ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 63

DOE Review Release 03/12/2010

Document Title:

Idaho National Laboratory and Argonne National 

Laboratory – West - Occupational Internal Dose

Document Number: ORAUT-TKBS-0007-5

Revision: 03

Effective Date: 03/02/2010

Type of Document: TBD

Supersedes: Revision 02 and ORAUT-

 TKBS-0026-5 Revision 01

Subject Expert(s): Brian P. Gleckler

Site Expert(s): Bryce L. Rich, Douglas Wenzel, Norman D. Rohrig, Boyd Levitt, and Henry Peterson

Approval: Signature on File Approval Date: 03/01/2010

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 03/01/2010

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 03/01/2010

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 03/01/2010

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/02/2010

Stuart L. Hinnefeld, Interim Director, DCAS

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-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 2 of 63

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

08/11/2004 00 New Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory – Occupational Internal Dose. First 

approved issue. Initiated by Norman D. Rohrig.

10/12/2004 00 PC-1 Approved page change revision. Deletes references to rigorous 

radiation program on page 39. Initiated by Norman D. Rohrig. 

Approval:

Signature on File 09/30/2004

Norman D. Rohrig, Document Owner

Signature on File 10/01/2004

Judson L. Kenoyer, Task 3 Manager

Signature on File 10/01/2004

Richard E. Toohey, Project Director

Signature on File 10/12/2004

James W. Neton, Associate Director for Science

01/13/2006 01 Approved issue of Revision 01. Revised to change table number 

format, responds to new information from Worker Outreach Effort 

and revises MDA tables. Incorporates 1-year decay time for Al fuel 

in Tables 5-18 and 5-24. Combined “INTEC & Unknown” and “Other 

Area” categories in Table 5-24 for 1971-80 into “All but ANL-W.” 

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

Norman D. Rohrig.

06/05/2006 01 PC-1 Page change revision to correct error on page 24 in Table 5-12 in 

Section 5.5 on sample volume in 1980s. Approved issue of Rev 01 

PC-1. No sections were deleted. Incorporates NIOSH formal review 

comments. This revision results in a reduction in assigned dose and 

no PER is required. Training required: As determined by the Task 

Manager. Initiated by Norman D. Rohrig. 

Approval:

Signature on File 05/23/2006

John M. Byrne, Document Owner

Signature on File 05/23/2006

John M. Byrne, Task 3 Manager

Signature on File 05/22/2006

Edward F. Maher, Task 5 Manager

Signature on File 05/24/2006

Kate Kimpan, Project Director

Signature on File 06/05/2006

James W. Neton, Associate Director for Science

07/30/2007 02 Revision initiated to add Purpose, Scope, and Attributions and 

Annotations sections. Added references. Updated the formatting 

and made editorial changes. Corrected error in Fecal MDA for 

Th-230 and Np-237 in Table 5-13. Constitutes a total rewrite of the 

document. Incorporates formal internal and NIOSH review 

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

Initiated by Jo Ann M. Jenkins.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 3 of 63

03/02/2010 03 Revision to incorporate the Argonne National Laboratory - West site. 

In order to resolve several issues identified in an independent audit 

of this TBD and other issues identified by ORAUT, the approaches 

for assessing the INEL bioassay data have been completely revised. 

The potential lung absorption types to be considered have also been 

expanded for several radionuclides. These changes generally result 

in lower organ doses. However, for specific scenarios they can 

result in higher doses to certain metabolic organs such as the 

thyroid. The layout of this TBD was completely reorganized to better 

match the layout recommended in ORAUT-PROC-0031. In addition, 

this revision fully incorporates the Argonne National Laboratory –

West back into this TBD and eliminates the ANL-W Occupational 

Internal Dose TBD. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by Jo 

Ann M. Jenkins. 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 4 of 63

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ...............................................................................................................7

5.1 Introduction ............................................................................................................................10

5.1.1 Purpose......................................................................................................................11

5.1.2 Scope.........................................................................................................................11

5.1.3 Historical Overview..................................................................................................... 11

5.1.3.1 Test Reactor Area........................................................................................ 12

5.1.3.2 Argonne National Laboratory – West............................................................ 13

5.1.3.3 Other Test Reactor Areas ............................................................................ 13

5.1.3.4 Idaho Chemical Processing Plant (ICPP)..................................................... 14

5.1.3.5 Other Nuclear Facilities and Processes........................................................ 15

5.1.4 Radionuclides of Concern and Solubility..................................................................... 16

5.1.5 Intake Modes and Particle Size Distributions .............................................................. 17

5.1.6 Internal Dosimetry Program........................................................................................ 17

5.1.6.1 Contamination Control ................................................................................. 19

5.1.6.2 Air Monitoring............................................................................................... 19

5.1.6.3 Bioassay Monitoring..................................................................................... 20

5.1.7 Recordkeeping ........................................................................................................... 25

5.1.7.1 Radiological Incident Records...................................................................... 30

5.2 In Vitro Minimum Detectable Activities, Counting Methods, and Reporting Protocols ............. 30

5.2.1 Urine Sample Analyses .............................................................................................. 30

5.2.2 Fecal Sample Analyses .............................................................................................. 31

5.3 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Protocols ............. 32

5.3.1 Whole-Body Counting................................................................................................. 32

5.3.1.1 General Information on Whole-Body Counting ............................................. 32

5.3.1.2 Attributing Positive WBCs to Cs-137 Fallout................................................. 34

5.3.2 Lung Counts...............................................................................................................35

5.3.3 Thyroid Counts ........................................................................................................... 35

5.3.4 Skull Counts ...............................................................................................................35

5.3.5 Wound Counts............................................................................................................ 36

5.4 Interferences and Uncertainties..............................................................................................36

5.4.1 Interferences ..............................................................................................................36

5.4.1.1 Contamination of Samples ........................................................................... 36

5.4.1.2 Naturally Occurring Uranium in Uranium Bioassay Measurements .............. 36

5.4.1.3 Cesium-137 from Fallout.............................................................................. 37

5.4.1.4 Intakes of Radionuclides for Medical Reasons ............................................. 37

5.4.2 Uncertainties ..............................................................................................................37

5.5 Intake and Internal Dose Assessment for Monitored Workers ................................................ 38

5.5.1 Activation and Fission Product Intakes ....................................................................... 38

5.5.1.1 Assessment of Missed Intakes..................................................................... 39

5.5.1.2 Assessment of Fitted Intakes ....................................................................... 39

5.5.1.3 Assessment of Unmonitored Intakes............................................................ 40

5.5.1.4 Decay Period Selections .............................................................................. 40

5.5.2 Actinide Intakes .......................................................................................................... 41

5.5.3 Contaminant Intakes for ICPP Workers Who Handled Recycled HEU ........................ 42
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5.5.4 Radioactive Iodine Intakes.......................................................................................... 43

5.5.4.1 General Guidance on Identified Radioactive Iodine Intakes ......................... 43

5.5.4.2 Special Considerations for Thyroid Cancer Cases ....................................... 44

5.5.5 Uranium Intakes for SMC Project Workers ................................................................. 45

5.5.6 Intakes for RWMC Workers ........................................................................................ 46

5.5.7 Intakes for Waste Reduction Operations Complex Workers ....................................... 47

5.5.8 Sodium-24 Intakes at Liquid-Sodium-Cooled Reactors............................................... 48

5.6 Intake and Internal Dose Assessment for Unmonitored Workers............................................ 48

5.7 Attributions and Annotations ..................................................................................................49

References ........................................................................................................................................53

Glossary ............................................................................................................................................59
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LIST OF TABLES

TABLE TITLE PAGE

5-1 Major airborne incidents at the TRA....................................................................................... 13

5-2 Major airborne incidents at the ICPP...................................................................................... 14

5-3 Primary radionuclides of concern ........................................................................................... 17

5-4 Surface contamination control and MDAs .............................................................................. 19

5-5 Routine Bioassay History Summary ....................................................................................... 20

5-6 Urinalysis results in 1959, 1960, and 1961............................................................................. 24

5-7 Summary statistics from the 1963 WBC program................................................................... 25

5-8 Internal dose assessment information before 1989 ................................................................ 26

5-9 Internal dose assessment information after 1989 ................................................................... 26

5-10 Column descriptions for urine sample record sheets before February 1955 ........................... 27

5-11 Column descriptions for urine sample record sheets from February 1955 and later ............... 28

5-12 Location codes that could be in worker dose files .................................................................. 29

5-13 Derived investigation levels in 1977 for acute exposures ....................................................... 30

5-14 Urine sample MDAs by period................................................................................................ 31

5-15 Fecal sample MDAs by period ............................................................................................... 32

5-16 Whole-body count MDAs by period ........................................................................................ 33

5-17 Mean body burdens of Cs-137 from fallout in the United States............................................. 34

5-18 Lung count MDAs by period ................................................................................................... 35

5-19 Thyroid count MDAs by period ............................................................................................... 35

5-20 Skull count MDAs by period ................................................................................................... 36

5-21 Wound count MDAs by period................................................................................................ 36

5-22 Actinide-to-Sr-90 ratios .......................................................................................................... 42

5-23 Actinide-to-Cs-137 ratios........................................................................................................ 42

5-24 Default ratios for recycled HEU intakes.................................................................................. 43

5-25 RaLa iodine............................................................................................................................44

5-26 I-129-to-indicator nuclide ratios .............................................................................................. 45

5-27 Uranium mass and activity ratios for SMC project DU............................................................ 46

5-28 SMC Project DU impurities..................................................................................................... 46

5-29 Radioactive low-level waste inventory in the active pits in the SDA........................................ 47

5-30 Radioactive waste inventory in the TSA ................................................................................. 47
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AEDE annual effective dose equivalent

AMAD activity median aerodynamic diameter

ANL-W Argonne National Laboratory–West (recently renamed the Materials and Fuels 

Complex [MFC])

ANP aircraft nuclear propulsion

ATR Advanced Test Reactor

CADRE guard force

CAM continuous air monitor

CDE committed dose equivalent

CEDE committed effective dose equivalent

CFA Central Facilities Area

CFR Code of Federal Regulations

Ci curie

COO Chicago Operations Office

cpm counts per minute

CPP Chemical Processing Plant

d day

DAC derived air concentration

DE dose equivalent

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

DU depleted uranium

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

ETR Engineering Test Reactor

F fast absorption type

FCF Fuel Cutting Facility

ft foot

g gram

gal gallon

GCRE Gas-Cooled Reactor Experiment

GSD geometric standard deviation

H&S Health and Safety

HFEF Hot Fuel Examination Facility

hr hour

HSD Health and Safety Division

HSL Health Services Laboratory

HTO tritiated water vapor
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ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

IDO Idaho Operations Office

IET initial engine test

IMBA Integrated Modules for Bioassay Analysis

in. inch

INEEL Idaho National Engineering & Environmental Laboratory

INEL Idaho National Engineering Laboratory

INL Idaho National Laboratory

INTEC Idaho Nuclear Technology and Engineering Center

L liter

LPTF Low Power Test Facility

m meter

M moderate absorption type

MAP mixed activation product

MDA minimum detectable activity

MDL minimum detectable level

MeV megavolt-electron, 1 million electron-volts

MFC Materials and Fuels Complex (new name for ANL-W)

MFP mixed fission product

mi mile

min minute

ml milliliter

mo month

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

OMRE Organic Moderated Reactor Experiment

ORAU Oak Ridge Associated Universities

ORAUT Oak Ridge Associated Universities Team

PBF Power Burst Facility

pCi picocurie

PF photofluorimetry

POC probability of causation
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RaLa radioactive lanthanum

RAM radiation (or remote) area monitor

RCIMS Radiation Control Information Management System

RDR Radiation Dosimetry and Records

RESL Radiological Environmental Sciences Laboratory

RFP Rocky Flats Plant

RWMC Radioactive Waste Management Complex

S slow absorption type

SDA Subsurface Disposal Area

SL-1 Stationary Low-Power Reactor

SMC Specific Manufacturing Capability

SPERT Special Power Excursion Reactor Test

SRDB Ref ID Site Research Database Reference Identification (number)

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

TRU transuranic

TSA Transuranic Storage Area

μCi microcurie

μg microgram

μm micrometer

U.S.C. United States Code

WBC whole-body counting/count

WERF Waste Experimental Reduction Facility

WROC Waste Reduction Operations Complex

yr year

ZPPR Zero Power Plutonium (later Physics) Reactor 

α alpha particle

β beta particle 

γ gamma ray or photon

σ standard deviation

3x3 3 inch by 3 inch

§ section or sections
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5.1 INTRODUCTION 

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

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

documents that provide 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

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2007):

] guidelines established under 

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

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

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007).

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

conventional structures

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 11 of 63

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

reasons.

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

justification, or clarification of the associated information, are presented in Section 5.7.

5.1.1 Purpose

The purpose of this TBD is to document the internal dosimetry program and practices at the Idaho 

National Laboratory, and to provide the technical basis to be used to evaluate the internal 

occupational radiation dose for EEOICPA claims.

5.1.2 Scope

This TBD provides supporting documentation to assist in the evaluation of occupational internal doses 

in accordance with the guidelines described in Internal Dose Reconstruction Implementation 

Guideline (NIOSH 2002). NIOSH considers the available data and methods for performing internal 

dose reconstruction to be adequate for estimating with sufficient accuracy the internal doses at the 

INEL Site throughout its entire history.

5.1.3 Historical Overview

In 1949, the U.S. Atomic Energy Commission (AEC) established the NRTS and started construction of 

facilities on a 572,000-acre site approximately 50 miles west of Idaho Falls in southeastern Idaho. 

The site has experienced several name changes since its creation. The following are the names the 

site has been called over the years and the approximate date ranges for those names: NRTS (1949–

1973), INEL (1974–1996), INEEL (1997–2004), and INL (2005 to the present). For convenience, 

INEL is used throughout the remainder of this document where it is unnecessary to distinguish

between one of its other names.

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

days of the AEC programs represented 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 the INEL that 

had a marked influence on the internal dosimetry programs at each of the facilities were:

• The original mission of the INEL was the development of high enriched uranium (HEU) reactor 

concepts, materials testing through high-flux test reactor operation, and chemical processing 

of those (valuable) fuels. The U-235 enrichment of the HEU at the INEL was over 50% and 

mostly over 90%. The production of weapons-grade nuclear materials was not a mission.

• The INEL began operations 8 to 10 years after Oak Ridge National Laboratory (ORNL) and 

the Hanford Site. During those developmental years significant technical progress in 

professional skills, instrumentation, analyses, procedures, and techniques was accomplished. 

