A, 

O, M, and S referred to the bioassay schedule frequency (weekly, quarterly, annually, other, monthly, 

or semiannually). If the schedule frequency changed the previous checked mark was crossed out, but 

the date of the change was not recorded. However, dose reconstructors should examine the 

recorded sample dates to determine the actual frequency. The other header information is selfexplanatory.

Figure 5-1. Early bioassay data card. 

The main body of the card contains the rows, generally one per analyte, for the results of the bioassay 

analysis. One sample could have had more than one analyte. The column headings are generally 

self-explanatory. The Element column could indicate an analysis type (e.g., α, β, or γ for gross alpha, 

gross beta, and gross gamma) or a specific radionuclide (e.g., Pu, 239Pu, U, and 3

H). The result 

column headed by “Estimated d/m/1,500 cc” consists of two columns. The first column records the 

numerical value but not necessarily the magnitude; the second column is the exponent of the order of 

magnitude (e.g., 0 is 100 or 1, 3 is 103 or 1,000). The numerical value in the first column is multiplied 

by the magnitude indicated in the second column.

The Remarks field could contain supplemental information to help the dose reconstructor interpret the 

results especially in the case of special samples. 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 31 of 42

Figure 5-2 is another example of the card used to record bioassay data starting in 1946 for monitored 

workers with an accompanying handwritten record of count data for some of the samples. In this 

example, the count data are not presented for all samples. 

The volume recorded as part of the count data is not the same as the volume recorded in the Volume 

Anal. (c.c.) column for these samples. The count data recorded under Net cts/min are all positive 

results, but the recorded results are 0 d/m/1,500 cm3

. No conversion factor has been discerned yet to 

allow the dose reconstructor to calculate the activity in disintegrations per minute from the counts per 

minute data. The dose reconstructor might also note whether the data for all the samples were 

recorded on the card. In this case, the result for the September 28, 1953, sample 53 was not 

recorded on the card.

Figure 5-2. Early bioassay card with supplemental count data. 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 32 of 42

Figure 5-3 is an example of the data card in the 1960s; it includes examples of gross alpha, gross 

beta, and gross gamma analyses as well as results of special uranium analyses for urine and fecal 

samples. The gross alpha results for routine urine samples are all zero in this example. The gross 

beta results are all positive but with the note that the value includes 40K. The dose reconstructor 

should remember the discussion preceding Table 5-7 that the normal range for beta bioassay 

measurements was 2,000 to 10,000 dpm/1,500 mL. In addition for the potassium measurements for 

some of the gross beta samples (all cases in this example), the 40K result is greater than the gross 

beta result. The gross gamma results were all cited as “Normal.” No DL for gross gamma 

measurements in this era has been discovered.

Figure 5-3. Bioassay card from the 1960s. 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 33 of 42

For the special samples starting June 26, 1968, there was a Request for Bioassay Analysis card as 

shown in Figure 5-4. This card might contain information about the radionuclides and the 

circumstances leading to the request for a special or nonroutine bioassay sample.

Figure 5-4. Request for Bioassay Analysis card. 

Figure 5-5 is an example of the bioassay report starting in the mid-1970s. Unlike the previous cards, 

this report lists only the minimum result data without supplemental information. However, for 

radionuclide-specific analyses, the reporting (less-than) levels are stated and quantified. The Type is 

the type of sample (U for urine and F for fecal). The Size is the volume in milliliters of the urine 

sample or weight in grams of the fecal sample. The analyte ALP indicates gross alpha measurements 

with proportional counting, and GAM indicates gamma-emitting isotopes measured by gamma 

spectroscopy. The result is stated as DPM, but it is not clear whether it is per sample, per milliliter, 

per liter, or per 1,500 mL. Dose reconstructors should assume a continuation of the practice for the 

bioassay data cards of disintegrations per minute per 1,500 mL for urine samples (except for tritium) 

and disintegrations per minute per sample for fecal samples. For tritium, results in this era appear to 

be reported in disintegrations per minute per milliliter with a reporting level on a 1977 report of 

2.5 x 104 dpm/mL. 

Figure 5-6 shows the next evolution of the bioassay reports in about 1986. These reports show the 

radionuclide- or isotope-specific results along with the value of the standard deviation. It has not been 

determined from available documentation whether all analytical uncertainties were included in this 

value of the standard deviation. The units of the results are clearly stated as disintegrations per 

minute per liter for urine sample results and disintegrations per minute per sample for fecal results.

