ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 29

Document Title:

Argonne National Laboratory – West – 

Occupational Internal Dose

Document Number: ORAUT-TKBS-0026-5 

Revision: 01 

Effective Date: 05/23/2007

Type of Document TBD

Supersedes: Revision 00

Subject Expert(s): Douglas Wenzel

Site Expert(s): Norman D. Rohrig

Approval: Signature on File Approval Date: 05/11/2007

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 05/11/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 05/11/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 05/17/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 05/23/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 2 of 29

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

01/20/2005 00 New technical basis document for the Argonne National Laboratory –

West – Occupational Internal Dose. First approved issue. Initiated 

by Norman D. Rohrig.

05/23/2007 01 Approved Revision 01 initiated to include Attributions and 

Annotations. Incorporates formal internal review comments. This 

revision results in no change to the assigned dose and no PER is 

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

Initiated by Jo Ann M. Jenkins.
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 3 of 29

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

5.1 Introduction ..............................................................................................................................8

5.1.1 Purpose........................................................................................................................9

5.1.2 Scope...........................................................................................................................9

5.1.3 Argonne National Laboratory–West............................................................................10

5.1.4 Radioactive Nuclides of Concern and Solubility..........................................................10

5.2 Radiological Protection Program Management and Support .................................................. 10

5.2.1 Bioassay Programs ....................................................................................................11

5.2.2 Internal Dose Records................................................................................................11

5.2.3 Data Codes and Investigation Levels..........................................................................11

5.3 Internal Dose Control ............................................................................................................. 14

5.3.1 Air Monitoring .............................................................................................................15

5.3.2 Early Technical and Analytical Capabilities................................................................. 15

5.3.3 Recent History of Bioassay Results............................................................................16

5.4 Minimum Detectable Activities ............................................................................................... 16

5.5 Development and Impact of Whole-Body Counting ................................................................ 17

5.6 Default for Missed Dose......................................................................................................... 21

5.7 Unmonitored Workers ............................................................................................................ 22

5.8 Attributions and Annotations .................................................................................................. 22

References ........................................................................................................................................23

Glossary ............................................................................................................................................27
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LIST OF TABLES

TABLE TITLE

5-1 Radionuclides of concern.......................................................................................................

PAGE

11

5-2 Routine bioassay history summary......................................................................................... 12

5-3 Internal dose assessment information after mid-1990 ............................................................ 13

5-4 Internal dose assessment information before mid-1990 ......................................................... 13

5-5 Analytical information that could be in claimant dose files ...................................................... 13

5-6 Area codes that could be in claimant dose files...................................................................... 14

5-7 Surface contamination control and MDAs .............................................................................. 15

5-8 Recent history of bioassay sampling ...................................................................................... 17

5-9 MDAs for urine samples by period ......................................................................................... 18

5-10 MDAs for fecal samples by period.......................................................................................... 19

5-11 In vivo MDAs by period .......................................................................................................... 19

5-12 Default table for missed dose................................................................................................. 21
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 5 of 29

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AECL administrative exposure control level

AEDE annual effective dose equivalent

ANL Argonne National Laboratory

ANL-W Argonne National Laboratory–West

ANPP Aircraft Nuclear Propulsion Program

ANSI American National Standards Institute

CAM continuous air monitor

CDE committed dose equivalent

CEDE committed effective dose equivalent

CFA Central Facilities Area

CFR Code of Federal Regulations

cm centimeter

COO Chicago Operations Office

CPP Chemical Processing Plant

cpm counts per minute

DAC derived air concentration

DE dose equivalent

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

EBR Experimental Breeder Reactor

ECF Expended Core Facility

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ERDA U.S. Energy Research and Development Administration

F fast (absorption type) 

FCF fuel cycle facility

g gram

GCRE Gas-Cooled Reactor Experiment

GE General Electric Company

hr hour

HFEF Hot Fuel Examination Facility

HSD Health and Safety Division

HSL Health Services Laboratory

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

IDO Idaho Operations Office

in. inch

INL Idaho National Laboratory

L liter

LPTF Low Power Test Facility
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 6 of 29

M moderate (absorption type) 

MDA minimum detectable activity

MFP mixed fission product 

mi mile

min minute

mL milliliter

ML-1 Mobile Low-Power Reactor No. 1

MPBB maximum permissible body burden

MPC maximum permissible concentration

mrem millirem

mrep millirep

MTR Materials Test Reactor

NaI(Tl) sodium iodide doped with thallium

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NRF Naval Reactors Facility

NRTS National Reactor Testing Station

PBF Power Burst Facility

pCi picocurie

POC probability of causation

RDR Radiation Dosimetry and Records

RWMC Radioactive Waste Management Complex

S slow (absorption type) 

SL-1 Stationary Low-Power Reactor No. 1

SMC Specific Manufacturing Capability

SPERT Special Power Excursion Reactor Test

STEP Safety Test Engineering Program

STPF Shield Test Pool Facility

TAN Test Area North

TBD technical basis document

TRA Test Reactor Area

TREAT Transient Reactor Experiment and Test (facility)

TRU transuranic

U.S.C. United States Code

WBC whole-body counting

yr year

ZPPR Zero Power Plutonium (later Physics) Reactor

α alpha particle

β beta particle 
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 7 of 29

γ gamma

σ standard deviation

μCi microcurie

μg microgram

§ section or sections
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 8 of 29

5.1 INTRODUCTION 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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 occupationally derived radiation 

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

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

external dosimetry monitoring 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. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] 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.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 9 of 29

5.1.1

This TBD discusses Argonne National Laboratory–West (ANL-W) internal dosimetry data for dose 

reconstruction and includes guidance for the appropriate use of that information.

Purpose

5.1.2

In 1949, the U.S. Atomic Energy Commission (AEC) established the National Reactor Testing Station 

(NRTS) and started construction of facilities on a 572,000-acre site approximately 50 mi west of Idaho 

Falls in southeastern Idaho. NRTS was later renamed the Idaho National Engineering Laboratory, 

then the Idaho National Engineering and Environmental Laboratory, and most recently the Idaho 

National Laboratory (INL). Argonne National Laboratory (ANL) near Chicago established a branch at 

the NRTS known as ANL-W where it built and operated several reactors that were of fundamental 

importance for the development of commercial nuclear power. ANL-W was operated under contract 

to the Chicago Field Office of the AEC/DOE by the University of Chicago from 1951 through January 

2005. ANL-W was merged with INL in February 2005, and Battelle Energy Alliance assumed all 

operations. 

Scope

Each of the original AEC laboratories was unique in both mission and location. Since the early days 

of the AEC programs, which represent the beginnings of the nuclear age, significant technical 

developments were a necessity, not the least of which were developments in radiation safety. Some 

of the unique characteristics of radiation safety (and internal dosimetry specifically) at ANL-W, which 

had a marked influence on the conduct of the internal dosimetry programs are as follows:

• The original mission of the ANL-W was uranium reactor concept development. The production 

of weapons-grade nuclear materials was not a mission.