Radiation safety programs and techniques from ORNL (ACC 1952) were adopted at the 

startup of the INEL facilities. 

• Two AEC field offices (Chicago and Idaho) were responsible for and had oversight of the INEL

programs included in this report. In addition, the Nuclear Navy under direction of the 

Pittsburgh Field Office administered programs and used facilities on the INEL for training and 

program development. Thus, there were three Federal organizations using the INEL and its 

infrastructure. This TBD does not apply to naval facilities or naval personnel, even if those 

personnel received exposure from AEC operations. 
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• In addition to the Federal agencies involved at the site, numerous contractors operated the 

many facilities for the agencies and shared support personnel to various degrees.

• To provide consistency of radiation safety programs at the INEL among a large variety of 

facilities and constantly changing contractors, the AEC established a Health and Safety (H&S) 

Laboratory at the INEL 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) technical support in quality assurance of external 

and internal radiation dose evaluation; (6) maintenance and documentation of personnel 

dosimetry records; and (7) research and development in these areas of responsibility. The 

name of this organization changed to Health Services Laboratory (HSL), then to the Health 

and Safety Division (HSD), then to the Idaho Center for Radiological and Environmental 

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

(RESL) [1].

• Although the design and administration of the radiation safety programs in the workplace were 

the responsibility of each facility contractor, AEC conducted oversight. Technical data, 

information (particularly in the instances of detectable worker intake), and analytical internal 

dose calculations and evaluations were exchanged between the HSL and each contractor [1].

As a consequence, and in spite of the constant changes at the INEL, basic assumptions about 

detection levels (e.g. MDAs, MDLs, etc...) and missed dose potential were relatively consistent across 

the years [2]. There were differences in the available nuclear materials from facility to facility, but by 

early 1958 gamma spectral analysis capabilities at the INEL allowed the significant bioassay results 

(those that would result in reportable internal dose) to be defined in terms of the specific radionuclides

(Hayden 1958). The practice in the case of a higher urine sample result was to attempt radionuclide 

identification through gamma spectral analysis and chemical separation. This TBD describes default 

assumptions for use in cases when the bioassay records for a worker do not include the radionuclide 

specific analyses and only include records for gross radioactivity measurements.

5.1.3.1 Test Reactor Area

The Test Reactor Area (TRA) complex includes hot cells, a gamma irradiation pool facility, research 

laboratories, and analytical laboratories. The reactors at the TRA, as well as the others at the INEL, 

were used for testing materials, experiments, neutron irradiation facilities, and so forth (Stacy 2000). 

They were not involved in the production of plutonium or any other weapons materials, unlike some 

reactors at other DOE sites. The MTR was the second operating reactor at the INEL and ran from 

March 1952 to 1970 in the TRA (Stacy 2000). The TRA has also hosted the Engineering Test 

Reactor (ETR; 1957 to 1981) and the ATR (1967 to present) along with six reactor-critical facilities 

that supported the test reactors (Stacy 2000). 

The uranium used in the TRA reactors is enriched to 93% U-235, and the fuel is clad in aluminum. 

The predominant activation product in the cladding is Na-24, which is formed by activation of sodium 

in the aluminum. Sodium-24 (Na-24) has a half-life of 15 hours and emits a high-energy gamma ray 

(2.75 MeV). The inhalation dose to personnel from Na-24 is insignificant in comparison to that from 

the fission products in the fuel [3]. There are minor levels of activation products of stainless steel 

(Co-58, Co-60, Cr-51, Mn-56, etc.) in the primary coolant system due to corrosion of its stainless-steel 

components [4]. 

Several factors contributed to unusual amounts of fission products in the primary coolant systems of 

the MTR and ETR during early operations. With cladding technology in its infancy, the quality of the 

cladding was not the best and fission products leaked through it [4]. Another factor was fuel 
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contaminated with tramp uranium on the outside of the cladding. During reactor operation, fission 

products in tramp uranium were released directly into the primary coolant system [4]. Reactor 

operators and the fuel manufacturer resolved these deficiencies over time; by the time the ATR 

became operational, the primary coolant system of the ATR was considerably less contaminated than 

that of the MTR or the ETR during their early years of operation. In addition, reactor components 

such as fuel elements, reactor loop components, and so forth that are removed from the reactor and 

placed in the canal are a source of contamination in the canal water. If these components are not 

cleaned adequately before they are removed from the canal, the activity on them can become 

airborne.

The majority of radioactivity releases from the TRA reactors to areas potentially occupied by workers 

consisted of noble gases that decayed to short-lived particulates. The principal dose to personnel 

from releases of noble gases was direct radiation rather than inhalation [5]. Table 5-1 provides a 

summary of the major airborne incidents that occurred at the TRA.

Table 5-1. Major airborne incidents at the TRA.

Date Incident

Radionuclide(s) 

released Reference

03/28/1954 GE-ANP-1 depressurization Noble gas Sommers 1954

12/17/1958 GEH-4 rupture at MTR Noble gas + iodine Sommers 1958

1961 Ag-110m spill at ETR Ag-110m Horan 1962

06/13/1967 GA-18-1 depressurization at ETR Ta-182 and Ta-183 Nertney et al. 1967

01/06/1977 Noble gas release at ATR Noble gas Sommers 1977

5.1.3.2 Argonne National Laboratory – West

The Argonne National Laboratory near Chicago established a branch at the INEL known as the 

Argonne National Laboratory – West (ANL-W), where it built and operated several reactors that were 

of fundamental importance for the development of commercial nuclear power. The 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. In February 2005, the ANL-W was merged into the INEL, and was 

subsequently renamed the Materials and Fuels Complex (MFC). However, for convenience, this TBD 

uses ANL-W, that has been this area’s name for most of its history and is the name that it is known as 

by most of the EEOICPA claimants.

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

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

the other at the current location on the southeast side of the INEL Site. Early reactor operations 

included physics critical experiments; power production; routine unmoderated reactor operation; 

uranium-fueled, plutonium-fueled, and breeder reactor designs; and self-destruction experiments.

5.1.3.3 Other Test Reactor Areas

As the primary nuclear reactor development laboratory in the United States, the INEL tested or 

evaluated more than 100 reactor concepts (DOE 1997). Fifty-two test reactors were designed, 

constructed, and operated (including operation-to-destruction tests) at the INEL. The INEL has 

experienced a number of episodic reactor events, both planned and accidental [for example, the 

military Stationary Low-Power Reactor (SL-1) accident on January 3, 1961 (Stacy 2000); a series of 

deliberate safety experiments by the ANL-W in which reactors were allowed to go prompt-critical with 

resultant reactor destruction (Stacy 2000); and the Aircraft Nuclear Propulsion (ANP) Program that 

operated initial engine tests (IETs) with large environmental releases in the 1950s (DOE 1991)]. 

External and internal doses to workers, both expected and accidental, were associated with these 

events (Till et al. 2002). 
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The largest internal exposures at the INEL resulted from accidental intakes associated with episodic 

events or planned major releases, for which the times and characterizations of the intake materials 

were well known [6]. These exposures were documented in each exposed employee’s file.

5.1.3.4 Idaho Chemical Processing Plant (ICPP)

Initially known as the Chemical Processing Plant (CPP), renamed as the Idaho Chemical Processing 

Plant (ICPP), and eventually renamed as the Idaho Nuclear Technology and Engineering Center 

(INTEC); this TBD uses ICPP for convenience where it is unnecessary to distinguish between one of 

its other names. The ICPP reprocessed highly enriched reactor fuel (U-235 enrichments of 50% to 

93%) for 39 years from 1952 to 1991. Aged mixed fission products (MFPs) were the predominant 

internal hazard, although enriched uranium isotopes and plutonium isotopes (Pu-238 enhanced) were 

limiting in specific process locations. 

The ICPP processes were remotely controlled but contact maintenance was required; that is, 

maintenance personnel entered process cells and repaired equipment by hand. The process 

equipment in the cells, which had walls of 5-ft-thick high-density concrete, were decontaminated by 

flushing and rinsing with concentrated acids and complexing agents before entry by health physics 

and maintenance personnel. These occasional operations were well planned, but they had high 

potentials for internal exposures [7].

Most internal doses experienced at the ICPP were from accidental releases. Table 5-2 lists unusual 

and episodic events that have occurred at the ICPP. The ICPP experienced not only operational 

containment barrier failures but also accidental criticality events in 1959, 1961, and 1978. Because 

the criticality accidents occurred in process vessels in heavily shielded cells, these events resulted in 

relatively minor worker intakes (Stacy 2000; AEC 1960; Horan 1962). These exposures are 

documented in the personnel dosimetry files. 

A variety of fuel types from a multitude of reactors were processed at the ICPP throughout its 

operating history. The dates when processing started for each of the major three fuel types were 

approximately 1953 for aluminum fuels, 1956 for zirconium fuels, and 1965 for stainless-steel fuels 

(Staiger 2003). There are relatively long half-life fission products that persist for the ICPP source 

terms. In most cases, the source terms were well tagged with beta-emitting radionuclides, which 

allowed beta/gamma-detecting CAMs and bioassay methods for beta/gamma-emitting radionuclides

to be used at the ICPP with the realization that they would also warn of possible alpha contamination 

or internal exposures [8]. 

The irradiated fuel was normally decayed a minimum of 90 days before shipment to the ICPP for 

processing (Allied Chemical undated), to minimize the radiological safety hazard of the relatively 

volatile halogens. After being received at the ICPP, the irradiated fuel was kept in storage until it had 

a minimum of 120 days of cooling; however, the processing of fuels often did not occur until months or 

years later (Allied Chemical undated). Because of this relatively long decay time, many of the short 

half-life radionuclides decayed considerably, leaving the actinides to make up a larger percentage of 

the total radionuclide inventory of the processed fuel [9]. 

Table 5-2. Major airborne incidents at the ICPP. 

Date Incident

Radionuclides

released Internal dose discussion Reference

05/15/57 Iodine release to Y-Cell I-131 Y-Cell modifications resulted in 8 

personnel receiving minor thyroid 

doses in the range of 600 mrem.

Vance 1957

03/20/58 Iodine release I-131 Radioactive iodine spread through 

makeup area to operating corridor. 

Thyroid intake to several health 

Rich 1958; 

Hayden 1958
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Date Incident

Radionuclides

released Internal dose discussion Reference

physics technicians and operators 

in the 40-μCi range.

10/16/59 Criticality accident - in 

shielded process system

Short-lived noble 

gases and I-131, 

I-132, I-133, etc.

Short-lived radioactive gases 

released to plant areas; internal 

doses reported as minimal.

Ginkel et al. 

1960

01/25/61 Criticality accident - in 

shielded process system

Short-lived noble 

gases & I-131, 

I-132, I--133, etc.

Short-lived radioactive gases 

released through process off-gas 

system to 76-m stack. Internal 

doses reported as minimal.

Paulus et al. 

1961

01/72 Release of ~1.0 Ci Ru106 from main stackRu-106 No internal doses detected. ERDA 1977

11/17/72 ICPP mass 

spectrometry Pu 

contamination incident

Pu-238, Pu-239 An exposure incident involving 

about a dozen personnel resulted 

in 50-yr exposure lung doses 

ranging up to about 4 rem.

Wenzel 1973, 

1974

10/17/78 Criticality accident in 

shielded process system

Short-lived noble 

gases

Short-lived lived radioactive gases 

released through process off-gas 

system to 76-m stack. Internal 

doses reported as minimal.

Casto 1980

11/85 N-Cell Pu uptake Pu-238 Internal exposures were far below 

DOE exposure limits, but showed 

a weakness in the radiological 

control program.

Henry and 

Slagle 1985

10/30/88 Release of ~0.2 Ci of 

Ru-106 from main stack

Ru-106 No internal doses detected. Hoff, Mitchell, 

and Moore 1989

One exception to this planned fuel aging was the radioactive lanthanum (RaLa) process, which 

operated in L cell of the 601/602 process building from February 1957 to 1963 (Stacy 2000). This 

process was designed to extract RaLa from green fuel from the MTR with as little decay as 

manageable (less than 2 days). Fuel was removed from the MTR, transported about 2 miles to the 

ICPP in a heavily shielded transport container by a straddle carrier, immediately dissolved, and the 

barium element was extracted. The Ba/La-140 product was shipped immediately to Los Alamos 

National Laboratory. The INEL’s Ra/La process released large quantities of volatile radioactive iodine

isotopes, because early designs for the process were inadequate for controlling iodine (Hayden 

1958). Several significant internal exposure incidents occurred in which I-131, I-132, and I-133

intakes occurred before personnel could respond to CAM alarms and take protective or corrective 

actions [10].

5.1.3.5 Other Nuclear Facilities and Processes

Other nuclear facilities at the INEL where intakes occurred include:

• The Radioactive Waste Management Complex (RWMC) handled radioactive wastes 

generated by nuclear facilities on the INEL Site and was the primary disposal location for 

materials from the Rocky Flats Plant. Although most waste came to the RWMC in packages, 

accidents occurred during handling and processing that resulted in releases (Hoff, Mitchell, 

and Moore 1989). This in turn caused intakes of aged MFPs, uranium isotopes, transuranic 

(TRU) radionuclides, and aged mixed activation products (MAPs) (INEEL 2001).

• The Specific Manufacturing Capability (SMC) Project is a depleted uranium (DU) specialtyparts production plant built in 1985 in the ANP Program hanger on the Test Area North (TAN) 

portion of the site. The SMC Project processes metric tons of DU metal for the production of 

military shielding units (Stacy 2000). The processes of cutting, machining, and handling 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 16 of 63

uranium metal produce environments in which both chronic and accidental intakes of DU have 

occurred.

• The U.S. Navy used the Naval Reactors Facility (NRF) for operating reactors and as a naval 

reactor training center. Because this is not a DOE program and not under the oversight of 

DOE, the NRF is not part of the dose reconstruction and compensation program. However, 

through the years the NRF has participated in limited coordination of radiological protection 

programs and site support services. It is possible that some INEL workers received internal 

doses from their support of work at the NRF [11].

5.1.4 Radionuclides of Concern and Solubility

The INEL’s facilities and activities have related primarily to experimental reactor design and 

development, irradiated fuel processing, DU parts production, and low- and high-level radioactive 

waste treatment and disposal. The Technical Basis Document for the Idaho National Laboratory –

Site Description (ORAUT 2007a) describes these activities in more detail. 