Some bioassay data reports starting in the 1990s list a standard set of radionuclides but show data 

only for some of the radionuclides; Figure 5-7 is an example. The key indicator is the zero value for 

the standard deviation. If the standard deviation is zero, the dose reconstructor should consider that 

no analysis was performed for that radionuclide.
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 34 of 42

Figure 5-5. Bioassay report from the mid-1970s. 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 35 of 42

Figure 5-6. Bioassay data report starting about 1986. 

Figure 5-7. Bioassay report with partial data.. 
Document No. ORAUT-TKBS-0036-5 Revision No. 00 Effective Date: 03/01/2006 Page 36 of 42

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Illinois, May 1.

ANL (Argonne National Laboratory), 1975, “Bioassay,” Chapter V-13 in Health and Safety Manual,

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ANL (Argonne National Laboratory), 1990, IDWBC-001, Argonne, Illinois, August.

ANSI (American National Standards Institute), 1996, Performance Criteria for Radiobioassay, 

HPS N 13.30, McLean, Virginia. 

Author unknown, 1958, “Analysis Results of Urine Samples Submitted by Great Lakes Carbon Co. 

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Industrial Hygiene and Safety, Argonne National Laboratory, Argonne, Illinois 

Bertelli, L., 2003, Technical Basis Document for Internal Dosimetry, Rev. 00, Argonne National 

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DOE (U.S. Department of Energy), 1999, Internal Dosimetry Program Guide, DOE G 441.1-3, 

Washington, D.C. Formerly G-10 CFR 835/C1.

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Washington, D.C.

Duffy, T. L.,1958. Memo to J. R. Novak. Urinary Uranium Results (for Armour Research Foundation) 

October 29, 1958.

Fairman, W. D. 1975. Bioassay Program Reports from January 1974-December 1974. Argonne 

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Fairman, W. D. 1976. Bioassay Program Reports from January 1975 – December 1975. Argonne 

National Laboratory. Argonne, Illinois.

Fairman, W. D., 1977, Bioassay Program Reports from January 1976-November 1976, Argonne 

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Fairman, W. D. 1978. Bioassay Program Reports from January 1977-November 1977. Argonne 

National Laboratory, Argonne, Illinois.

Fleury, E. R., 1947, No title, letter to R. S. Stone (U.S. Atomic Energy Commission), June 12, 1947. 

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GLOSSARY

accelerator

Device for imparting kinetic energy to charged particles.

activation products

Elements present in the structure of the reactor or the fuel rods that become activated by 

neutrons, as opposed to fission products that result from fission.

bioassay

Measurement of amount or concentration of material (usually radioactive material) in the body 

or in biological material excreted or removed from the body and analyzed for purposes of 

estimating the quantity of material in the body.

boiling-water reactor

Nuclear reactor in which the coolant, water, is permitted to boil as it absorbs the heat of the 

nuclear reaction. 

breeder reactor

Nuclear reactor in which the operation produces a net increase in fissionable atoms.

committed

In relation to dose equivalent, refers to a total or time-integrated amount for 50 years after 

intake or onset of a chronic intake (or for a different period if so specified).

committed dose equivalent

Dose equivalent calculated to be received by an organ or tissue over a 50-year period 

following an acute intake of a radionuclide into the body or onset of chronic intake. It does not 

include contributions from radiation sources external to the body.

committed effective dose equivalent (CEDE) 

Sum of the effective dose equivalents to various tissues or organs in the body each multiplied 

by the appropriate tissue weighting factor and committed for a 50-year period following an 

acute intake or onset of chronic intake. It does not include contributions from external dose.

control rod

Device manipulated within a nuclear reactor constructed of material to absorb neutrons for the 

purpose of slowing or increasing the nuclear reaction.

criticality

A sustained nuclear chain reaction (can be controlled or not). 

curie

Special unit of radioactivity equal to 3.7x1010 disintegrations per second. 

cyclotron

Accelerator capable of large beam currents where the beam is injected in the center of a 

circular magnet. A fixed radio frequency signal applied to two D-shaped electrodes 

accelerates the beam as it passes from one electrode to the other as the potential alternates. 