• The beginning of ANL-W activities was 8 to 10 years later than those at ANL-East, Oak Ridge 

National Laboratory, and the Hanford Site. During those developmental years significant 

technical progress in professional skills, instrumentation, analyses, procedures, and 

techniques were accomplished. 

• The Chicago Field Office was responsible for and had oversight of the ANL-W program. 

• To provide necessary consistency of radiation safety programs at the NRTS, the AEC through 

the Idaho Operations Office (IDO) established a Health and Safety Laboratory (HSL) to 

provide technical support in the areas of (1) environmental surveillance, (2) external dosimetry 

(personnel dosimeters of all types), (3) portable radiation detection instrumentation 

inventories, calibration, and maintenance, (4) internal in vitro and in vivo bioassay analytical 

laboratories, (5) maintenance and documentation of personnel dosimetry records, and 

(6) research and development in these areas of responsibility. The name of this organization 

changed several times to Health and Safety Division (HSD), Idaho Center for Radiological and 

Environmental Sciences, and most recently to the Radiological and Environmental Sciences 

Laboratory. Technical data, dosimetry services, information (particularly in the instances of 

detectable worker intake), and analytical internal dose calculations and evaluations were 

exchanged between the AEC HSL and ANL-W.

As a consequence, basic assumptions about minimum detectable activities (MDAs) or minimum 

detectable levels, missed dose potential, etc. are relatively consistent across the years. As early as 

1955 or 1956 gamma spectral analysis capabilities allowed the significant bioassay results (those that 

would result in reportable internal dose) to be defined in terms of the specific radionuclides. The 
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 10 of 29

practice in the case of a higher urine sample result was to attempt radionuclide identification through 

gamma spectral analysis and chemical separation. This document describes default assumptions for 

use in cases when the bioassay records for a claimant do not include specific radionuclide analyses 

but rather only record gross beta or gross alpha results.

The majority of this document provides background information to aid the internal dose reconstructor 

through increased general understanding, data interpretation, defaults, etc. Section 5.1.3 provides a 

facility description, and Section 5.1.4 details the radionuclides of concern. Section 5.2 describes the 

INL radiological protection program as it evolved over the years, Section 5.3 discusses internal dose 

control, and Sections 5.4 and 5.5 describe MDAs and whole-body counting (WBC), respectively. 

Sections 5.6 and 5.7 suggest approaches for the treatment of missed dose and unmonitored workers.

5.1.3

ANL-W, which began operations in 1951, continued to conduct nuclear power developmental 

programs through January 2005. Although ANL-W received internal dosimetry support from the INL

service laboratories, its radiological safety programs operated under authority of the Chicago 

Operations Office (COO). 

Argonne National Laboratory–West

Nine experimental reactors under the technical direction of ANL-W were operated at two locations, 

one on the southwest side of the Site near the Radioactive Waste Management Complex (RWMC) 

and the others at the current location on the southeast side of the site. Early reactor operations 

included physics critical experiments; power production; routine unmoderated operation; uraniumfueled, plutonium-fueled, and breeder reactor designs; and self-destruction experiments. 

A series of deliberate safety experiments were conducted by ANL-W in which reactors were allowed 

to go prompt-critical with resultant reactor destruction (Stacy 2000). External and internal doses to 

workers, both expected and accidental, were associated with these activities (RAC 2002). 

5.1.4

ANL-W facilities and activities related primarily to experimental reactor design and development. The 

latest revision of the ANL-W Site Description TBD (ORAUT 2004) describes these activities in more 

detail. The radionuclides of concern as listed in Table 5-1 from an internal dose standpoint are mixed 

fission products (MFPs) (primarily aged), uranium and decay products, and transuranic (TRU) and 

decay products. The uranium had various enrichments from depleted to at least 68% enriched, both 

used in Experimental Breeder Reactor No. 2 (EBR-II). Most internal doses were identified after an 

incident rather than as a result of routine bioassay measurements. Through the years at ANL-W, 

plotting urine and fecal elimination curves has shown that many nuclides appear to be eliminated 

slowly, which indicates a relatively insoluble material. The chemical explanation is that radioactive 

materials oxidize rapidly and form less soluble compounds. A default assumption of type M or S 

would be appropriate, where International Commission on Radiological Protection (ICRP) Publication 

68 (ICRP 1995) specifies these for the nuclide. Strontium is an exception; type F should be assumed.

Radioactive Nuclides of Concern and Solubility

5.2 RADIOLOGICAL PROTECTION PROGRAM MANAGEMENT AND SUPPORT

The contract with the University of Chicago to operate the ANL-W facilities did not change during the 

54 years of operations. The COO provided oversight for ANL-W programs and facilities. The ANL-W 

program used site support services, including internal dosimetry support.
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Table 5-1. Radionuclides of concern (Nielsen 1996a). 

Radionuclides Sources/characteristics

H-3 (HTO) EBR-II Reactor Facility & Sodium Components Maintenance Shop

Mn-54, Fe-59, Co-58, Co-60 EBR-II primary source. Levels low and decreasing since decommissioning.

Sr/Y-90 All facilities handling fission products

Cs-134, Cs-137 All facilities handling fission products

U-234, 235, 238 Hot cell, hoods, glove boxes, waste, reactor fuel, research areas

Pu-238, 239 FCF, HFEF, ZPPR, Analytical Laboratory

Am-241 FCF, HFEF, ZPPR, Analytical Laboratory

The personnel dosimetry records were documented and are permanently maintained. Records about 

individual facility or contractor field monitoring programs (air-monitoring data, personnel contamination 

records, etc.) were maintained by ANL-W. The field monitoring data were not available for use in this 

report. 

The HSL had radionuclide identification capabilities from the early 1960s. Positive bioassay results 

(analyses in which the results exceeded 2σ counting statistics) were normally followed by a 

confirmatory analysis to identify specific radionuclides. In the cases where only gross beta or gross 

alpha bioassay results are available, the results are normally within 2σ. However, if it is necessary to 

evaluate intakes from the gross beta or gross alpha results, the radionuclide defaults should be 

90Sr/90Y and 239Pu, respectively. 

5.2.1

Routine bioassay of radiation workers was conducted from the beginning of operations. However, 

formal documentation of the ANL-W bioassay programs was not found for periods before 1989. 

Some of the data sheets on individuals indicate that urine bioassay sampling occurred annually from 

1959 to 1966. Whole-body counts began in 1963 and were typically annual. Table 5-2 lists the 

reconstructed history of routine bioassay frequency.