Table 5-3 lists the INEL’s radionuclides of concern from these programs and as documented in INEEL 

M&O Contractor Technical Basis for Internal Dosimetry, General Technical Basis and Facility Specific 

Documents (INEEL 2001). These radionuclides are those for which internal doses were determined 

in the past and/or for which detection methods were developed. 

The INEL Site stored and processed nuclear materials from all over the world (ORAUT 2007a). 

These materials included, but were not limited to, nuclear wastes from other DOE sites and irradiated 

nuclear fuels from universities, test reactors, commercial power plants, INEL reactors, and U.S. 

Department of Defense (DOD) Projects. Because of this, it is not possible to limit the material types 

to a single selection when there are multiple possibilities identified by the International Commission on 

Radiological Protection (ICRP). In addition, materials change over time as they age, typically 

becoming more insoluble. Therefore, all material types included in ICRP Publication 68 (ICRP 1995), 

as well as super-S plutonium, for a given radionuclide should be considered when performing a dose 

reconstruction and the one yielding the highest dose should be assigned with the following 

exceptions.

• Because the available information on the INEL Site does not indicate that strontium titanate 

(SrTiO3) was ever present at the site and because strontium titanate was an uncommon 

strontium compound, strontium only needs to be assessed as type F material.

• With the exception of I-131 intakes calculated using the approach described in the Fission and 

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

Analyses (ORAUT 2007c), intakes of radioactive iodine need only be assessed as type F 

material using the inhalation intake route in the IMBA program. Because the various intake 

route and material type options for iodine in the IMBA program result in relatively insignificant 

dose differences and the inhalation of type F material results in a bounding dose estimate, it is 

reasonable to limit all assessments of radioactive iodine intakes to the inhalation of type F 

material. When using the approach described in the Fission and Activation Product 

Assignment for Internal Dose-Related Gross Beta and Gross Gamma Analyses, the intake 

route and material type options specified for iodine intakes in that document shall be used.

• Radionuclides where decay products or trace atoms are bound in a matrix of another 

radionuclide. Examples of this include Sr-90 and its decay product Y-90, and trace atoms of 

Am-241 in a plutonium matrix. In these examples, the Y-90 tracks with the type F Sr-90 even 

though Y-90 is only associated with type M and S materials, and Am-241 will track with type S 

plutonium even though Am-241 is only associated with type M material (ORAU 2007b).
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 17 of 63

• If the assessment of the bioassay data provides an inarguably better fit for a specific material 

type, only that material type need be used for the dose assessment (ORAUT 2007b).

Table 5-3. Primary radionuclides of concern.

Element Radionuclides

Hydrogen H-3 [assume tritiated water vapor (HTO)]a

Chromium Cr-51

Manganese Mn-54

Iron Fe-59

Cobalt Co-58, Co-60

Zinc Zn-65

Strontium/Yttrium Sr-89, Sr/Y-90

Zirconium/Niobium Zr/Nb-95

Molybdenum Mo-99

Technetium Tc-99

Ruthenium Ru-103, Ru-106

Silver Ag-110m

Antimony Sb-122, Sb-125

Tellurium Te-132

Iodine I-129, I-131, I-132, I-133, I-135

Cesium Cs-134, Cs-137

Barium/Lanthanum Ba/La-140

Cerium Ce-141, Ce-144

Europium Eu-152, Eu-154, Eu-155

Gadolinium Gd-153

Tantalum Ta-182

Mercury Hg-203 (assume inorganic)a

Protactinium Pa-233

Uranium U-233, U-234, U-235, U-236, U-238

Neptunium Np-237

Plutonium Pu-238, Pu-239/240

Americium Am-241

a. Based on information provided in the document titled INEEL M&O 

Contractor Technical Basis for Internal Dosimetry (INEEL 2001).

5.1.5 Intake Modes and Particle Size Distributions

Unless specific information is provided, the default guidelines regarding intake mode and particle size 

distribution in the Technical Information Bulletin: Internal Dose Reconstruction (ORAUT 2007b)

should be followed. 

For the SMC Project, specific particle size information is provided in Section 5.5.5.

5.1.6 Internal Dosimetry Program

The radiological protection program was established to provide timely detection of barrier or 

ventilation failure. The program consisted of continuous and retrospective air and effluent monitoring 

combined with personnel and surface contamination monitoring [12]. Detection of barrier failure 

provided the information for making decisions on evacuating personnel, increasing personnel 

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

As a consequence of consistent policy to avoid detectable internal exposures, coupled with the time 

and technical complexity of an internal dose evaluation, the general policy at the INEL for internal 

exposures has been preventive in nature [13]. In general, radiological materials handled at the site 

were of relatively low volume and mass and of higher activity concentration rather than metric tons of 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 18 of 63

materials of low specific activity. The consistent INEL policy and practice was to require respiratory 

protection on jobs when the possibility of airborne contamination was thought to exist regardless of 

the actual measured air or surface contamination (ACC 1952).

The changes in contractors at the INEL during its 60 plus year history resulted in relatively frequent 

management changes at most of the facilities. However, the contract with the University of Chicago 

to operate the ANL-W facilities did not change until 2005. Summaries of the contractor changes are 

provided in Tables 2-1 and 2-2 of the Technical Basis Document for the Idaho National Laboratory –

Site Description (ORAUT 2007a). 

The primary oversight for the INEL, which included most projects on the site and all support functions, 

was initially assigned to the AEC Idaho Operations Office (IDO) [14]. The AEC created the H&S 

Laboratory to provide a variety of health and safety support functions to the entire site, which included 

external and internal dosimetry, health physics instrumentation, fire protection, medical services, and 

environmental surveillance [14]. Once the NRF was constructed, the U.S. Navy provided oversight for 

that portion of the INEL Site. In addition, the Chicago Operations Office (COO) provided oversight for 

ANL-W programs and facilities that were contracted by the University of Chicago, until February 2005. 

During that period, the ANL-W used site support services, including internal dosimetry support, but the 

contractor reported the dosimetry results to the COO versus the IDO [14]. In February 2005, 

oversight of the INEL and ANL-W programs and facilities were combined under the same primary IDO 

contract.

The INEL personnel dosimetry records have been and are documented and permanently maintained 

by the various organizations throughout the site’s history [15]. Records about individual facility or 

contractor field monitoring programs (air-monitoring data, personnel contamination records, etc.) were 

maintained by individual contractors and/or site areas and are not maintained in a single 

recordkeeping system. The field monitoring data were not available for use in this TBD. 

In spite of the frequent changes in operational responsibility through the years and the movement of 

workers among facilities, there has been a basic level of consistency in the internal dosimetry 

programs at the INEL, particularly the bioassay analytical techniques and calculation processes [16]. 

The field programs monitored the workplace and identified work groups to be included in the routine 

bioassay programs and workers who needed special bioassays. Although these programs were 

implemented by the individual contractors, there was routine interaction with the H&S Laboratory 

professionals in interpretation of dosimetry results as well as in determination of necessary corrective 

practices or procedures [17]. 

Employees were typically assigned to individual facilities and were monitored for specific radiological 

hazards associated with the work. During periods when a single prime contractor was responsible for 

programs at most facilities or for site-wide support personnel, workers in certain crafts (e.g., 

maintenance, specialty operators, and some health physics technicians) worked at several facilities 

and were exposed to a variety of radioactive materials in a variety of work situations [18]. 

Internal dose reconstruction for personnel who worked at a number of the INEL facilities should rely 

on specific bioassay data (radionuclides, quantities, etc.) when available. The procedures and 

technical capabilities for collecting and analyzing bioassay samples at the different facilities were 

basically equivalent [19]. In addition, both the individual facilities and the H&S laboratories 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 (Bhatt 2002).
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5.1.6.1 Contamination Control

The contamination control limits for the detection and control of released activity beyond the control 

boundaries related to instrumentation capabilities and the basic philosophy of acceptance of 

detectable contamination. 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 personnel were always set 

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

evaluations and actions. Beta/gamma MDAs typically were a factor of 5 below the limits. Table 5-4 is

a summary of control limits primarily from an early health physics manual for the ICPP (ACC 1952) 

and current operating procedures.

Table 5-4. Surface contamination control and MDAs. 

Period

Surface 

location

Detection 

technique Control levels Typical MDA

1952–

1960s

Plant/equipment Smears 500 dpm β and

20 dpm α per 100 cm2

150 dpm β and

10 dpm α per 100 cm2

Personal 

clothing

Portable survey 

instruments

1500 dpm β and

500 dpm α per 100 cm2

1,000 dpm β and

500 dpm α per 100 cm2

Personal skin Portable survey 

instruments

Any detectable reported, 

e.g. 1,000 dpm β and

500 dpm α per 100 cm2

1,000 dpm β and

500 dpm α per 100 cm2

Shipments Smears 500 dpm β and

20 dpm α per 100 cm2

150 dpm β and

10 dpm α per 100 cm2 

Portable survey 

instruments

0.1 mrep/hr β and

500 dpm α per 100 cm2

0.01 mrep/hr β and

500 dpm α per 100 cm2

1970s–

present

Plant/equipment

surfaces

Smears 300 dpm β and

20 dpm α per 100 cm2

30 dpm β and

10 dpm α per 100 cm2

Personnel Portable survey 

instruments

Any detectable reported, 

e.g. 300 dpm β and

200 dpm α per 80-100 cm2

300 dpm β and

200 dpm α per 80–100 cm2

5.1.6.2 Air Monitoring

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

exposure prevention program. Beta/gamma-detecting CAMs were used from the beginning of all 

facility and program operations in routinely occupied areas. With the exception of the SMC Project

that was started in 1985, the primary contaminant radionuclides by activity were either MFPs or 

MAPs, which were beta/gamma emitters with maximum permissible concentrations/derived air 

concentrations (MPCs/DACs) above 1E-09 μCi/cm3

. TRU materials and uranium were available at 

some of the INEL facilities, but they were nearly always well-tagged with beta/gamma-emitting 

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

contamination or internal exposures.

The CPP Health Physics Manual from 1952 (ACC 1952) describes a CAM and three other airsampling systems. The manual required use of a filter-type respirator when airborne activity 

exceeded 1E-08 μCi/cm3 for beta/gamma activity or 1E-11 μCi/cm3 for alpha activity (ACC 1952). An 

army assault-type mask was required when levels exceeded this by a factor of 10. Positive-pressure 

air masks were required if levels larger by a factor of 1,000 occurred (ACC 1952).

The CAM systems provided real-time air activity evaluations (although it is not clear what the set 

points for alarms were), and fixed air samplers at several locations provided retrospective data and an 

average air concentration of beta/gamma emitters in an area or building [20]. The fixed air filter 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 20 of 63

samples were counted for both beta and alpha activity. Later, alpha CAMs were provided in select 

facilities where alpha contaminants could be controlling [21]. CAMs were calibrated, and training 

programs for health physicists were established for interpreting CAM responses for such variations as 

situations, radionuclides, response times, and filter accumulations [1]. If personnel were required to 

work in an area or building where known air contamination was present, respirators were worn to 

reduce internal contamination intake to levels below detectable amounts [1].

In general, workers were asked to submit to bioassay measurements whenever they were in an area 

where a CAM alarm sounded. In addition, the fixed location and retrospective air-sampling system 

signaled the need for bioassay if elevated air sample results were detected [22]. 

5.1.6.3 Bioassay Monitoring

Routine bioassay of radiation workers has occurred since the beginning of site operations. However, 

formal documentation of the bioassay programs was not found for periods before 1981. Some of the 

data sheets on individuals indicate that bioassay sampling occurred routinely every 6 months in 1953 

[23]. Table 5-5 lists the reconstructed history of routine bioassay frequency. The INEL’s monitoring 

and analytical programs were also designed to initiate an investigation of any potential internal intake 

as indicated by off-normal workplace indicators such as positive air sampling, personnel 

contamination, etc (INEEL 2001). Those investigations often included performing non-routine 

bioassay measurements for the potentially exposed workers. In addition, most of the recorded 

bioassay analyses performed at the INEL did not result in detectable radionuclides (ORAUT 2010a, 

2010b).

Table 5-5. Routine Bioassay History Summary.a

Year

Typical 

frequency Type Groups analyzed/sampled Investigating level Comments Reference

1953–

1960

Annual In vitro urine Radiation workers Unknown Frequency is 

inferred from 

individual data 

sheets.

Individual 

data sheets; 

Table 5-10; 

Horan 1959; 

AEC 1961

1961 Annual In vitro urine; in 

vivo

Radiation workers Unknown Frequency is 

inferred from 

individual data 

sheets.

Table 5-10; 

Horan 1962

1962–

1972

Annual In vitro urine; in 

vivo

Radiation workers Unknown Frequency is 

inferred from 

individual data 

sheets.

Horan 1962

Dodd 1963

1973–

1981

Annual In vitro urine Radiation workers Reporting

Annual DE >10% quarterly 

standard in ERDA Manual 

Chapter 0524 (ERDA 1975).

Frequency is 

inferred from 

individual data 

sheets.

AEC 1968;

AEC 1975; 

ERDA 1975

When internal 

intake 

suspected

Fecal

Annual In vivo

1982–

1987

Annual In vitro urine CPP-603 workers; fuel 

reprocessing operators

Reporting

50-yr CDE >10% quarterly 

standard in ERDA Manual 

Chapter 0524 (ERDA 1975).

Staggered to 

monitor group 

throughout the year.

Author 

unknown

When internal 1981

intake 

suspected

Fecal Waste reprocessing 

operators; shift laboratory 

workers; health physics 

technicians

Annual In vivo Selected radiochemistry 

workers; maintenance 

workers; denitrator operators.

Termination In vivo All radiation workers.
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Year

Typical 

frequency Type Groups analyzed/sampled Investigating level Comments Reference

1988–

1989

1 to 6 mo In vivo All radiation workers. Investigating

Lung, 50-yr CDE 

>0.5–1.0 rem.

Bone surface, 50-yr CDE 

>1.0–2.0 rem.

Other organs, 50-yr CDE 

>0.5-1.0 rem.

Staggered so that 

worker received

some sort of 

analysis or 

sampling every 3 

mo.

Tschaeche 

Annual In vitro fecal 1988

18 to 24 mo In vitro urine

Termination In vivo When internal exposure 

In vitro suspected.

New hire In vitro Depending on review of 

In vivo radiation dose history.

1990–

1994

Annual In vivo All radiation workers where

exposure to surface or 

airborne radioactive 

contamination could give at 

least 0.1-mrem AEDE from 

occupational sources, or give 

an organ or tissue DE >5 rem 

annual.