The radius of the beam increases as the energy increases.
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decision level (Dc)

The value of a net observation (result) at or above which a decision is made that a positive 

quantity of analyte is present.

decommissioning 

Removal of a facility from service, usually involving decontamination of radioactivity to 

specified levels and often involving demolition of the facility. 

decontamination 

Reduction or removal of radioactive material from a structure, area, object, or person. 

Decontamination can occur through (1) treating the surface to remove or decrease the 

contamination or (2) allowing natural radioactive decay to occur over time.

detection limit (DL)

A general term relating to the smallest amount of material detectable as a function of the 

measurement method and instrument background.

disintegrations per minute

Rate of radioactivity decay in a sample.

dose

Specific amount of energy from ionizing radiation or a toxic substance absorbed per unit of 

mass.

Dynamitron 

Direct current accelerator manufactured by Radiation Dynamics. The terminal is charged with 

a many stage rectifier system excited by radio frequency. 

enriched uranium

Uranium enhanced from its natural state to contain a higher abundance of the isotope 235U. 

exposure

(1) A measure of X- or gamma-ray radiation capable of ionizing air in units of roentgen. (2) 

Qualitatively used to describe the time and concentration of radionuclides to which a worker is 

exposed. 

fission

Nuclear transformation characterized by the splitting of a nucleus into at least two other nuclei 

and the release of a relatively large amount of energy.

fission product

Elements or compounds resulting from fission.

hot cell

Specialized shielded laboratory in which radioactive materials can be handled with the aid of 

remotely operated manipulators. The walls and windows of the laboratory are made of 

materials designed to protect workers from radiation. 

intake

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

ingestion or through wounds.
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in vitro 

Of or relating to a process that takes place under artificial conditions or outside a living 

organism (e.g., in the laboratory). From Latin meaning in glass. 

in vivo 

Of or relating to a process that takes place within a living organism. From Latin meaning in 

life. 

linear accelerator

Single-pass accelerator

minimum detectable amount or activity (MDA) 

Smallest amount or activity of a radionuclide in a sample (or organ) that will yield a result 

above the detection level with a specific probability of a Type II (false negative) error while 

accepting a specific probability of a Type I (false positive) error. 

minimum detectable concentration (MDC)

MDA expressed in units of concentration (e.g., dpm/L). 

neutron

Basic nuclear particle that is electrically neutral with nearly the same mass as a hydrogen 

atom.

neutron, fast

Neutrons with energy equal or greater than 10 kilo electron volts. 

Phoswich detector

A thin NaI detector combined with a thicker CsI detector which provides anticoincidence 

discrimination based on pulse rise time against higher energy photons which will excite both 

detectors. This reduces the low energy background enhancing detection of plutonium xrays.

positive level

Argonne Internal Dosimetry Program term for detection level: A level of a bioassay 

measurement at which the analyte to be detected (as opposed to being detectable). 

radioactivity

Spontaneous emission of radiation, generally alpha or beta particles, gamma- and X-rays, and 

neutrons from unstable nuclei.

radioactive waste

Byproducts of nuclear processes that are radioactive and have no useful recyclable purpose 

(see nuclear waste).

radionuclide

Radioactive species of an atom characterized by the constitution of its nucleus specified by the 

number of protons, neutrons, atomic number, and mass number. 

rem

Unit of dose equivalent equal to the product of the rad absorbed and the quality factor.
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synchrotron

Roughly circular accelerator in which the particles travel in synchronized bunches at fixed 

radius.

tandem Van de Graaff or Dynamitron

Accelerator in which a charge exchange occurs in the terminal on either a thin foil or in a gas 

stripper tube.

tritium (H-3, 3

H)

Radioactive isotope of hydrogen having one proton and two neutrons. Tritium gas is produced 

in nuclear reactors and used to boost the explosive power of most modern nuclear weapons. 

It is also a constituent of irradiated water associated with reactor operations.

Van de Graaff accelerator 

Accelerator in which constant high-voltage on a terminal in a pressurized tank accelerates the 

beam. The voltage is generated by charge carried on an insulating belt.

water-moderated reactor

Reactor in which water slows the speed of neutrons from fissioning atoms. 

X-ray

(1) Ionizing electromagnetic radiation of external nuclear origin with energies less than 

250 kilo electron volts. (2) Radiograph.

zero power

Operating a reactor to maintain a chain reaction at an extremely low power level producing 

very little heat. Zero power reactors are used as sensitive laboratory tools to pretest 

experimental loadings of test reactors and for other analytical purposes. Also called low 

power.