Bioassay Programs

5.2.2

Formal or legal internal dose data were maintained by the DOE HSD in individual hard-copy folders 

until mid-1990 when all technical support service functions, including those related to internal 

dosimetry, were transferred to the INL prime contractor. At that time, in vitro analytical functions were 

transferred to an onsite INL analytical laboratory. The in vivo counting laboratory provided support 

directly through the Radiation Dosimetry and Records (RDR) organization, which administered

external and internal dosimetry support programs. The subject matter expert reviewed, validated, and 

prepared official internal dose assessments. The RDR unit functions included documentation and 

records custodial responsibilities.

Internal Dose Records

5.2.3

The changes in standards and regulations influenced the level of internal dose evaluation and 

documentation, but did not change the fact that all (negative as well as positive) bioassay data were 

recorded in the individual dosimetry files.

Data Codes and Investigation Levels 

The information that was used in internal dose assessments and analytical data sheets varied through

the years. Table 5-3 describes coded information that could appear in records from after mid-1990. 

Table 5-4 describes internal dose information that could appear in records from before mid-1990. 

Table 5-5 contains analytical nomenclature. Table 5-6 contains INL codes for various site areas.
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 12 of 29

Table 5-2. Routine bioassay history summary. 

Year

Typical 

frequency Type

Groups 

analyzed/sampled Investigating level Comments Reference

1953–

1960

Annually In vitro

urine

Radiation workers Frequency is 

inferred from 

individual 

data sheets.

Individual data 

sheets

Table 5-9 

AEC 1959

AEC 1961

1961–

1972

Annually In vitro 

urine 

In vivo

Radiation workers Frequency is 

inferred from 

individual 

data sheets.

AEC 1962

AEC 1963

1973–

1981

When intake 

suspected

Fecal Radiation workers

Annual DE >10% 

quarterly standard in 

ERDA Manual Chapter 

0524

Reporting Frequency is 

inferred from 

individual 

data sheets.

AEC 1968 

AEC 1975

Annually In vivo ERDA 1975

1982–

1988

When intake 

suspected

Fecal Radiation workers

CDE >10% quarterly 

standard in DOE 

5480.1A Chapter 11

Reporting DOE 1981

Annual In vivo

Termination In vivo Radiation workers

1989–

1995

Annual In vivo Radiation workers, 

available all employees

ES&H 1989

Radiation 

Workers 

based on 

exposure 

potential

In vitro 

fecal

Medium to highest risk 

working with plutonium 

contamination

In vitro 

urine

Uranium or tritium 

workers

Termination In vivo When determined by 

Radiation Fire and 

Safety Engineering

In vitro

New hire In vitro

In vivo

1995–

2004

Appropriate 

to the facility 

mission, 

potential 

uptakes.

In vivo Radiation workers that 

enter radiological buffer 

areas or areas of greater 

radiological controls and 

are likely to receive 

intakes resulting in a 

CEDE of 0.1 rem or 

more. Type of bioassay 

based on source term. 

Urine requested when 

pure beta, uranium, or 

TRU was of interest. 

Feces requested 

primarily for uranium and 

TRU source terms.

In accordance with 

DOE 5480.11 and 

10 CFR 835. 

Reporting

Workers that could 

receive 0.1 rem 

CEDE.

Declared pregnant 

workers when 

embryo/fetus could 

receive 0.05 rem DE. 

Follow-up for 

any 

suspected 

intake of 

radionuclide

s and to 

more 

accurately 

identify and 

characterize 

the amount 

of intake and 

excretion 

pattern.

Nielson 1996a

In vitro DOE 1988

Urine

Fecal

When work 

place 

monitoring 

indicates 

significant 

potential for 

intakes.

Internal doses resulting 

from all confirmed 

intakes are to be 

evaluated.

Investigating

1995–

2004

New hire To determine internal 

conditions from previous 

uptakes or to establish 

baseline for those 

continuing to work as 

radiation workers.

Termination In vivo All employees

CDE = committed dose equivalent; CEDE = committed effective dose equivalent; DE = dose equivalent; ERDA = U.S. Energy 

Research and Development Administration.
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 13 of 29

Table 5-3. Internal dose assessment information after mid-1990. 

Coded information Description

Name & SS no. Exposed employee by name and Social Security Number.

Asmt. nos. This assessment number is the calendar year, and a consecutive numbered 

assessment for that employee during that specific year.

Intake date Month/day/year of employee intake.

Radionuclide class & 

amt.

Specific radionuclide followed immediately by ICRP Publication 30 solubility class 

symbol D, W, or Y (ICRP 1979). Amount in microcuries or becquerels.

CEDE rem Calculated CEDE in rem.

Organ (max.) Organ that received the maximum dose from the specified intake.

Organ CDE rem CDE calculated for the listed organ in rem.

Employer and exp. 

location

Abbreviation of DOE site contractor and the plant site of exposure (can include the 

building number).

Year–total CEDE CEDE exposures are summed for the year of intake for each employee.

Year–TL organ CDE Organ CDE total (TL) exposures are summed for the year of intake for each 

employee.

Table 5-4 Internal dose assessment information before mid-1990. 

Dose information Description

Name, Soc. Sec. no. Employee name, Social Security Number, and (Contractor Abbreviation/Plant or 

Facility).

Nuclide Radionuclide symbol followed by ICRP solubility class (D, W, or Y) (ICRP 1979).

Intake period Month and year for single exposure or period of time by month and year in which 

exposure occurred.

CEDE rem Calculated CEDE in rem.

AEDE rem Calculated AEDE in rem.

Year Year for which the AEDE was calculated.

AEDE = annual effective dose equivalent.

Table 5-5. Analytical information that could be in claimant dose files. 

Analytical information Description

Sample no. Sample log number.

Date and time Generally clear interpretation.

Sample description Name of the employee, numerical sample number frequently included, additional 

special analyses performed (e.g., Sr-90, Y Separation, etc.). *

Anal. for Generally gross beta and/or gross gamma. Sample aliquots evaporated for gross 

beta or counted directly in a deep-well NaI scintillation counter. Specific isotopic 

analysis based upon chemical separation or gamma spectrum also listed in this 

column.

Quantity used Size of the sample aliquot, generally in mL.

U+ or K+ trans. Note to indicate analytical correction for natural potassium and uranium.

Count time Counting either used preset time or preset counts. Time in minutes recorded in 

either case.

Total count Total number of counts recorded.

Gross count, cpm Cpm determined by dividing total counts by time of count.

Bkgd., cpm Background cpm recorded.

Net count, cpm Gross cpm minus background cpm.

K-40 corr., cpm Additional background from K-40 identified. K-40 is not a facility occupational 

exposure product; ignore from an internal dose reconstruction

Foreign activity, cpm 

and dpm

Net counts corrected for K-40 and then converted to dpm based upon counter 

calibration. Uncertainty also included, which is recorded as 1σ based upon 

counting statistics.