Reporting

In accordance with DOE 

Order 5480.11 (DOE 1988). 

Workers who could receive 

0.1 rem AEDE or 5 rem ADE 

organ or tissue dose.

Bioassay requested 

when workplace 

monitoring program 

indicated >0.02 

annual limit of 

intake.

Follow-up triggered 

by positive results 

from the workplace 

monitoring program, 

positive routine 

bioassay sample, or 

in response to 

incidents involving 

suspected intakes.

King 1990; 

6 mo In vitro fecal Rich 1990

Annual In vitro urine

New hire In vivo Worked at a facility where 

gamma-emitting 

radionuclides were handled.

Investigating

AEDE ≥0.01 rem.

In vitro urine; 

fecal

Worked in U manufacturing 

or recovery facilities; worked 

with transuranic materials.

Termination In vivo; in vitro Any employee suspected of 

having an internal exposure 

or on a scheduled monitoring 

program.

1995 Appropriate to 

the facility 

mission, 

potential 

uptakes.

In vivo All radiation workers who

enter radiological buffer 

areas or areas of higher 

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.

Reporting

In accordance with DOE 

5480.11 (DOE 1988) and 

10 CFR Part 835: 

Workers who could receive 

0.1 rem CEDE.

Declared pregnant workers 

when embryo/fetus could 

receive 0.05 rem DE.

Each facility had a 

specific Technical 

Basis Document for 

Internal Dosimetry.

Andersen, 

Perry, and 

Ruhter 1995

When 

workplace 

monitoring 

indicated 

significant 

potential for 

intakes.

In vitro urine; 

fecal

Random sampling was

performed to demonstrate the 

adequacy of the radiological 

controls in limiting the internal 

intake of radionuclides. 

Employees were selected at 

random from both nonradiation workers and a 

radiation worker population.

Investigating

Internal doses resulting from 

all confirmed intakes were to 

be evaluated.

Follow-up for any 

suspected intake of 

radionuclides and to 

more accurately 

identify and 

characterize the 

amount of intake 

and excretion 

pattern.

1995–

2000

New hire Based on screening to 

determine internal conditions 

from previous uptakes or to 

establish baseline for those 

continuing to work as 

radiation workers.

Termination In vivo Any employee who was on a 

scheduled monitoring 

program.

In vitro

2001 As developed 

by individual 

facilities based 

on analysis 

tables 

developed for 

each 

radionuclide.

In vivo All radiation workers. Reporting

In accordance with DOE 

5480.11 (DOE 1988) and 10 

CFR Part 835.

Workers who could receive 

0.1 rem CEDE.

Declared pregnant workers 

when embryo/fetus could 

receive 0.05 rem DE.

Bioassay is 

mandatory when an 

employee or visitor 

is involved in an 

event where the 

internal uptake of 

radionuclides was 

likely to have 

occurred.

INEEL 2001

In vitro urine 

fecal

Termination In vivo Any employee who was on a 
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In vitro scheduled monitoring 

program.

Investigating

Uranium, >1.0 μg/L

In vitro activity detected >2σ.

In vivo >2.33 σ.

Default trigger levels 

exceeded. 

a. AEDE = annual effective dose equivalent; CDE = committed dose equivalent; ERDA = U.S. Energy Research and Development 

Administration. 

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

documented the analytical detection capability of the INEL in the early 1950s and 1960s. Internal 

monitoring programs were in place when facility operations began in late 1951. For example, during 

ANP Program-IET activity in 1956, particulate and liquid caustic filter samples of effluent were 

analyzed with gamma spectroscopy and specific chemical separations of the identified radionuclides 

(Ebersole 1956). This analytical capability to identify radionuclides by their 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.

In the early days a gross beta 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 strontium). The gross beta assay used a 5 mL volume, but that was 

replaced with the gross gamma assay supplemented with strontium analyses that typically used a 75

mL volume [24]. 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.

In 1958, the IDO HSD acquired a 256-channel gamma spectrometer with a 3x3 (3 inch by 3 inch)

sodium iodide thallium-doped [NaI(Tl)] detector counting system for analyses of gamma-emitting 

radionuclides. In 1960, the HSD obtained a 3x3 well counter for gamma analysis, which replaced 

gross beta counting as the routine analytical procedure for urine samples. The Annual Report of 

Health and Safety Division 1960 (AEC 1961, p. 59) indicates that approximately 1.5E-06 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.

The Annual Report of Health and Safety Division 1960 (AEC 1961) outlined a basic philosophy in

relation to gamma counting of bioassay samples. Gamma counting would be effective in all situations 

except exposures 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 solution, gross beta counted, and the strontium 

analysis not completed unless a detectable count was obtained on the precipitate. A 100 mL sample 

of urine permitted the detection of approximately 8E-08 μCi/mL of Sr-90.

Special and routine bioassay measurements were performed and documented by the DOE analytical 

laboratory. Puphal (1994) reported on the procedures used for bioassay in the Analytical Chemistry 

Branch beginning in 1960. These procedures were collected and placed into a procedures manual in 

1982 for periodic revision (Bodnar and Percival 1982). There was another version of the procedures 

after the analytical work was transferred to Westinghouse Idaho Nuclear Company (INEEL 2002).

Table 5-6 reproduces reports of the urinalysis results for 1959, 1960, and 1961 as obtained from 

Table 4 of AEC (1960), Table IV of AEC (1961), and Table XIII of Horan (1962). These results are not 

identical but quite similar to those from the newer database. The practice was to perform a gross beta 

or gamma analysis and identify specific radionuclides if the gross counts indicated activity above 

background levels. The total number of urinalyses in 1959 was 11,066; 3,524 people had radiation 

badges; and 715 received external doses above 500 mrem. These numbers demonstrate that 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 23 of 63

workers provided urine samples multiple times during the year. The 1963 annual summary report 

(Dodd 1964) describes the year’s follow-up analyses for the whole-body counting program. As shown 

in Table 5-7, many of the individuals were counted multiple times. 

Images of the INEL’s urine sample results through 1986 and in vivo measurements results through 

early 1996 were provided to the Oak Ridge Associated Universities Team (ORAUT), and the results 

were entered into a database. This database has more than 140,000 in vitro bioassay measurement 

results and over 95,000 in vivo bioassay measurement results (ORAUT 2010a, ORAUT 2010b). Of 

the in vitro bioassay measurements, over 61% of them were gross beta or gross gamma radioactivity 

in urine measurements (ORAUT 2010a). Of the in vivo bioassay measurements, over 70% of them 

were whole body counts (ORAUT 2010b). Prior to 1961, the predominant type of bioassay 

measurement being performed at the INEL was urine sampling with a gross beta radioactivity 

analysis. In December 1960, the gross beta radioactivity analysis was discontinued entirely (AEC 

1961). Starting in 1961, the gross gamma radioactivity analysis replaced the gross beta radioactivity 

analysis as the predominant type of analysis performed on urine samples. After 1960, urine samples 

were still being periodically analyzed for strontium radioactivity; however, the strontium analysis was 

never the predominant analysis performed on the urine samples (ORAUT 2010b). In 1961, whole 

body counting was introduced and was often performed along with in vitro bioassay, until the early 

1970’s. In the early 1970s, in vitro bioassay was largely replaced by whole-body counting [1]. 

Significant intakes of radioactivity were uncommon at the INEL. This determination is supported by 

an evaluation of the bioassay data in ORAUT’s database. Of the more than 89,000 urine samples 

that were analyzed for gross beta radioactivity and gross gamma radioactivity, less than 2% of those 

measurements were above the MDA values provided in Table 5-14 (ORAUT 2010a). Of more than 

69,000 whole body counts, which is typically a more sensitive measurement than urinalysis, less than 

10% of those measurements were reported as having results above their respective detection levels 

(ORAUT 2010b). In addition, over 92% of the whole body counts results that were above their 

respective detection levels were still below the 0.1 μCi reporting level (ORAUT 2010b), which 

indicates that large intakes were uncommon at the INEL. 

During the earlier years at the INEL, workers were not routinely bioassayed for plutonium. Because 

plutonium was not separated from the spent nuclear fuel at the INEL, the plutonium was always 

present with the more readily detectable mixed fission products that were also in the irradiated fuel.

Therefore, in the vast majority of the plutonium exposure scenarios, the plutonium would have been 

present with the product and waste streams containing mixed fission products, and any intakes of 

radioactivity would have been more readily detectable by performing bioassay measurements for 

mixed fission products. The available bioassay data indicates that certain ANL-W and ICPP workers

started receiving routine plutonium bioassays starting around the late 1970’s.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/0

2/2010 Page 24 of 63

Table 5-6. Urinalysis results in 1959, 1960, and 1961 (AEC 1960, 1961; Horan 1962).

Nuclide/

element of 

interest

Type 

activity

Total number performed

Statistically significant

Highest resulta Number Percent

1959 1960 1961 1959 1960 1961 1959 1960 1961 1959 1960 1961

Gross βb 8,546 8,546 30 65 15 5 0.76 0.18 17 18,820 ±632 

dpm/5 mL

992 ±40 

dpm/5 mL

172 ±16 dpm/mL

Gross γc 2,433 2,712 9,120 174 129 --d 7.15 4.76 4 35,972 ±310 

dpm/5 mL

19,817 ±105 

dpm/5 mL

1,900,235 ±876 

dpm/75 mL

Co-60 β -- -- 1 -- -- 0 -- -- 0 -- -- 300 ±75 

dpm/450 mL

Sr-90 β 3 105 3,248 3 0 2 100 0 2 4.12E-2 

dpm/mL

Insignificant 183 ±8 dpm/75 

mL

Sr-91 β 20 37 2 19 0 -- 95 0 0 388 ±1.6 Insignificant 4 ±8 dpm/mL

I

e β -- 9 -- -- 2 -- 22 0 9992 ±80 

dpm/mL

--

Cs-137 β -- -- 40 -- -- 0 -- -- 0 -- -- 1,460 ±10% 

dpm/1700 mL

Ba-139 α 20 16 0 -- 80 0 -- 120 ±0.8 

dpm/mL

0 --

Thf α 7 0 -- 0 0 -- 0 0 -- Insignificant 0 Insignificant

U α -- -- 4 -- -- -- -- -- 0 -- -- 10 μg/L

U-233 α 17 3 -- 1 0 -- 0.06 0 -- 180 ±4.0 

dpm/mL

Insignificant --

Pu-239 α 18 0 29 0 0 0 0 0 Insignificant Insignificant 2E-9 μCi/mL

Am-241 α 2 0 -- 0 0 -- 0 0 -- Insignificant 0 --

Totals 11,066 11,352 12,494 278 146 -- 2.51 1.29 4

a. All except two I-131 exposures in 1961 listed under gross gamma activity are less than 10% of the permissible body burden for the radionuclide of interest.

b. If only gross β analyses are available, the default should be Sr-90. 

c. In many instances positive gross gamma results have been attributed to radioactive isotopes of iodine.

d. -- = no data reported.

e. Iodine isotope(s) not identified in references. Assume I-131.

f. Thorium isotope(s) not identified in references. Assume Th-228.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 25 of 63

Table 5-7. Summary statistics from the 1963 WBC program (Dodd 1964, p. 19).

Radionuclide Times reported Number of individuals Maximum activity (μCi)

Cr-51 15 10 1.2

Co-60/Fe-59 848 387 1.5

Mn-54 98 51 0.16

Co-58 62 50 0.03

Zn-65 505 171 1.20

Zr/Nb-95a 427 232 1.66

Ru-103/Ru-106b 93 75 0.22

Ag-110m 583 186 0.93

Sb-122 2 2 0.08

I-131 110 82 5.0

Cs-134 361 168 0.14

Cs-137 2,332 573 1.32

Ba/La-140 90 51 0.07

Ce-141/Ce-144 59 49 0.16

Ta-182 50 36 0.02

Hg-203 28 6 0.16

Pa-233 13 10 0.48

Np-239 1 1 1.68

Sb-125 3 3 0.1

Mo/Tc-99 8 5 0.72

I-132 8 7 <0.1

I-133 3 3 <0.1

Te-132 6 6 <0.1

Hg-197 7 3 0.7

a. Consider Zr-95.

b. Consider Ru-106.

5.1.7 Recordkeeping

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

until 1989 when all technical support service functions, including those related to internal dosimetry, 

were transferred to the INEL Site’s prime contractor. At that time, in vitro analytical functions were 

transferred to an onsite analytical laboratory. The in vivo counting laboratory provides support directly 

through the Radiation Dosimetry and Records (RDR) organization, which administers external and 

internal dosimetry support programs. The current contractor’s subject matter expert reviews, 

validates, and prepares official internal dose assessments. A DOE staff member at RESL is 

responsible for oversight of the INEL’s internal dosimetry program functions and provides quality 

assurance. The RDR unit functions include documentation and records custodial responsibilities. In 

1999, the Radiation Control Information Management System (RCIMS) database was placed in 

service to support the radiation protection program, including internal dosimetry. RCIMS lists reported 

internal doses as CEDE when an individual’s dose history is prepared (INEEL 2001).

The following information is important to internal dose reconstruction because the worker files from 

DOE can contain a variety of internal dose information including the calculated internal doses as well 

as the in vitro and in vivo individual bioassay results. The changing 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.

The information used in internal dose assessments and analytical data sheets has varied through the 

years. Table 5-8 describes internal dose information that could appear in pre-1989 records. Table

5-9 describes coded information that could appear in records after 1989. Tables 5-10 and 5-11 

provide the column descriptions for the sample record sheets that were used for the urine sample 

data. Column descriptions are provided for the sample record sheets used before February 1955 and 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 26 of 63

for the sheets used during the period of February 1955 and later. Table 5-12 provides the INEL’s 

location codes for various work areas.

Table 5-8. Internal dose assessment information before 1989.

Dose information Description

Name, Social Security # 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 by month and year in which 

exposure occurred.

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

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

CEDE rem Calculated CEDE in rem.

AEDE rem Calculated AEDE in rem.

Year Year for which the AEDE was calculated.

Table 5-9. Internal dose assessment information after 1989.

Coded information Description

Name & Social 

Security #

Exposed employee by name and Social Security number.

Asmt. nos. This assessment number is the calendar year (e.g., 83) 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.

AEDE rem Calculated AEDE in rem.

Year Year for which the AEDE was calculated.

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 exposures are summed for the year of intake for each employee.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 27 of 63

Table 5-10. Column descriptions for urine sample record sheets before February 1955.

Analytical 

information Description

Sample no. Sample log number.

Date and hour Generally clear interpretation.

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

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

Sampling data This column is subdivided in to three separate columns: Rate, Time, and Quant. The 

columns appear to be applicable to air sampling data and are not used for urine 

samples.

Anal. for Means analyzed for. Generally gross beta-gamma and/or gross gamma. The gross 

beta-gamma analyses are typically denoted by either a “beta-gamma”, “beta”, “B”, “B-I”,

“β”, or “βγ” in this column. The gross gamma analyses are typically denoted by either a 

“gamma”, “I”, or “γ” in this column. Note that an “I” in this column does not indicate an 

iodine analysis unless specifically labeled as such. For gross beta-gamma analyses, a 

5 ml sample aliquot was typically evaporated and counted using an end-window 

Geiger-Mueller detector. For the gross gamma analyses, a 50 ml sample aliquot was 

typically counted directly in a deep-well NaI scintillation detector. Identifiers for specific 

isotopic analyses, based on chemical separations or gamma spectrometry were also 

listed in this column.

Quantity used Size of the sample aliquot in whatever units are indicated. Typically, the aliquot sizes 

are provided in ml.

Date or time For urine samples the header for this column is usually crossed out and replaced with 

“(K+

) Trans”, which indicates that an aliquot of the sample was also analyzed for K-40. 

When this is the case, the reported value in this column is the percent transmission 

reading from a flame photometer analysis that was performed on an aliquot of the 

sample that was analyzed for its potassium content. The percent transmission value is 

then used to determine the amount of the radioactivity in the urine sample that is 

attributable to naturally occurring K-40. However, before February 1955, the sample

results appear to have never been corrected for the radioactivity that was attributable to 

K-40.

Count time Counting methods used either preset times or preset counts. The sample count time 

was typically recorded in minutes in either case. For some early gross beta-gamma 

analyses, the sample count time was recorded in seconds.

Inst. read. This is the reading from the counting instrument which is recorded in the units 

indicated. Typically, this is the total number of counts recorded for the sample.

Corr. This column is used for the background count rate. A 1σ counting error is sometimes 

also reported with the background count rate.

Corr. read. This column is used for the net count rate, which is calculated by subtracting the 

background count rate from the calculated gross count rate.

Mass equiv. This column is often used to report the amount of K-40 in dpm that the sample 

contains. Before February 1955, there is no indication that the sample results were 

corrected to account for the K-40 activity. 

Corrected value This column is used to report the net counting results in units of activity, usually dpm. 

When the results are reported in dpm, it should be assumed that they are actually 

representing units of dpm/sample, unless indicated otherwise.
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Table 5-11. Column descriptions for urine sample record sheets from February 1955 and later.

Analytical 

information Description

Sample no. Sample log number.

Date and hour Generally clear interpretation.

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

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

Anal. for Means analyzed for. Generally gross beta-gamma and/or gross gamma. The gross 

beta-gamma analyses are typically denoted by either a “beta-gamma”, “beta”, “B”, “B-I”, 

“β”, or “βγ” in this column. The gross gamma analyses are typically denoted by either a 

“gamma”, “I”, or “γ” in this column. Note that an “I” in this column does not indicate an 

iodine analysis unless specifically labeled as such. For gross beta-gamma analyses, a 

5 ml sample aliquot was typically evaporated and counted using an end-window 

Geiger-Mueller detector. For the gross gamma analyses, a 50 ml sample aliquot was 

typically counted directly in a deep-well NaI scintillation detector. Identifiers for specific 

isotopic analyses, based on chemical separations or gamma spectrometry were also 

listed in this column.

Quantity used Size of the sample aliquot in ml, unless other units are otherwise indicated.

U+ or K+ trans. Indicates that an analytical correction for natural uranium or potassium might be 

applied to the recorded results. The value in this column is almost always associated 

with a correction for natural potassium and the natural uranium correction is relatively 

rare. The reported value in this column is the percent transmission reading from a 

flame photometer analysis that was performed on an aliquot of the sample. The 

percent transmission value is then used to determine the amount of the radioactivity in 

the urine sample that is attributable to natural uranium or potassium.

Count time Counting methods used either preset times or preset counts. The sample count time 

was typically recorded in minutes in either case. For some early gross beta-gamma 

analyses, the sample count time was recorded in seconds.

Total count Total number of counts recorded for the sample.

Gross count, cpm The gross count rate was determined by dividing total counts by the count time.

Bkgd., cpm Background count rate recorded. A 1σ counting error is sometimes also reported with 

the background count rate.

Net count, cpm The net count rate, which is calculated by subtracting the background count rate from 

the gross count rate.

K-40 corr., cpm Starting in 1955 and when the value in this column is reported in cpm, a K-40 

correction has likely been applied to the recorded analytical results. The K-40 

correction is applied by subtracting the value in this column from the recorded net count 

rate and then dividing the adjusted net count rate by the detector’s counting efficiency.

Foreign activity, 

cpm and dpm

Unless stated otherwise, the results in this column are reported in units of cpm and

dpm. Net count rate, which is sometimes corrected for K-40, and then converted to 

dpm based on the detector’s counting efficiency. Uncertainty also included, which is 

recorded based on 1σ counting statistics. When the results are reported in dpm, it 

should be assumed that they are actually representing units of dpm/sample, unless 

indicated otherwise. At times the dpm units are crossed out and the results are 

recorded in other units. The other units will then be specified, and are typically units of 

dpm/ml or μCi/ml. In some instances the units may be in μg/L, which indicate the 

results for a uranium analysis whether or not a uranium analysis was previously 

specified.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 29 of 63

Table 5-12. Location codes that could be in worker dose files. 

Location code Descriptiona Location code Descriptiona

1 AEC Headquarters Building 28 Onsite site survey

2 ANL-W - EBR-I 29 Offsite site survey

3, 34, 35 CFA 30 SL-1 - ANP program

4, 42, 45 MTR, TRA 31 STPF

5, 53, 55 ICPP 33 CFA Laundry

6, 32, 63, 64, 

65, 764, 765

NRF 66 Non-security

67 Division of Compliance

7 TAN - ANP program 68 STEP

8 Services 69 LPTF (Phillips and AEC)

9 NX (X is construction) at NRF 71 CADRE (guard force)

10, 105 AX at TAN 125 PBF construction

11, 113, 115 CX at ICPP 133, 135 PBF/WROC

12 EX at EBR 160, 161 IRC

13 SPERT 162 WCB

14 OMRE 163 EROB

15 SX at SPERT 164 TSA/TSB

16 SL-1 165 WAC

17, 333 MX at MTR 244, 245 ARA I

18, 814, 815 WP, RWMC 344, 345 ARA II

19, 772, 775 TAN (Phillips and AEC) 354, 355 ARA III

20, 261, 264 TREAT 555 Guards

21 LX 754/755 DOE – ID/RESL

22 GCRE 773 SMC (B&W cask)

23 OX 774, 776 SMC

24 ARHG 825 RWMC construction

25 No information available 835 RWMC storage

26, 263, 265 ANL-W 845 Pit 9

27 ML-1

a. See the acronyms and abbreviations list.

Federal regulations about permissible internal dose and formal reporting requirements to the AEC, 

ERDA, and DOE changed periodically during the site’s history. As the internal dose limits were 

reduced over time, the frequency of monitoring typically increases, new analytical methods are 

developed for existing types of bioassay to detect lower quantities, and new types of bioassay are 

developed and implemented (e.g., whole body counting, chest counting, routine fecal sampling, 

etc...). As a result, a site’s ability to detect intakes was progressively enhanced over time, because of 

the changes in the permissible levels of internal dose. During the early years internal dose was 

usually considered separately from external dose in terms of meeting specific exposure limits, and the 

calculated dose was reported and documented only if specific dose levels were exceeded (Aoki 

1979). AEC and ERDA policies required periodic urinalyses or in vivo counting or evaluation of air 

concentrations if the whole-body dose or committed dose could exceed 300 mrem in a calendar 

quarter (AEC 1958, 1963, 1968, 1975; ERDA 1975). 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) have also changed little from the early 

years to the present [22]. Dickson (1977) established official investigation levels (Table 5-13) for 

acute uptakes of radionuclides corresponding to one-tenth of the quarterly radiation standard. Later 

procedures (DOE 1988) set specific limits on those positive bioassay results that could result in 

100-mrem AEDE or above as the point at which follow-up and reporting was required. With the DOE 

Radiological Control Manual (e.g., DOE 1994), this changed to 100-mrem CEDE. In addition, a 

calculated dose of 10 mrem or above would be recorded as an internal dose (DOE 1994). 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 30 of 63

Table 5-13. Derived investigation levels in 1977 for acute 

exposures (Dickson 1977). 

Radionuclide

Inhalation lung burden

(μCi)

Ingestion total activity

(μCi)

Cr-51 20 500

Mn-54 0.4 30

Co-57 2 90

Co-60 0.09 9

Zn-65 0.6 30

Zr-95 0.3 20

Ru-106 0.06 3

Sb-125 0.3 30

Cs-134 0.1 3

Cs-137 0.1 4

Ce-144 0.06 3

Pu-239 Whenever detected

Am-241 Whenever detected

Sr-90 (bone) When detected by skull counting

I-131 (thyroid) Initial content 0.27 Not provided

5.1.7.1 Radiological Incident Records

When an incident occurred, it was the policy to investigate thoroughly and identify all individuals 

involved in the incident [26]. Therefore, when there is no evidence in the incident file or the 

individual’s dosimetry file that an individual was involved and no other evidence supporting that an 

individual was involved in the incident, dose reconstructors should assume that the individual was not 

involved. 

5.2 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

No bioassay data appear to have been collected before 1952; however, very few of the radiological 

facilities at the INEL were operational before 1952.

In compliance with the November 1998 Code of Federal Regulations requirement (10 CFR Part 835) 

for the DOE Laboratory Accreditation Program (DOELAP), and based on American National 

Standards Institute N13.30, Performance Criteria for Radiobioassay (HPS 1996), both the in vitro and 

in vivo radiobioassay laboratories at the INEL received DOELAP accreditation in February 1998 

(INEEL 2001, p 11). In accordance with this accreditation, MDAs and decision levels at the 95% 

confidence level are performed. Tables 5-14 through 5-17 and 5-18 through 5-21 list the current 

MDAs for the various bioassay methods employed at the INEL along with values from historical 

documents and the recommended periods of use for the MDA values.

5.2.1 Urine Sample Analyses

The majority of the urine samples taken at the site were single voidings; 24 samples were used for 

special sampling purposes (i.e., follow-up samples, primarily to extend the sensitivity). Urine sample 

results are typically reported in units of activity per sample along with the total sample and/or aliquot 

size. Table 5-14 provides the known MDAs for this in vitro bioassay method.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 31 of 63

Table 5-14. Urine sample MDAs by period.

Radiation/

radionuclide Period 

Typical 

volume 

(ml)

Typical MDA 

(dpm/sample)

Typical dailya

MDA (dpm/d) Reference

Gross β 1951–1953 5 86 24,000 Data Sheet

1954–1960 5 93b 26,000 Ebersole and Flygare 1957

Gross γ 1957–1964 75 580b 10,800 AEC 1961

1965–1971 75 205 3,800 Data Sheet

H-3 1972–1994 3 0.5 dpm/ml

b 700 AEC 1972, 1974

1995–present 3 dpm/ml 4,200 Andersen, Perry, and Ruhter 1995, 

INEEL 2001

Co-60 1957–1958 50 51 1,400 Database

Sr-90 1953–6/14/62 75 37 700 Database

6/15/62–1970 75 20 370 Database

1971–1989 75 1.7b 32 AEC 1972, 1974

1990–present 500 min 1.9 5 Andersen, Perry, and Ruhter 1995

I-131 1957–1970 75 370 6,900 Database

Cs-134 1974–present 400 2 7 Database

Cs-137 1961–present 400 410 1,435 Database

Th-230 1974–present 1,000 0.1b 0.14 AEC 1974

Np-237 1974–present 1,000 0.1b 0.14 AEC 1974

U (PF) 1954–1961 0.1 1E-5 g U/Lc 14 μg Uc Database

1962–1971 0.1 5E-6 g U/Lc 7 μg Uc Database

U (KPA) 1985–present 0.2 μg U/L 0.28 μg U Rich 1990

U-233/234 1979–1986 700 0.52 1.0 Database

1995–present 500 min 0.091 0.25 Andersen, Perry, and Ruhter 1995

U-235 1970–1979 1,000 0.22 0.31 Rich 1990

1980–1994 700 0.13 0.26 Database

1995–present 500 min 0.084 0.24 Andersen, Perry, and Ruhter 1995

U-238 1970–1979 1,000 0.22 0.31 Rich 1990

1980–1994 700 0.21 0.42 Database

1995–present 500 min 0.067 0.19 Andersen, Perry, and Ruhter 1995

Pu-238 1981–1984 700 0.072 0.14 Database

1990–1994 1,000 0.13 0.18 Rich 1990

1995–present 500 min 0.049 0.14 Andersen, Perry, and Ruhter 1995

Pu-239/240 1964–1970 1,000 0.93b 1.3 Dodd 1964

1971–1973 1,000 1.03b 1.4 AEC 1972

1974–1979 1,000 0.47b 0.66 AEC 1974

1980–1989 700 0.073 0.14 Database

1990–1994 1,000 0.060 0.084 Rich 1990

1995–present 500 min 0.060 0.17 Andersen, Perry, and Ruhter 1995

Am-241 1977–1979 1,000 0.16b 0.22 AEC 1974

1980–1989 700 0.29 0.6 Database

1990–1994 1,000 0.2 0.28 Rich 1990

1995–present 500 min 0.051 0.14 Andersen, Perry, and Ruhter 1995

Cm-244 1974–present 1,000 0.155b 0.22 AEC 1974

a. Based on 1,400-mL daily volume and typical sample size, and would need to be adjusted for different sample volumes. 

b. MDA calculated from an inferred 2σ uncertainty.

c. Smallest reported value. Not MDA.

5.2.2 Fecal Sample Analyses

At times, fecal samples were collected from workers to assess potential intakes of certain 

radionuclides. Table 5-15 provides the known MDAs for this in vitro bioassay method.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 32 of 63

Table 5-15. Fecal sample MDAs by period.