Dpm per a volume The activity is for the sample volume listed.

Result in μg/L These results are for uranium whether stated or not.
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 14 of 29

Table 5-6. Area codes that could be in claimant dose files. 

Area code Description Area code Description

1 AEC Headquarters Building 20, 261, 264 TREAT

2 EBR-I 21 LX

3, 034, 035 CFA 22 GCRE

4, 042,045 MTR, TRA 23 OX

5, 053, 055 ICPP 24 ARHG

6 NRF 25 No information available

7 TAN (GE) 26, 263,265 EBR-II

8 Services 27 ML-1

9 NX (X is construction) at NRF 28 Onsite survey

10 AX at TAN 29 Offsite survey

11, 113 CX at CPP 30 ANPP at SL-1

12 EX at EBR 31 STPF

13, 133,135 SPERT, PBF 65 ECF

14 MORE 66 Non-Security

15 SX at SPERT 67 Division of Compliance

16 SL-1 68 STEP

17, 333 MX at MTR 69 LPTF (Phillips & AEC)

18, 814,815 WP, RWMC 71 CADRE

19, 772, 775 TAN (Phillips & AEC) 774,776 SMC

Bold indicates ANL-W facilities.

Each individual analytical result was documented and placed in individual exposure files, regardless 

of the formal reporting requirements. The investigation levels (the levels at which positive bioassay 

results triggered follow-up sampling to verify that detectable activity had been taken into the body) 

also changed little from the early years to 2005. Later procedures (DOE 1988) set specific limits on 

those positive bioassay results that could result in an AEDE of 100 mrem or above as the point at 

which follow-up and reporting was required. With the DOE Radiological Control Manual (DOE 1994), 

this changed to a CEDE of 100 mrem. In addition, a calculated dose of 10 mrem or above would be 

recorded as an internal dose. These procedural limits did not materially affect the bioassay sampling 

frequency or the recording of even undetectable radioactivity in bioassay samples, although the 

request for and number of follow-up samples and analyses could have been different as a function of 

the formal regulations in effect.

5.3 INTERNAL DOSE CONTROL

The radiological protection program was designed to detect barrier or ventilation failure in a timely 

manner. The program consisted of continuous and retrospective air and effluent monitoring combined 

with personnel and surface contamination monitoring. Detection of barrier failure provided the 

information for making decisions on personnel evacuation, increasing personal protection equipment 

(e.g., respirators), and requesting bioassay analyses to identify possible internal intakes. 

The contamination control limits for the detection and control of released activity were dependent on 

instrumentation capabilities. As a result of increased emphasis on exposures that were as low as 

reasonably achievable, some reduction in acceptable release levels was implemented. The 

contamination control limits for alpha on plant surfaces and particularly on personnel were set close to 

the MDA, such that any detectable contamination was a signal for preventive action and follow-up 

evaluations. Beta/gamma MDAs typically were a factor of 5 below the limits. Table 5-7 is a summary 

of control limits primarily from the CPP [Chemical Processing Plant] Health Physics Manual (ACC 

1952) and the operating procedures in effect in 2005. 
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Table 5-7. Surface contamination control and MDAs. 

Period Surface location

Detection 

technique Control levels MDA - typical

1952–

1960s

Plant/equipment Smears 500 dpm β & 20 dpm α per 

100 cm2 150 dpm β & 10 dpm α per 100 cm2

Personal clothing Portable survey 

instruments

1500 dpm β & 500 dpm α per 

100 cm2 1,000 dpm β & 500 dpm α per 100 cm2

Personal skin Portable survey 

instruments

Any detectable reported, e.g. 

1,000 dpm β & 500 dpm α per 

100 cm2 

1,000 dpm β & 500 dpm α per 

100 cm2

Shipments Smears/portable 

survey 

instrument

500 dpm β & 20 dpm α per 

100 cm2

 – smears

0.1 mrep/hr β & 500 dpm α

per 100 cm2

150 dpm β & 10 dpm α per 

100 cm2 smears. 0.01 

mrep/hr β & 500 dpm α per 

100 cm2

1970s–

2005

Plant/equipment 

surfaces

Smears 300 dpm β & 20 dpm α per 

100 cm2 30 dpm β & 10 dpm α per 100 cm2

Personnel Portable survey 

instruments

Any detectable reported, e.g. 

300 dpm β & 200 dpm α per 

80-100 cm2 

300 dpm β & 200 dpm α per 

80-100 cm2

5.3.1

The monitoring of radioactivity in the air in occupied areas was a basic element of the internal 

exposure prevention program. Beta/gamma continuous air monitors (CAMs) were used for 

beta/gamma emitters with maximum permissible concentrations (MPCs) or derived air concentrations 

(DACs) above 1 × 10-9 μCi/cm3

. TRU materials and uranium were nearly always well tagged with 

beta/gamma activity that allowed beta/gamma-detecting CAMs to be used to warn of possible alpha 

contamination and possible internal exposures. 

Air Monitoring

In general, workers were asked to submit to bioassay whenever they were in an area where a CAM 

alarm sounded. In addition, the fixed location and retrospective air-sampling system would signal the 

need for bioassay if elevated air sample results were detected.

At ANL-W fixed air sampling heads were used in the Fixed Air Sampling System beginning in 1976 

(Courtney et al. 1980). Concentrations averaged between 1 × 10-15 and 5 × 10-15 μCi/cm-3 for alpha 

emitters and between 0.4 × 10-13 and 4.0 × 10-13 μCi/cm-3 for beta emitters in the first 3 years of 

operation. This information can be used to assign an intake for an unmonitored worker.

5.3.2

DOE HSL technical reports and annual reports, along with facility memoranda and reports, 

documented the analytical detection capability in the 1950s and 1960s. Radionuclide identification by 

energy spectra was available and used for urine and other bioassay samples. Specific separations 

(e.g., strontium, iodine) were available to quantify the radioactive components of a variety of samples 

of interest.

Early Technical and Analytical Capabilities

In the early days, a gross beta urine bioassay measurement was made on an evaporated aliquot, or a 

gamma count was made directly on a liquid sample, or both. Any detectable activity triggered a 

specific chemical separation analysis (generally for strontium). Early analyses for plutonium generally 

were gross alpha counts on a plutonium separation; later, alpha spectroscopy was used to count and 

better characterize the results.
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 16 of 29

In 1958, the IDO HSD acquired a 256-channel gamma spectrometer with a 3- by 3-in. sodium iodide 

thallium-doped [NaI(Tl)] detector counting system for analyses of gamma-emitting radionuclides. In 

1960, the HSD obtained a 3- by 3-in. well counter for gamma analysis to replace the gross beta 

counting as the routine analytical procedure for urine samples. AEC (1961, p. 59) states, 

“Approximately 1.5 × 10-6 μCi/mL of MFPs can be detected in 75 mL of urine in a 5-minute count, 

which is about the same as was obtained with the gross beta procedure in a 20-minute count.” 