Radiation/

radionuclide Period

Fecala

(pCi/sample) Reference

Co-60 1963–present 10 Rich 1990

Sr-90 1963–1994 10 Rich 1990

1995–present 1.9 Andersen, Perry, and Ruhter 1995; INEEL 2001

Cs-134 1963–present 10 Rich 1990

Cs-137 1963–1999 0.01 Rich 1990

2000–present 0.3 BBI 2000

Th-230 1974–present 0.03 AEC 1974

Np-237 1974–present 0.03 AEC 1974

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

2003–present 0.05 Bhatt 2003

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

2003–present 0.09 Bhatt 2003

U-238 1970–1994 0.5 Rich 1990

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

2003–present 0.09 Bhatt 2003

Pu-238 1974–1994 0.03 AEC 1974

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

2003–present 0.02 Bhatt 2003

Pu-239/240 1964–1973 0.4b Dodd 1964

1974–1994 0.02 AEC 1974

1995–present 0.03 Andersen, Perry, and Ruhter 1995; BBI 2000; INEEL 2001; 

Bhatt 2002

Am-241 1974–1994 0.07 AEC 1974

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

2002–present 0.04 Bhatt 2002

Cm-244 1974–present 0.02 AEC 1974

Cf-252 1974–present 0.02 AEC 1974

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

b. MDA calculated from an inferred 2σ uncertainty.

5.3 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

5.3.1 Whole-Body Counting

5.3.1.1 General Information on Whole-Body Counting

Whole-body counting (WBC) was introduced at the INEL in 1961. As early as 1961 one of the 

fundamental conclusions from experience at the INEL with in vivo and in vitro internal dosimetry 

analytical techniques was that a large proportion of the internal exposures to the INEL workers were

to insoluble materials. Radionuclides (e.g. Sb-125, Ag-110m, Zn-65, and Zr/Nb-95) were detected by 

an in vivo count and not in the urine. Concurrent analyses of feces and urine demonstrated the main 

elimination route to be by feces, with so little voided in the urine as to be undetectable even in a 24

hour specimen (Horan 1962; Sill, Anderson, and Percival 1964). WBC was demonstrated to detect 

activity as low as 0.01 μCi in a 10 minute count (Horan 1962). This detection level was several orders 

of magnitude more sensitive than the maximum permissible body burdens (MPBBs) for most 

beta/gamma-emitting fission and activation products.

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

required physical examinations, (2) employees who were suspected of having a possible internal 

intake, and (3) selected groups of individuals who were scheduled for semi-routine analyses by health 

physics supervisors (Sommers 1961). In 1963 approximately 1,650 counts were performed; only 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 33 of 63

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.

Table 5-16. Whole-body count MDAs by period. 

Radiation/

radionuclide Period

In vivo 

MDA (nCi)

Count 

time 

(min) Reference

Cr-51 1961–2000 5 10 Percival and Anderson 1962

2001–present 32 5 INEEL 2001

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

1995

2001–present 2.6 5 INEEL 2001

2001–present 1.3 10 INEEL 2001

Fe-59 1962–2001 4.5 5 INEEL 2001

2001–present 1.5 10 INEEL 2001

Co-58 1961–2000 5 10 Percival and Anderson 1962

2001–present 2.5 5 INEEL 2001

2001–present 1.1 10 INEEL 2001

Co-60 1961–2000 5 10 Percival and Anderson 1962

1971–1988 5a 10 AEC 1972; 1974

1989–2000 7 10 Martin 1989; Grothaus 1993; Andersen, Perry and 

Ruhter 1995

2001–present 2.5 5 INEEL 2001

2001–present 1.1 10 INEEL 2001

Zn-65 1961–2000 5 10 Percival and Anderson 1962

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

Ruhter 1995

2001–present 4.9 5 INEEL 2001

2001–present 2 10 INEEL 2001

Zr/Nb-95 1961–2000 5 10 Percival and Anderson 1962

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

Ruhter 1995

2001–present 2.6 5 INEEL 2001

Ru-106 2001–present 27 5 INEEL 2001

2001–present 7.6 10 INEEL 2001

Ag-110m 1961–2000 5 10 Percival and Anderson 1962

Sb-125 1961–1988 5 10 Percival and Anderson 1962

1989–present 14 10 Martin 1989; Grothaus 1993

I-131 1961–2000 5 10 Percival and Anderson 1962

2001–present 3.8 5 INEEL 2001

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

Ruhter 1995

2001–present 3 5 INEEL 2001

2001–present 0.96 10 INEEL 2001

Cs-137 1961–1970 5 10 Percival and Anderson 1962

1971–1988 2 10 AEC 1972, AEC 1974, Tschaeche 1988

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

Ruhter 1995

2001–present 3.1 5 INEEL 2001

2001–present 1.9 10 INEEL 2001

Ba/La-140 2001–present 12 5 INEEL 2001

Ce-141 2001–present 9.9 5 INEEL 2001

2001–present 3.2 10 INEEL 2001
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 34 of 63

Radiation/

radionuclide Period

In vivo 

MDA (nCi)

Count 

time 

(min) Reference

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

Ruhter 1995

2001–present 44 5 INEEL 2001

2001–present 15 10 INEEL 2001

Eu-152 2001–present 4 10 INEEL 2001

Eu-154 2001–present 2 10 INEEL 2001

Eu-155 2001–present 1 10 INEEL 2001

Ga-153 2001–present 6.5 10 INEEL 2001

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

Ruhter 1995

Ta-182 1961–1962 5 10 Percival and Anderson 1962

Hg-203 1961–1962 5 10 Percival and Anderson 1962

a. MDA calculated from an inferred 2σ uncertainty.

5.3.1.2 Attributing Positive WBCs to Cs-137 Fallout

Fallout affected everyone in North America, and body burdens of Cs-137 measurable in the wholebody counters were common in the 1960s and 1970s. National Council on Radiation Protection and 

Measurements (NCRP) Report 94 (NCRP 1987) provides mean body burdens of Cs-137 for the 

United States for 1953–1977, which are summarized in Table 5-17.

Table 5-17. Mean body burdens of Cs-137 from fallout in the 

United States.

Year

Body burden

(nCi) Year

Body burden

(nCi)

1953 0.27 1966 9.7

1954 1.1 1967 5.6

1955 2.2 1968 3.5

1956 4.3 1969 2.7

1957 5.1 1970 2.7

1958 6.5 1971 2.7

1959 8.1 1972 2.7

1960 6.8 1973 2.7

1961 4.6 1974 1.6

1962 6.0 1975 1.1

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

a. Values in this table were obtained from NCRP Report 94(NCRP 1987). 

If WBC results show detection of only K-40 and Cs-137, and the Cs-137 result is less than the values 

in Table 5-17, the dose reconstructor can attribute the positive Cs-137 results to fallout and not assign 

doses based on those results. If other radionuclides are present during the same measurement time 

period in the WBC or are detected by another bioassay method, it should be assumed that the Cs-137 

is entirely from occupational sources. The values in Table 5-17 are only intended to attribute an entire 

WBC’s result to fallout Cs-137 and are not intended to reduce the calculated intakes by subtracting 

the contribution attributable to fallout Cs-137. Therefore, the values in Table 5-17 should not be 

subtracted from WBC results that are greater than the Table 5-17 values.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 35 of 63

5.3.2 Lung Counts

At times, lung counts were performed on workers to assess potential intakes of certain radionuclides. 

Table 5-18 provides the known MDAs for this in vivo bioassay method.

Table 5-18. Lung count MDAs by period. 

Radionuclide Period

In vivo

MDAa (nCi)

Count 

time (min) Reference

Ce-141 2001–present 0.11 60 INEEL 2001

Ce-144 2001–present 0.44 60 INEEL 2001

Eu-152 2001–present 0.18 60 INEEL 2001

Ga-153 2001–present 0.096 60 INEEL 2001

Th-234 2001–present 1.4 60 INEEL 2001

U-235 1990–2000 0.2 Rich 1990

2001–present 0.11 60 INEEL 2001

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

Pu-238 1984–1988 16 Tschaeche 1988

1989–1995 26 60 Martin 1989; Rich 1990; Grothaus 1993; 

Andersen, Perry, and Ruhter 1995

1996–present 54 60 INEEL 2001

Pu-239/240 1974–1983 74 100 AEC 1974

1984–1988 20 Tschaeche 1988

1989–1995 80 60 Martin 1989; Rich 1990; Grothaus 1993; 

Andersen, Perry, and Ruhter 1995

1996–present 140 60 INEEL 2001

Am-241 1984–1988 0.15 Tschaeche 1988

1989–1995 0.6 60 Martin 1989; Rich 1990; Grothaus 1993; 

Andersen, Perry, and Ruhter 1995

1996–present 0.14 60 INEEL 2001

a. If an MDA value is not available in this table, it can be inferred from the reported 1σ uncertainty as being approximately 

equal to the 2σ value.

5.3.3 Thyroid Counts

At times, thyroid counts were performed on workers to assess potential intakes of radioactive iodine. 

Table 5-19 provides the known MDAs for this in vivo bioassay method.

Table 5-19. Thyroid count MDAs by period. 

Radionuclide Period

In vivo 

MDA (nCi)

Count time 

(min) Reference

I-131 1971-1974 1E-05 10 AEC 1972, 1974

1990–1992 2 10 Rich 1990

1993–2000 0.3 10 Grothaus 1993

2001–present 0.13 INEEL 2001

5.3.4 Skull Counts

At times skull counts were performed to assess potential intakes of Sr/Y-90. The skull counts 

measured Bremsstrahlung radiation that is emitted from the skull after the bone has had an uptake of 

radioactive strontium. Table 5-20 provides the MDAs for this in vivo bioassay method.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 36 of 63

Table 5-20. Skull count MDAs by period. 

Radionuclide Period

In vivo 

MDA (nCi)

Count time 

(min) Reference

Sr/Y-90 1968–1977 70a 10 Voelz 1969; AEC 1972, 1974

1978–present 34 10 Martin 1989; Grothaus 1993

a. MDA calculated from an inferred 2σ uncertainty.

5.3.5 Wound Counts

Wound counts have been performed at the INEL to assess potential intakes from contaminated 

wounds. Table 5-21 provides the known MDAs for this in vivo bioassay method.

Table 5-21. Wound count MDAs by period. 

Radionuclide Period

In vivo 

MDA (nCi)

Count time

(min) Reference

U-235 1993 0.2 20 Grothaus 1993

Pu-238 1993 1 20 Grothaus 1993

Pu-239/240 1993 2 20 Grothaus 1993

Am-241 1993 0.1 20 Grothaus 1993

5.4 INTERFERENCES AND UNCERTAINTIES

5.4.1 Interferences

5.4.1.1 Contamination of Samples

The most common type of interference encountered with bioassay measurements is due to 

contaminated in vitro samples and contaminated workers for in vivo measurements. This type of 

interference can produce erroneously high bioassay measurement results that can sometimes be 

identified as outliers. Notes regarding suspected contamination are sometimes found in the bioassay 

records. When no such notes are found in the bioassay records, a dose reconstructor may still be 

able to determine if a bioassay measurement is an outlier by using the subsequent bioassay 

measurement results and information regarding the affected worker’s work locations and the activities 

being performed at those locations before the elevated measurement. However, a significant amount 

of caution should be exercised when determining if an INEL bioassay measurement result is an 

outlier, because the most common significant intakes at the INEL Site were attributable to 

radionuclides that are quickly eliminated from the body, which often do not show up in subsequent 

bioassay measurements. Therefore, bioassay results above their detection or reporting limits should 

be considered to be real, unless there is conclusive information to the contrary.

5.4.1.2 Naturally Occurring Uranium in Uranium Bioassay Measurements

Uranium is present in secular equilibrium throughout the Earth. Consequently, there is a continuous 

source of intake of naturally occurring uranium via ingestion of drinking water and food and inhalation 

of particulate matter, which can interfere with bioassay measurements being used to detect intakes of 

uranium from occupational exposures. Because the uranium bioassay measurements performed for 

the ICPP workers who were exposed to HEU were isotopic analyses and because the mixture of 

uranium isotopes in recycled HEU is significantly different from natural uranium, positive sample 

results attributable to naturally occurring uranium can be readily distinguished from results attributable 

to HEU. However, the Kinetic Phosphorescence Analysis (KPA) that is performed on SMC Project 

urine samples cannot distinguish between different types of uranium, be it naturally occurring, 

depleted, or enriched. Therefore, the uranium in urine sample results for SMC Project workers have 

been adjusted to discount the interference from naturally occurring uranium. Consequently, the 
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reported bioassay results in the worker files reflect the subtraction of 0.16 μg U/L from the value 

determined in the laboratory bioassay result (King 2001).

The 0.16 μg U/L adjustment value is based on measured background levels of uranium in urine 

samples submitted in 1987, 1994, and 1998 by SMC Project workers who were not radiological 

workers. Those measurements were assumed to represent the non-occupational excretion levels of 

the SMC Project worker population, because performing background measurements using members 

of the public was not feasible. The background measurement results ranged from 0.04 to 0.33 μg U/L 

with wide fluctuations in individual measurements, some as high as 1.0 μg U/L (King 2001). The 

average reported uranium concentration was 0.157 ±0.109 μg U/L at 1σ uncertainty. Therefore, 

0.16 μg U/L is used as the non-occupational component of uranium excretion for SMC Project 

workers, and is subtracted from each urine result before assessment of occupational internal dose. 

ICRP Publication 23 (ICRP 1975) lists the daily intake of naturally occurring uranium as 1.9 μg U/d 

and a daily urine excretion rate of 1.4 L for Reference Man. Using a uranium intake of 1.9 μg U/d and

a daily urine excretion rate of 1.4 L, the typical urinary concentration ranges between 0.4 and 0.5 μg

U/L for naturally occurring uranium (King 2001). Therefore, the adjustment value used at the INEL is 

consistent with the ICRP reference values for naturally occurring uranium. 

5.4.1.3 Cesium-137 from Fallout

As indicated above, fallout from the atmospheric testing of nuclear weapons affected everyone in

North America and body burdens of Cs-137 measurable in whole-body counters throughout the 

United States were common in the 1960s and 1970s. Even though the United States stopped its 

atmospheric testing of nuclear weapons by 1963, Cs-137 intakes from fallout were still often detected 

after that because of the Cs-137 that was deposited on soil and vegetation. Significantly higher 

Cs-137 intakes from fallout are sometimes identified in persons who consume game meats. This is 

due to specific game animals whose diets include plants that tend to concentrate the Cs-137 in the 

environment. 