AEC (1961) outlined a basic philosophy in relation to gamma counting of bioassay samples. Gamma 

counting would be effective in all situations except for exposure to pure strontium isotopes. To guard 

against this unlikely possibility, the procedure of performing a strontium analysis for individual workers 

at risk (radiation workers) every 2 years and at termination was established. Because of the 

improbability of finding detectable activity, all activities were to be precipitated by oxalic acid in a weak 

acid solution and counted for gross beta, and the strontium-specific analysis not completed unless a 

detectable count was obtained on the precipitate. A 100-mL sample of urine permitted the detection 

of approximately 8 × 10-8 μCi/mL of 90Sr.

The special and routine bioassay sample analyses were performed and documented by the DOE 

analytical laboratory. The individual analytical data were recorded. At ANL-W, there was a personnel 

dose file on each individual that contains results of all samplings and notations on all contamination 

incidents. In all cases, copies of the bioassay results should be in the individual’s dose file in the 

NIOSH-Office of Compensation and Support Claims Tracking System database.

5.3.3

Table 5-8 provides a recent history of bioassay sampling based on annual skin contamination and 

exposure reports (Carr 1990; Holson 1991; Kirchner 1992; Marshall 2002; Nielsen 1992a to 1995b, 

1995b to 2004; Vroman 2001). Blank entries mean that the information was not available. The skin 

contaminations tended to be a precursor for positive bioassay, and the number was countable and 

tracked since at least 1984. The associated activities have not been extremely high. In 1993 there 

were a large number of skin and clothing contaminations in relation the Fuel Cycle Facility (FCF), 

which was being refurbished. Almost all workers received a whole-body count until the late 1990s. In 

1998, an FCF seal tube repair led to 11 positive counts on five people. The highest calculated CEDE 

was 1 mrem. The other positive count was from a medical test about 3 years before. The tritium 

assay was about monthly, so many fewer people were involved and the doses were generally trivial 

because of the high monitoring sensitivity. The number of tritium assays was cut back because of the 

history of not recording anything significant in dose from tritium. The plutonium fecal assay led to 

mostly false positive results for which a subsequent fecal assay failed to show any activity. In 1993 a 

person received a hand wound and was chelated, which resulted in two positive urine counts but no 

positive fecal count (Burke 1993). A CEDE of 100 mrem was assigned. This recent history suggests 

that rather modest internal exposures occurred at ANL-W.

Recent History of Bioassay Results

5.4 MINIMUM DETECTABLE ACTIVITIES

In compliance with the November 1998 DOE requirements (10 CFR Part 835) for the DOE Laboratory 

Accreditation Program (DOELAP), and based on ANSI Standard N13.30, An American National 

Standard, Performance Criteria for Radiobioassay (HPS 1996), the in vitro and in vivo radiobioassay 

laboratories at INL received DOELAP accreditation in February 1998. In 
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 17 of 29

Table 5-8. Recent history of bioassay sampling. 

Year

# Skin 

contam 

events

# Clothing 

contam 

events

Maximum 

beta 

contam # WBCs

Positive 

WBCs

# Fecal 

Pu 

assay

Pos in 

vitroa

# 

Tritium 

assay

Pos 

tritium

Max T 

dose

1984 20 cpm until 

1992

mrem

1985 17

1986 7 dpm after

1987 5 1992

1988 16

1989 18 8,000

1990 10 542 0 44 0

1991 10 12,000 648 0 60 0 287 29 14

1992 15 7 8,000 609 0 69 0 40 8 <1

1993 23 58 500,000 1,082 0 33 2 urine 48 6 <1

1994 7 21 139,000 837 0 70 1 FP 46 14 <1

1995 3 10 784,000 758 0 106 1 FP 106 10 <1

1996 2 11 186,000 273 0 75 1 FP 139 13 <1

1997 5 10 135,000 329 1 76 2 FP 1 P 138 8 <1

1998 5 21 309 12 56 2 P 152 16 <1

1999 7 5 311 1 105 2 FP 365 84 <1

2000 2 4 318 1 136 0 697 50 <1

2001 1 7 254 1 143 1 P 552 73 <1

2002 4 8 212 1 143 3 P 40 7 <1

2003 2 8 222 0 127 1 FP 9 1 <1

a. FP – false positive P - Positive

accordance with this accreditation, MDAs and decision levels at the 95% confidence level are 

performed. Tables 5-9 and 5-10 list the MDAs by period for urine and fecal sample analysis, 

respectively. The earliest reference (Ebersole and Flygare 1957) defines a detection limit as “twice 

the standard deviation of its determination.” The detection limits in early references have therefore 

been multiplied by 2.33 to determine an MDA similar to the values from the American National 

Standards Institute (ANSI). A large majority of the urine samples were single voidings; 24-hr samples 

were used for special sampling purposes (i.e., follow-up samples, primarily to extend the sensitivity). 

If the urine sample aliquot size was stated, the MDA for urine samples was given in both picocuries 

per milliliter and picocuries per sample; if the size was not stated, only the reference value was 

provided. Most bioassay data have units of activity per sample with the units changing from 

disintegrations per minute to microcuries to becquerels and back again. The listed MDAs are those 

for the primary samples. The recommended periods for the MDA values are included in the tables. 

Table 5-11 lists the 2005 in vivo MDAs along with values from historical documents and the 

recommended periods for use of the MDA values.

5.5 DEVELOPMENT AND IMPACT OF WHOLE-BODY COUNTING

WBC was introduced at the INL in 1961. As early as 1961 one of the fundamental conclusions from 

experience with in vivo and in vitro internal dosimetry analytical techniques was the fact that a large 

proportion of the internal exposures was to insoluble materials. Radioactive nuclides (e.g. 125Sb, 

110mAg, 65Zn, and 95Zr/Nb) were detected by an in vivo count but were not detected in the urine. 

Concurrent analyses of fecal and urine excreta demonstrated the main elimination route to be by the 

feces, with so little voided in the urine as to be undetectable even in a 24-hour specimen (AEC 1962; 

Sill, Anderson, and Percival 1964). WBC was demonstrated to detect activity as low as 0.01 μCi in a 

10-minute count (AEC 1962). This detection level was several orders of magnitude more sensitive 

than the maximum permissible body burdens (MPBBs) for most beta/gamma fission and activation 
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 18 of 29

Table 5-9. MDAs for urine samples by period. 

Radiation/

radionuclide Period 

Urine

(pCi/mL)

Urine

(pCi/sample)

Nominal sample 

volume (mL) Reference

Gross β 1951–1953 8.4 Data sheet

1954–1959 9.3 46 5 Ebersole & Flygare 1957a

1960–70 3.5 AEC 1961a

1971–2005 1 5 5 AEC 1972a

Gross γ 1960–1964 3.5 260 75 AEC 1961a

1965–1971 2.3 Data Sheet

1972–2005 4.6 345 75 AEC 1972a

H-3 1972–1994 35 105 3 AEC 1972a , AEC 1974

1995–2005 1.4 Andersen, Perry, and Ruhter 

1995; Rielly 2001

Co-60 1963–2005 0.01 Rich 1990

Sr-90 1953–1959 9.2 Ebersole and Flygare 1957

1960–1970 0.2 20 100 AEC 1961

1971–1989 0.023 1.7 75 AEC 1972a, AEC 1974a

1990–1994 0.01 Rich 1990

1995–2005 1.9 Andersen, Perry, and Ruhter 

1995

I-131 1963–2005 0.01 Rich 1990

Cs-134 1963–2005 0.01 Rich 1990

Cs-137 1963–2005 0.01 Rich 1990

Th-230 1974–2005 4.6 E-5 0.046 1000 AEC 1974 a

Np-237 1974–2005 4.6 E-5 0.046 1000 AEC 1974 a

U (FP) 1954–1985 14 μg/L Assumed value

U (KPA) 1985–2005 0.2 μg/L Rich 1990

U-233/234 1970–1994 1.E-4 Rich 1990

1995–2005 0.041 Andersen, Perry, and Ruhter 

1995

U-235 1970–1994 1.E-4 Rich 1990

1995–2005 0.038 Andersen, Perry, and Ruhter 

1995

U-238 1970–1994 1.E-4 Rich 1990

1995–2005 0.030 Andersen, Perry, and Ruhter 

1995

Pu-238 1974–1989 2.3 E-5 0.023 1000 AEC 1974 a

1990–1994 6 E-5 Rich 1990

1995–2005 0.022 Andersen, Perry, and Ruhter 

1995

Pu-239/240 1964–1970 4.E-4 AEC 1964a

1971–1973 5.E-4 0.5 1000 AEC 1972a

1974–1989 2 E-5 0.02 1000 AEC 1974 a

1990–1994 6.E-5 Rich 1990

1995–2005 0.027 Andersen, Perry, and Ruhter 

1995

Am-241 1974–1989 7 E-5 0.07 1000 AEC 1974 a

1990–1994 2. E-4 Rich 1990

1995–2005 0.023 Andersen, Perry, and Ruhter 

1995

Cm-244 1974–2005 1.4 E-5 0.014 1000 AEC 1974 a

Cf-252 1974–2005 1.4 E-5 0.014 1000 AEC 1974 a

a. MDA calculated from inferred 2σ uncertainty.
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 19 of 29

Table 5-10. MDAs for fecal samples by period. 

Radiation/

radionuclide Period

Fecala

(pCi/sample) Reference

Co-60 1963–2005 10 Rich 1990

Sr-90 1963–1994 10 Rich 1990

1995–2005 1.9 Andersen, Perry, and Ruhter 1995; Rielly 2001

Cs-134 1963–2005 10 Rich 1990

Cs-137 1963–1999 10 Rich 1990

2000–2005 0.3 BBI 2000

Th-230 1974–2005 0.03 AEC 1974

Np-237 1974–2005 0.03 AEC 1974

U-233/234 1970–2002 0.041 Andersen, Perry, and Ruhter 1995; Rielly 2001

2003–2005 0.05 Bhatt 2003

U-235 1970–2003 0.038 Andersen, Perry, and Ruhter 1995; Rielly 2001

2003–2005 0.09 Bhatt 2003

U-238 1970–1994 0.5 Rich 1990

1995–2002 0.03 Andersen, Perry, and Ruhter 1995; BBI 2000; Rielly 2001

2003–2005 0.09 Bhatt 2003

Pu-238 1974–1994 0.03 AEC 1974

1995–2002 0.022 Andersen, Perry, and Ruhter 1995; Rielly 2001

2003–2005 0.02 Bhatt 2003

Pu-239/240 1964–1973 0.4 AEC 1964b

1974–1994 0.02 AEC 1974

1995–2005 0.03 Andersen, Perry, and Ruhter 1995; BBI 2000; Rielly 2001; Bhatt 

2002

Am-241 1974–1994 0.07 AEC 1974

1995–2001 0.023 Andersen, Perry, and Ruhter 1995; Rielly 2001

2002–2005 0.04 Bhatt 2002

Cm-244 1974–2005 0.02 AEC 1974

Cf-252 1974–2005 0.02 AEC 1974

a. When sample size is not identified in individual’s records, assume the activity is that excreted per day.

b. MDA calculated from inferred 2σ uncertainty.

Table 5-11. In vivo MDAs by period. 

Radiation/ 

radionuclide Period

In vivo result

(nCi)

Count time 

(min) Reference

Cr-51 1962–2000 12 10 Percival and Anderson 1962a

2001–2005 32 5 Rielly 2001

Mn-54 1962–2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, Ruhter 

1995

2001–2005 2.6 5 Rielly 2001

2001–2005 1.3 10 Rielly 2001

Fe-59 1962–2001 4.5 5 Rielly 2001

2001–2005 1.5 10 Rielly 2001

Co-58 1962–2000 12 10 Percival and Anderson 1962a

2001–2005 2.5 5 Rielly 2001

2001–2005 1.1 10 Rielly 2001

Co-60 1962–1970 12 10 Percival and Anderson 1962a

1971–1988 5 10 AEC 1972a; AEC 1974

1989 7 10 Martin 1989

1990–1992 2 (lung) Rich 1990

1993–2000 7 10 Grothaus 1993; Andersen, Perry and Ruhter 1995

2001–2005 2.5 5 Rielly 2001

2001–2005 1.1 10 Rielly 2001

Zn-65 1962–1988 12 10 Percival and Anderson 1962a

1989–2000 10 10 Martin 1989; Grothaus 1993; Andersen, Perry, and 

Ruhter 1995
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 20 of 29

Radiation/

radionuclide Period

In vivo result

(nCi)