Because the doses attributable to fallout Cs-137 are not considered to be occupational doses, Cs-137 

doses attributable to fallout did not need to be assigned by a site. However, no information has been 

found to indicate that the INEL adjusted its WBC results to eliminate the contribution from nonoccupational Cs-137 intakes. Therefore, the dose reconstructor can assume that the reported WBC

results are the total measured Cs-137 results. 

As indicated above, the doses attributable to fallout Cs-137 are not considered to be occupational 

doses. Therefore, under the circumstance discussed in Section 5.3.1, the dose reconstructor can 

attribute positive WBC results to fallout Cs-137 and not assign any intakes or doses based on those 

WBCs.

5.4.1.4 Intakes of Radionuclides for Medical Reasons

Bioassay measurements at the INEL periodically detected intakes of radionuclides that were received 

for medical reasons. In such cases, there are notes in the worker’s bioassay records identifying the 

intake as a medical intake. Because medical intakes are not occupational intakes, they are not 

covered under the EEOICPA and should not be included in the dose reconstructions.

5.4.2 Uncertainties

The uncertainty values for all types of bioassay measurements are typically included in the INEL’s 

bioassay records that are provided by the DOE. When measurement-specific uncertainty values are 

available, those values are preferred for the data analysis over generic values. When the uncertainty 

values are not included with the bioassay records, the uncertainty values to be used for the data 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 38 of 63

analysis should be determined in accordance with the recommendations in the Technical Information 

Bulletin: Internal Dose Reconstruction (ORAUT 2007b). 

5.5 INTAKE AND INTERNAL DOSE ASSESSMENT FOR MONITORED WORKERS

This section is intended to apply to the periods of employment that an INEL worker was monitored for 

internal dose. For INEL workers who were not monitored for some periods of their employment, the 

recommendations in Section 5.6 should be followed to assess the potential internal doses associated 

with the unmonitored periods of their employment.

When bioassay data are used to assess the intakes and internal doses for monitored INEL workers, 

the recommendations in the Technical Information Bulletin: Internal Dose Reconstruction (ORAUT 

2007b) should be followed. This technical information bulletin is also known as OTIB-0060, and shall 

be referred to as OTIB-0060 throughout the remainder of Section 5.5. In addition, when workers have 

monitoring data for specific radionuclides, the radionuclide-specific data are the preferred data for 

estimating intakes and internal doses for those radionuclides. Because the majority of INEL workers 

were only monitored for beta- and/or gamma-emitting radionuclides and because the reported 

bioassay results typically do not provide any indication of the specific radionuclides involved with the 

intake or potential intake, intakes of specific radionuclides will typically need to be assigned based on 

ratios that are applied to the estimated Sr-90 and/or Cs-137 intakes. The details on when this needs 

to be done and on how to do it are provided in the following subsections.

5.5.1 Activation and Fission Product Intakes

Because the majority of the bioassay measurements performed for INEL workers did not determine 

the potential mixtures of the activation and fission products that the workers were exposed to and 

because those potential mixtures have a significant impact on the workers’ doses, a method for 

estimating the mixtures of the activation and fission products needed to be selected for the INEL Site. 

As a result, the method described in the Technical Information Bulletin: Fission and Activation Product 

Assignment for Internal Dose-Related Gross Beta and Gross Gamma Analyses (ORAUT 2007c) was 

determined to be appropriate and applicable for the INEL Site, because some INEL reactor data were

used to develop that method. This technical information bulletin is also known as OTIB-0054, and 

shall be referred to as OTIB-0054 throughout the remainder of Section 5.5.

NOTE: Because the radionuclides associated with work performed for the SMC Project were limited 

to isotopes of uranium and the impurities found in recycled uranium, the approach described in this 

section is not applicable to work performed for the SMC Project. To estimate intakes and internal 

doses for SMC Project workers, refer to Section 5.5.5 below.

NOTE: Because the radioactive materials associated with some of the work areas at the RWMC did 

not have radionuclide compositions that were completely represented by OTIB-0054 (ORAUT 2007c), 

the intakes received at the RWMC need to be evaluated on a case-by-case basis. To estimate 

intakes and internal doses for RWMC workers, refer to Section 5.5.6 below.

Section 3.0 of OTIB-0054 (ORAUT 2007c) identifies two exclusions that were potentially applicable to 

the ICPP. The following paragraphs describe why those exclusions do not apply to the ICPP. 

Even though radioactive lanthanum (RaLa) work was performed at the ICPP between 1956 and 1963, 

this exclusion in OTIB-0054 (ORAUT 2007c) does not apply to the ICPP because the irradiated

nuclear fuel that was reprocessed at the ICPP as part of RaLa operations accounted for less than 1% 

of the total amounts of irradiated reactor fuel reprocessed at the ICPP during each of those years. 

Therefore, the radioactive materials associated with the RaLa operations only account for an 

insignificant fraction of the total amount of irradiated reactor fuels that were reprocessed at the ICPP 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 39 of 63

during each of the years between 1956 and 1963, and those materials are not considered to be 

representative of the radioactive materials that the majority of workers were exposed to. However, 

there were a limited number of ICPP workers who did receive intakes of radioactive materials due to 

the RaLa operations that were performed. Based on the bioassay results for those exposure 

incidents, the gamma spectrometry results indicated that the intakes from the RaLa operations were 

usually limited to radioactive isotopes of iodine versus a mixture of all the activation and fission 

products present in the source term for the RaLa operations. When the dosimetry records for an 

ICPP worker indicates that they received an intake from the RaLa operations or when there are 

positive bioassay results during the years of 1956-1963 that include radioactive isotopes of iodine, 

those potential intakes should be assessed in accordance with OTIB-0060 (ORAUT 2007b) and 

Section 5.5.4 below.

The available information on the ICPP indicated that some activation and/or fission products might

have been extracted and concentrated at the ICPP. However, there is no information to indicate this 

was a routine process or that the activation and/or fission products were separated in quantities that 

were capable of causing a significant alteration to the radioactive source term that the workers were 

exposed to. Therefore, this exclusion does not apply to the ICPP. 

5.5.1.1 Assessment of Missed Intakes

The majority of the bioassay measurements at the INEL were performed to detect the predominant

beta/gamma-emitting activation and fission products that were found in the irradiated nuclear reactor 

fuels. As a result, urine samples were typically only analyzed for gross beta, gross gamma, and/or 

strontium radioactivity. When the urine samples are only analyzed for gross beta, gross gamma, 

and/or strontium radioactivity, missed Sr-90 and/or Cs-137 intakes should be assessed in accordance

with OTIB-0054 (ORAUT 2007c) and OTIB-0060 (ORAUT 2007b). Similarly, missed Cs-137 intakes 

should be assessed using the WBC data, in accordance with OTIB-0060 (ORAUT 2007b).

NOTE: For some of the strontium analyses only a preliminary count of the gross beta radioactivity 

was performed on the portion of the sample that was precipitated by oxalic acid, because the samples 

were below the detection limits. These sample results were typically labeled as “β(ox)” results in the 

INEL’s bioassay records, and should be treated as chemically processed beta samples when 

following the OTIB-0054 recommendations. When the preliminary β(ox) result was above the 

detection limits the processing for the strontium analysis was completed and only the completed 

strontium analysis results were reported.

5.5.1.2 Assessment of Fitted Intakes

Fitted intakes are assessed when there are positive bioassay measurement results. A bioassay 

measurement is considered to be positive when its result is greater than the reporting level (this could

be the MDA, detection level, or some other value that the site used), and is considered to be negative 

when its result is less than or equal to the reporting level (ORAUT 2007b). 

When a positive bioassay result can be attributed to specific radionuclides, the fitted intake should be 

assessed for those specific radionuclides, in accordance with OTIB-0060 (ORAUT 2007b). 

Unfortunately, the radionuclides responsible for many positive urine sample results cannot be 

determined when only a gross radioactivity analysis was performed on them. In those instances, a 

Sr-90 or Cs-137 intake should be calculated in accordance with the recommendations provided in 

OTIB-0054 (ORAUT 2007c) and OTIB-0060 (ORAUT 2007b). Even though no information was 

provided in the dosimetry records to indicate that the positive urine sample results were attributable to 

either Na-24 or radioactive iodine, it may be possible to determine if a radioactive iodine or Na-24 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 40 of 63

intake occurred. Sections 5.5.4 and 5.5.6 provides guidance on how to determine if an intake was 

likely attributable to Na-24 or an isotope of radioactive iodine.

5.5.1.3 Assessment of Unmonitored Intakes

When missed and/or fitted intakes are calculated for Sr-90 and/or Cs-137, the unmonitored intakes for 

all other dosimetrically significant activation and fission products should then be calculated using the 

ratios provided in Table 7-3 of OTIB-0054 (ORAUT 2007c), unless a bioassay result was reported for 

any of the radionuclides listed in Table 7-3. The use of actual bioassay data (e.g. radionuclidespecific urine sample, fecal sample, or in vivo measurement results) is preferred any time radionuclide 

specific results are provided.

5.5.1.4 Decay Period Selections

The decay periods to be used for the selection of the appropriate urine activity fractions and 

associated radionuclide ratios for the intake estimates are to be based on a combination of the 

recommendations in Table 5-12 of OTIB-0054 (ORAUT 2007c) and the available site-specific 

information. 

For operating reactors, a decay period of 10 days is appropriate, until the reactor is shut down 

permanently. Once a reactor has been shut down permanently for a year, a decay period of 1 year 

can be used for the intake estimates. Because of the uncertainties with some of the reactor operation 

end dates (e.g. sometimes only the year of the shutdown is known), the use of the 40 and 180 day 

decay periods is not recommended for the shutdown period for the INEL reactors. 

Because the minimum decay periods for the spent and/or irradiated reactor fuel storage facilities and 

fuel examinations facilities at ANL-W and TAN could not be determined at the time this TBD was 

prepared, the recommendations for the INEL reactors also apply to these irradiated or irradiated 

reactor fuel storage and/or examination facilities, unless documentation supporting different decay 

periods is located for those facilities. 

For the ICPP, intake estimates based on the 40 day, 180 day, and 1 year decay periods should each 

be assessed, and the decay period resulting in the highest internal dose for a given organ should be 

used for the dose assignment. Because the ICPP reprocessed irradiated fuels with varying decay 

periods, this approach is considered to be reasonable for all types of claims regardless of the 

compensability decision. At the ICPP, some irradiated fuels were reprocessed with less than a 40 day 

decay period because of the RaLa operations. However, a decay period shorter than 40 days is not 

justified because of the relatively small amount of reactor fuel involved with the RaLa operations. The 

RaLa operations performed at the ICPP between 1957 and 1963 involved processing irradiated fuel 

with only a 2 day decay period; however, as indicated above, this fuel represented less than 1% of the 

fuel that was reprocessed during each of those years. Irradiated fuels were also stored at the ICPP’s 

Fuel Storage Facility (Building CPP-603). Table 5-12 of OTIB-0054 (ORAUT 2007c) recommends a 

10 day decay period for spent fuel storage facilities when a more appropriate decay period is 

unknown. However, site-specific information indicates that, with the exception of the RaLa 

operations, the irradiated reactor fuels had a minimum of 90 days of cooling before being received at 

the ICPP’s Fuel Storage Facility (Allied Chemical undated). Therefore, the decay period 

recommendations for the ICPP also apply to the ICPP’s Fuel Storage Facility.

For the INEL’s waste management facilities, a decay period of 1 year is appropriate for the intake 

estimates.
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5.5.2 Actinide Intakes

Because uranium was the only actinide that was routinely separated from the irradiated nuclear fuels 

at the INEL Site and because the much more readily detectable fission products were present along 

with the actinides, the potential intakes and doses attributable to the actinides can be estimated by 

applying ratios to the estimated Sr-90 and/or Cs-137 intakes for a worker. In the event that a worker 

was monitored for uranium intakes or intakes of other actinides, the actual bioassay results should be 

used to estimate the worker’s internal doses for those radionuclides. 

NOTE: The recommendations in this section apply to all INEL workers with the exception of the 

workers on the SMC Project. For SMC Project workers, the recommendations in Sections 5.5.4 

should be followed to account for those workers’ potential actinide intakes and doses.

Actinide to Sr-90 and actinide to Cs-137 ratios are provided in Tables 5-22 and 5-23 for the three 

major fuel types that were present at the INEL Site. For the ANL-W reactor facilities, the actinide 

ratios for stainless-steel fuels should be used. With the exception of the ANL-W reactor facilities, the 

actinide ratios for aluminum fuels should be used for the remaining reactor sites. Even though some 

reactors did not use aluminum fuel types, the aluminum fuel actinide ratios for those reactors likely 

provide a reasonable overestimate of the actinides in the other fuel types, because the fuels from nonaluminum-fueled reactors typically had much lower burnups than what was assumed for the aluminum 

fuel calculations. 

For the ICPP, the actinide ratios for the aluminum fuels likely provide a reasonable overestimate of 

the actinides present at the ICPP before 1971, because the non-aluminum fuels that were 

reprocessed at the ICPP before 1971 had much lower burnups than what was assumed for the 

aluminum fuels. Because the ICPP was reprocessing a combination of different reactor fuels that had 

higher burnups by 1971 and because the ICPP workers might have been simultaneously exposed to 

an unknown mixture of those fuel types, the maximum ratio between the three major fuel types (i.e., 

aluminum, zirconium, and stainless-steel fuels) for a given actinide should be used for ICPP workers 

after 1970. 

For the INEL’s waste management facilities, the maximum actinide ratios should be used for the 

intake estimates.

The radioactive source term that was used to generate the values in Tables 5-22 and 5-23 was based 

on the information provided in an INEL document titled Determination of the Normalized Mass of 

Individual Radionuclides in the Dissolver Product for Aluminum, Zirconium and Stainless Steel Fuels 

Previously Processed at INTEC (Wenzel 2000). This document indicated that the source term values 

were generated using the ORIGEN2 computer model and that calculations were only performed for 

the three main types of irradiated reactor fuel that were reprocessed at the ICPP (also known as 

INTEC). In addition, only a single source term calculation was performed for each of the reactor fuel 

types using the typical composition, configuration, nuclear fuel burnup, and post-irradiation decay 

periods that were applicable to those reactor fuels. Because no uncertainty information was provided 

for the ORIGEN2 calculations, the source term information and the actinide ratios that were calculated 

for this TBD will be treated as constants.