Count time 

(min) Reference

2001–2005 4.9 5 Rielly 2001

2001–2005 2 10 Rielly 2001

Sr/Y-90 1968–1977 70 (skull) 10 AEC 1969a; AEC 1972a; AEC 1974

1978–2005 34 (skull) 10 Martin 1989; Grothaus 1993

Zr/Nb-95 1962–1988 12 10 Percival and Anderson 1962a

1989–2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and 

Ruhter 1995

2001–2005 2.6 5 Rielly 2001

Ru-106 2001–2005 27 5 Rielly 2001

2001–2005 7.6 10 Rielly 2001

Ag-110m 1962–2005 12 10 Percival and Anderson 1962a

Sb-125 1962–2005 14 10 Martin 1989; Grothaus 1993

I-131 1962–1989 12 10 Percival and Anderson 1962

1990–1992 2 (thyroid) 10 Rich 1990

1993–2000 0.3 (thyroid) 10 Grothaus 1993

2001–2005 3.8 5 Rielly 2001

2001–2005 0.13 (thyroid) Rielly 2001

Cs-134 1989–2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and 

Ruhter 1995

1990–2005 2(lung) Rich 1990

2001–2005 3 5 Rielly 2001

2001–2005 0.96 10 Rielly 2001

Cs-137 1962–1970 12 10 Percival and Anderson 1962a

1971–1998 5 10 AEC 1972a; AEC 1974; Martin 1989; Grothaus 1993; 

Andersen, Perry, and Ruhter 1995

1999–2000 2. (lung) 10 Rich 1990

2001–2005 3.1 5 Rielly 2001

2001–2005 1.9 10 Rielly 2001

Ba/La-140 1962–2005 12 5 Rielly 2001

Ce-141 1962–2005 9.9 5 Rielly 2001

2001–2005 3.2 10 Rielly 2001

2001–2005 0.11 (lung) 60 Reilly 2001

Ce-144 1962–2000 50 10 Martin 1989; Grothaus 1993; Andersen, Perry, and 

Ruhter 1995

2001–2005 44 5 Rielly 2001

2001–2005 15 10 Rielly 2001

2001–2005 0.44 (lung) 60 Rielly 2001

Eu-152 1962–2005 4 10 Rielly 2001

2001–2005 0.18 (lung) 60 Rielly 2001

Eu-154 1962–2005 2 10 Rielly 2001

Eu-155 1962–2005 1 10 Rielly 2001

Ga-153 1962–2005 6.5 10 Rielly 2001

2001–2005 0.096 (lung) 60 Rielly 2001

Hf-181 1962–2005 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and 

Ruhter 1995

Ta-182 1962–2005 12 10 Percival and Anderson 1962a

Hg-203 1962–2005 12 10 Percival and Anderson 1962a

Th-230 1974–2005 1,000 AEC 1974

Th-234 2001–2005 1.4 (lung) 60 Rielly 2001

Np-237 1974–2005 AEC 1974

U-235 1993 0.2 (wound) 20 Grothaus 1993

1962–2005 0.2 (lung) Rich 1990

2001 0.11 (lung) 60 Rielly 2001

U-dep/nat 1989 3 (lung) 60 Martin 1989; Grothaus 1993

Pu-238 1989–1998 26 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and 

Ruhter 1995

1993 1 (wound) 20 Grothaus 1993

1999–2000 30 (lung) Rich 1990
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 21 of 29

Radiation/

radionuclide Period

In vivo result

(nCi)

Count time 

(min) Reference

2001 54 (lung) 60 Rielly 2001

Pu-239/240 1971–1993 30 100 AEC 1972a

1974–1988 74 (lung) 100 AEC 1974

1989–1988 80 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and 

Ruhter 1995

1993 2 (wound) 20 Grothaus 1993

1990–2000 30 (lung) Rich 1990

2001–2005 140 (lung) 60 Rielly 2001

Am-241 1989–1999 0.6 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and 

Ruhter 1995

1993 0.1 (wound) 20 Grothaus 1993

1990–2000 0.2 (lung) Rich 1990

2001–2005 0.14 (lung) 60 Rielly 2001

a. MDA calculated from inferred 2σ uncertainty.

products. As a consequence, the in vivo counting program was used to count (1) all terminating 

employees, (2) employees suspected of having a possible internal intake, and (3) any interested 

employee on an annual basis. Only those activities greater than 0.1 μCi were further quantified. This 

level was determined to be less than one-tenth of the MPBB for most of the gamma-emitting isotopes.

5.6 DEFAULT FOR MISSED DOSE

Based on the ANL-W operational environment characteristics, a number of missed dose default 

assumptions have been derived. The breeder reactors have a fast neutron spectrum and stainlesssteel clad fuel. Rather than include all of the radionuclides in the default summary table for missed 

dose, only 90Sr, 137Cs, 144Ce, and 239Pu are included for stainless-steel fuels. Cesium-137 was 

selected because it is most commonly reported in the in vivo results rather than for its dose 

contribution. The potential missed inhalation dose from the other radionuclides is accounted for by 

weighting the dose from these selected radionuclides by the weighting factors. This gives an 

equivalent to 100% of the dose from the radionuclide distributions for the three types of fuels. 

Table 5-12 is a summary of these recommended defaults. 

Table 5-12. Default table for missed dose. 

Period Based on Recommendation Basis

Start up date 

through 1960

Urine

gross β

Calculate chronic Sr-90 intake that 

results in a urine activity of 0.4 × gross β

Typical β activity is 0.33 Sr-90, 0.33 

Y-90 and 0.33 Cs-137. Use of 0.4 is 

favorable to claimants.

Cs-137 intake = Sr-90 intake Half-lives and fission yields of Cs-137 

are Sr-90 are approximately equal.

Pu-239 intake = 0.004 × Sr-90 intake Pu:Sr-90 ratio of 0.003 weighted by a 

factor of 1.2. 

Ce-144 intake = 2.4 × Sr-90 intake Ce-144:Sr-90 ratio of 2 weighted by a 

factor of 1.2. 

1961–1980 In vivo

Cs-137

Calculate chronic Cs-137 intake that 

results in the in vivo measurement

Not applicable.

Sr-90 intake = Cs-137 intake when no in 

vitro measurement

Half-lives and fission yields are 

approximately equal.

Pu-239 intake = 0.004 × Cs-137 intake 

when no in vitro measurement

Pu:Cs-137 ratio of 0.003 in stainlesssteel fuel weighted by a factor of 1.2. 

Ce-144 intake = 2.4 × Cs-137 assigned 

intake when no measurement

Ce-144:Cs-137 ratio of 2 in stainlesssteel fuel weighted by a factor of 1.2.

1981–2005 Bioassay Use bioassay results. Not applicable.
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 22 of 29

For most of the history of the ANL-W facilities, personnel dosimeters were issued to all workers who 

entered the security access control points at each facility, regardless of work assignments. For 

example, administrative and clerical personnel were required to wear these radiation-monitoring 

dosimeters even though they were not exposed to elevated background levels or internal dose 

potential. Moreover, WBC was made available to any worker regardless of the likelihood of exposure 

to radiation. If exposure was likely, the worker was requested to have a whole-body count (ES&H 

1989). 

5.7 UNMONITORED WORKERS

If no detectable external dose or WBC information was recorded and no in vitro samples were 

recorded, the person should be considered an unmonitored worker for internal dose purposes and 

only the environmental dose should be included.

As noted above, ANL-W personnel dosimeters were issued to all workers. Many of these workers, 

due to the nature of their work environment, would not have inhaled or ingested radioactivity and 

therefore would not have been subject to routine bioassay.

Most radioactivity encountered was well tagged with beta/gamma activity, which produced 

measurable direct radiation doses on a personnel dosimeter. Therefore, the probability that a worker 

with no external dose received a significant unmonitored internal intake of radioactive material is very 

low. Individuals who were not issued a personal dosimeter and have no record of internal dose 

monitoring should be assigned only the environmental dose for the facility.

At ANL-W each construction job was evaluated to determine if radiation exposure or internal dose 

could be received. When construction work was done in an area with potential radiation exposure, 

including internal dose exposure, construction workers were monitored in the same fashion as 

radiation workers (Andersen, Perry and Ruhter 1995 and Reilly 2001). Construction workers who 

were issued personnel dosimeters should be treated the same as facility employees who were issued 

personnel dosimeters. Construction workers who were not issued a personnel dosimeter should be 

assigned only the environmental dose for the facility.

5.8 ATTRIBUTIONS AND ANNOTATIONS

All information that requires identification was addressed via references integrated into the reference 

section of this document.

Norman Rohrig served as the initial Document Owner for this document. Mr. Rohrig was previously 

employed at INL, which shared boundaries with ANL-W and used the same dosimetry systems. His 

work involved management, direction, or implementation of radiation protection and/or health physics 

program policies, procedures, or practices in relation to atomic weapons activities at the site. This 

revision has been overseen by a new Document Owner, who is fully responsible for the content of this 

document, including all findings and conclusions. Mr. Rohrig continues to serve as a Site Expert for 

this document because he possesses or is aware of information relevant to reconstruction of radiation 

doses that were experienced by claimants who worked at the site. In all cases where such 

information or previous studies or writings are included or relied upon by the Document Owner, those 

materials are fully attributed to the source. 
Document No. ORAUT-TKBS-0026-5 Revision No. 01 Effective Date: 05/23/2007 Page 23 of 29

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Division, Idaho Falls, Idaho.

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AEC (U.S. Atomic Energy Commission), 1961, Annual Report of Health & Safety Division, 1960, 

IDO-12019, Idaho Operations Office, Idaho Falls, Idaho.

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AEC (U.S. Atomic Energy Commission), 1963, Annual Progress Report 1962, Health & Safety 

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Grothaus, G. E., 1993, “MDAs For In-Vivo Counting,” letter to C. Bloom, EG&G, ES&Q, February 19.

Holson, C. E., 1991, “1990 Radiation Exposure Summary,” memorandum to G. C. Marshall, 

ANL-W-SES-(CEH)-91-42, February 20.

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Kirchner, D. S., 1992, “1991 Radiation Exposure Summary,” memorandum to G. C. Marshall, 

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GLOSSARY

absorption

As used in this internal dosimetry section, absorption refers to the material being transported 

to fluids and other organs as well as radiation energy being imparted. 

activation

The process of inducing radioactivity by irradiation.

becquerel

A unit of radioactivity equal to one-disintegration per second.

beta radiation

Radiation consisting of electrons or positrons emitted at high velocity from the nuclei of certain 

radioactive elements. Most direct fission products emit beta radiation. 

breeder reactor

A nuclear reactor concept in which the operation produces a net increase in fissionable 

material.

cladding

The outer layer of material encasing a reactor fuel element (e.g., aluminum or zirconium). 

Cladding promotes the transfer of heat from the fuel to the coolant and contains fission 

products and activation products that result from the fissioning of the fuel.

contamination, radioactive

Radioactive material where it does not belong. 

core

That part of the reactor consisting of the fuel and some of the control elements for reactor 

operation.

curie

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

decontaminate

Removing a contaminant, such as a radioactive material, from an undesired location.

dosimeter

A device used to measure accumulated radiation exposure. 

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation. 

fission

A 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

Radionuclides resulting from fission.
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fuel reprocessing

A chemical process, usually involving several steps, that recovers 235U and other fissionable 

products from spent fuel.

gamma rays

Short wave length electromagnetic radiation (photons) originating in atomic nuclei and 

accompanying many nuclear reactions (e.g., fission, radioactive decay, and neutron capture) 

in an energy range of 10,000 electron volts to 9 million electron volts.

half-life

The time it takes for one-half of any given number of unstable atoms to decay (disintegrate).

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

in vitro

In glass. Outside the living body and in an artificial environment. Internal bioassay sampling, 

such as fecal samples or urine samples.

in vivo 

In the living. In the living body of a plant or animal. Bioassay sampling by whole-body 

counting. 

isotope

Nuclides having the same number of protons in their nuclei (same atomic number), but having 

a differing number of neutrons (different mass number).

millirem

A unit of radiation dose equivalent (or equivalent dose) equal to one-thousandth of a rem (see 

rem).

microcurie

A measure of radioactivity equal to one-millionth of a curie. 

mixed waste

Waste that contains hazardous and radioactive materials. 

natural uranium

Uranium that has not been through an enrichment process.

neutron

A basic particle in a nuclear reaction, electrically neutral, with nearly the same mass as a 1

H 

atom.

nuclear waste

A general term used for the byproduct unusable material resulting from nuclear reactions, 

including high-level, intermediate, low-level, mixed and TRU waste.

nucleus

That part of an atom consisting of the total positive electrical charge and most of the mass. 
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quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

radiation

Energy transferred through air or some other media in the form of particles or waves (see 

ionizing radiation).

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma or X-rays, or 

neutrons from unstable atoms.

radionuclide

A radioactive species of an atom characterized by the constitution of its nucleus specified by 

atomic number (the number of protons), and the mass number (equal to the number of protons 

plus neutrons). 

rem

A unit of dose equivalent, equal to the product of the absorbed dose and the quality factor. 

The word derives from roentgen equivalent in man. 

shielding

Any material or obstruction that absorbs (or attenuates) radiation to protect personnel or 

materials.

spent nuclear fuel

Reactor fuel containing fission and activation products that can no longer economically sustain 

a chain reaction.

type

Refers to the rate of absorption from lung to blood of inhaled radioactive materials and 

includes types F (fast), M (moderate), and S (slow).

transuranic (TRU) waste

Contaminated waste materials with nuclides having an atomic number greater than 92, a halflife over 20 yr and concentration greater than or equal to 100 nCi/g. 

U.S. Atomic Energy Commission

An agency established by the U.S. Government for oversight of nuclear weapons and power 

production; a predecessor to the U.S. Department of Energy.

X-ray

Ionizing electromagnetic radiation of extranuclear (outside the nucleus) origin.

zirconium

A metallic element highly resistant to corrosion and often used to make cladding for nuclear 

fuel. It is sometimes alloyed in small amounts in the fuel itself.