Because the number of favorable to claimant simplifying approaches or assumptions used to develop 

the ratios for the actinides was minimal, the distribution(s) associated with the indicator radionuclide 

intakes will need to be accounted for. Therefore, the distribution associated with the unmonitored 

actinide intakes and internal doses should be indicative of the distributions associated with the 

indicator radionuclide intakes that they were based on. For a fitted dose calculation, the distribution 

should be lognormal with a geometric standard deviation (GSD) of 3.0 (ORAUT 2007b). For a missed 

dose calculation, the distribution should be triangular (ORAUT 2007b).
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 42 of 63

Table 5-22. Actinide-to-Sr-90 ratios.a

Actinide

Reactor fuel typesb

Al Zr SS Max

Ac 8.0E-12 1.3E-11 2.3E-10 2.3E-10

Ac-227 Ac-227 Ac-227 Ac-227

Th 2.4E-08 6.4E-08 2.3E-07 2.3E-07

Th-228 Th-228 Th-228 Th-228

Pa 1.2E-10 1.1E-10 3.8E-09 3.8E-09

Pa-231 Pa-231 Pa-231 Pa-231

U 5.6E-05 6.2E-06 1.4E-03 1.4E-03

U-234 U-236 U-234 U-234

Np 3.4E-06 3.7E-06 6.8E-07 3.7E-06

Np-237 Np-237 Np-237 Np-237

Pu 8.7E-03 1.5E-02 3.7E-03 1.5E-02

Pu-238 Pu-238 Pu-239 Pu-238

Am 1.4E-04 3.9E-06 9.0E-08 1.4E-04

Am-241 Am-241 Am-241 Am-241

Cm 4.9E-05 1.8E-06 1.1E-10 4.9E-05

Cm-244 Cm-244 Cm-242 Cm-244

a. The values in this table were obtained from the MS Excel workbook titled INEL 

– Actinide Ratios (ORAUT 2009a).

b. Al = aluminum; Zr = zirconium; SS = stainless-steel; Max = maximum; The 

actinide isotopes to use for the dose calculations are provided below each 

actinide ratio; The actinide isotopes listed are the predominant alpha-emitting 

actinides in the source term for a given fuel type.

Table 5-23. Actinide-to-Cs-137 ratios.

a

Actinide

Reactor fuel typesb

Al Zr SS Max

Ac 7.6E-12 1.3E-11 2.1E-10 2.1E-10

Ac-227 Ac-227 Ac-227 Ac-227

Th 2.3E-08 6.2E-08 2.1E-07 2.1E-07

Th-228 Th-228 Th-228 Th-228

Pa 1.2E-10 1.1E-10 3.5E-09 3.5E-09

Pa-231 Pa-231 Pa-231 Pa-231

U 5.3E-05 6.0E-06 1.3E-03 1.3E-03

U-234 U-236 U-234 U-234

Np 3.2E-06 3.5E-06 6.2E-07 3.5E-06

Np-237 Np-237 Np-237 Np-237

Pu 8.3E-03 1.4E-02 3.4E-03 1.4E-02

Pu-238 Pu-238 Pu-239 Pu-238

Am 1.3E-04 3.7E-06 8.3E-08 1.3E-04

Am-241 Am-241 Am-241 Am-241

Cm 4.7E-05 1.7E-06 1.0E-10 4.7E-05

Cm-244 Cm-244 Cm-242 Cm-244

a. The values in this table were obtained from the MS Excel workbook titled INEL 

– Actinide Ratios (ORAUT 2009a).

b. Al = aluminum; Zr = zirconium; SS = stainless-steel; Max = maximum; The 

actinide isotopes to use for the dose calculations are provided below each 

actinide ratio; The actinide isotopes listed are the predominant alpha-emitting 

actinides in the source term for a given fuel type.

5.5.3 Contaminant Intakes for ICPP Workers Who Handled Recycled HEU

If an ICPP worker has bioassay data for uranium, the worker likely handled recycled HEU that was 

separated from the irradiated reactor fuels. Because the actinide ratios for the recycled HEU were 

significantly different than the ratios for the irradiated reactor fuel, a different set of ratios needs to be 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 43 of 63

used to assess unmonitored radionuclide intakes for periods that a worker handled recycled HEU. 

Because the majority of the ICPP’s HEU product was shipped to the Y-12 Plant, the activity ratios for 

the ICPP HEU are based on information provided for recycled uranium in the Technical Basis 

Document for the Y-12 National Security Complex – Occupational Internal Dose (ORAUT 2006). The 

default ratios to estimate the unmonitored radionuclide intakes are provided in Table 5-24. The

unmonitored radionuclide intakes are calculated by multiplying the default ratios by the total uranium 

intake that was calculated based on an assessment of the bioassay data.

Table 5-24. Default ratios for recycled HEU intakes.a

Radionuclide

Default ratio

(pCi/μg total U)

Tc-99 1.33

Th-228 0.05

Np-237 0.06

Pu-238 0.02

a. The values in this table are based on the values provided in 

Table 5-8 of the Y-12 Plant TBD (ORAUT 2006). The 

values have also been adjusted in accordance with that TBD 

for use on all types of cases, including best estimate cases.

It should be noted that the worker might have also been exposed to the radionuclides in the irradiated 

reactor fuels prior to the uranium being separated. This would be indicated by bioassay data for those 

workers, which would include either urine sample data analyzed for gross beta, gross gamma, and/or 

Sr-90 radioactivity or a WBC data for the same period that the uranium bioassay data were provided. 

In those cases, the potential unmonitored radionuclide exposures should also be assessed in 

accordance with the recommendations in Sections 5.5.1 and 5.5.2 (i.e., assessed in addition to the 

unmonitored intakes assessed for recycled HEU handlers).

5.5.4 Radioactive Iodine Intakes

5.5.4.1 General Guidance on Identified Radioactive Iodine Intakes

At the INEL, the results of the bioassay measurements may be dominated by radioactive iodine. 

When this is the case, it is unlikely that there was a detectable intake of the other activation and 

fission products. In those instances, the positive bioassay results only need to be assessed for the 

radioactive isotopes of iodine that might have been present. 

The INEL Site often performed a gamma spectrometry analysis to identify and/or confirm the 

radionuclides that were associated with a bioassay result that INEL dosimetry personnel considered 

to be significant or suspected to be associated with a radioactive iodine intake. In some instances, 

the determination that an intake was an iodine intake might have been based on process knowledge 

or other workplace indicators. In those instances, the isotopes involved with the intake are written 

next to the positive bioassay result, which is usually a gross gamma in urine result and sometimes 

might be a gross beta in urine result. However, isotopic mixtures or the quantities of the specific 

radionuclides involved with the intake are not provided in most circumstances. One common 

exception to this is when the intake was significant enough to warrant an incident investigation and 

the generation of an incident report. Because WBCs are a gamma spectrometry type of analysis, the 

specific radionuclides that are above the detection limits or reporting levels are always reported. 

When those results are above the reporting levels, which were typically 0.1 μCi, the measured 

quantities of each radionuclide are provided. 

Isotope-specific information provided with positive bioassay results should be used to assess the 

intakes that are attributable to radioactive iodine. If multiple iodine isotopes are identified and no 

information regarding the specific mixture is provided, the bioassay results should be assessed 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 44 of 63

assuming that the intake was attributable solely to each iodine isotope and the isotope generating the 

highest dose estimate shall be used for the assigned dose. When there is no isotopic information 

provided for a radioactive iodine intake, the dose reconstructor should assume that the iodine isotope 

is solely I-131, unless there is information to indicate that the worker’s intake occurred during RaLa 

operations at the ICPP. 

If a worker’s radioactive iodine intake occurred during RaLa operations at the ICPP, the isotopic 

mixture provided in Table 5-25 should be used to assess the radioactive iodine intakes and doses. 

The isotopic fractions for the iodine isotopes provided in Table 5-25 are based on an ORIGEN 

computer program run for the MTR fuel with a 2 day decay period, which was the fuel type and typical 

decay period used for the RaLa operations. The isotopes listed in Table 5-25 account for more than 

99% of the iodine radioactivity.

Table 5-25. RaLa iodine.

Iodine isotope Isotopic fractions

I-131 0.3245

I-132 0.4553

I-133 0.2203

5.5.4.2 Special Considerations for Thyroid Cancer Cases

Because thyroid cancer cases are significantly impacted by intakes of radioactive iodine, some 

special considerations regarding the assessment of the missed and/or unmonitored iodine doses 

need to be made for those types of cases. For intakes based on positive bioassay results, the 

recommendations provided in the previous sections are still applicable.

Because a WBC is capable of detecting most short-lived isotopes of iodine, a missed intake and dose 

for I-131 can be calculated from negative WBCs in lieu of the unmonitored I-131 intakes that would be 

calculated in accordance to the recommendations in Section 5.5.1.3 above. 

When using the OTIB-0054 (ORAUT 2007c) approach for assigning unmonitored iodine intakes after 

a 40 day decay period, intakes of the much longer-lived I-129 need to be accounted for in the 

unmonitored internal doses to the thyroid, because the short-lived isotopes have completely decayed 

away and are no longer present. Until the much shorter lived isotopes of iodine decayed away, the 

internal doses to the thyroid that were attributable to I-129 were considered to be insignificant relevant 

to the shorter lived isotopes. It should also be noted that I-129’s contribution to the internal doses to 

organs other than the thyroid are still considered to be insignificant relative to the other activation and 

fission products that might have been present; therefore, internal doses attributable to I-129 only need 

to be considered for thyroid cancers. 

Because OTIB-0054 (ORAUT 2007c) does not provide I-129-to-indicator nuclide ratios and because 

the aged iodine is likely in a less volatile form, an approach similar to the approach being used for 

assessing unmonitored actinide intakes in Section 5.5.2 will be used to assess unmonitored I-129 

intakes. Table 5-26 provides I-129 ratios for indicator radionuclides that can be used to estimate the

unmonitored I-129 intakes when the applicable decay period exceeds 40 days. Because elemental 

iodine is very reactive and because it has had over 40 days to react with other materials, I-129 should 

only be assessed as an inhalation intake and a material having type F lung absorption properties.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 45 of 63

Table 5-26. I-129-to-indicator nuclide ratios.a

Indicator nuclide Ratio

Sr-90 4.2E-07

Cs-137 3.9E-07

a. The values in this table were obtained from the 

MS Excel workbook titled INEL – I-129 Ratios

for Internal TBD (ORAUT 2009b).

The radioactive source term that was used to generate the values in Table 5-26 was based on the 

information provided in an INEL document titled Determination of the Normalized Mass of Individual 

Radionuclides in the Dissolver Product for Aluminum, Zirconium and Stainless Steel Fuels Previously 

Processed at INTEC (Wenzel 2000). This document indicated that the source term values were 

generated using the ORIGEN2 computer model and that calculations were only performed for the 

three main types of irradiated reactor fuel that were reprocessed at the ICPP (also known as INTEC). 

In addition, only a single source term calculation was performed for each of the reactor fuel types 

using the typical composition, configuration, nuclear fuel burnup, and post-irradiation decay periods 

that were applicable to those reactor fuels. Because no uncertainty information was provided for the 

ORIGEN2 calculations, the source term information and the actinide ratios that were calculated for 

this TBD will be treated as constants.

Because the number of favorable to claimant simplifying approaches or assumptions used to develop 

the ratios for I-129 was minimal, the distribution(s) associated with the indicator radionuclide intakes 

will need to be accounted for. Therefore, the distribution associated with the unmonitored I-129 

intakes and internal doses should be indicative of the distributions associated with the indicator 

radionuclide intakes that they were based on. For a fitted dose calculation, the distribution should be 

lognormal with a geometric standard deviation (GSD) of 3.0 (ORAUT 2007b). For a missed dose 

calculation, the distribution should be triangular (ORAUT 2007b).

5.5.5 Uranium Intakes for SMC Project Workers

The SMC Project was located at TAN at the old ANP Program facilities (ORAUT 2007a). In 1985, the 

SMC Project started manufacturing DU armor for the U.S. Army (ORAUT 2007a). As of June 2000, 

records show that the SMC Project has received 10,129,000 lb of DU for processing (Lewis et al. 

2000). Of this, 4,726,000 lb were received from the DOE Feed Materials Production Center (FMPC) 

at Fernald, Ohio, and 5,403,000 lb were received from the RFP (Lewis et al. 2000). The records also 

indicate that the DU being used by the SMC Project was recycled DU, and contained other radioactive 

impurities (Lewis et al. 2000). 

The processing of DU at the SMC Project consists of rolling and cutting billets. Recasting is also 

performed and the SMC Project requires that only metallic DU be used for the recasting process. No 

chemical processing beyond recasting takes place at the SMC Project. The transuranic impurities in 

the DU are neither concentrated nor diluted by the recasting process (Lewis et al. 2000). During DU 

parts fabrication, small quantities of finely divided uranium metal and oxides present inhalation and 

ingestion potential, as indicated by routine positive personnel bioassays (King 2001).

Air monitoring is the primary method used at the SMC Project to evaluate the potential for exposure to 

airborne DU. Fixed-head air sampling throughout the plant, supplemented by CAMs, provides the 

routine information to evaluate the effectiveness of control programs and to indicate potential internal 

intake. Exposures to concentrations above 0.1 DAC generally indicate the use of respiratory 

protection and require bioassay follow-up (King 2001).

The uranium values provided in Table 5-27 are representative of the typical composition of recycled 

DU. In addition to the uranium isotopes, DU contains two beta-emitting radionuclides. Average 

values for the radiologically significant impurities in the SMC Project’s DU are provided in Table 5-28. 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 46 of 63

Table 5-27. Uranium mass and activity ratios for SMC project DU.a

Uranium isotope Mass % Activity % Total (pCi/μg DU)

U-234 0.0010 15.46 0.402

U-235 0.1991 1.07

U-236 0.0003 0.05

U-238 99.7996 83.42

a. The values in this table are the default values for DU that are provided in the 

IMBA computer program, because the site-specific information (King 2001) did 

not account for the U-236 that is in recycled DU.

Table 5-28. SMC Project DU impurities. 

Nuclide

Average impurity level

(pCi/g DU)a (pCi/pCi DU)

Np-237 1.82 4.90E-06

Pu-238 0.272 7.33E-07

Pu-239/240 0.406 1.09E-06

Am-241 2.78 7.49E-06

Tc-99 154 4.15E-04

a. These values were obtained from INEEL/EXT-2000-00959 (Lewis et al. 2000).

Inhalation Absorption Type: Respirable particulates associated with SMC Project operations are 

probably a mixture of metal and metal oxides. The actual exposures are undoubtedly due to mixtures 

of absorption types. During the 18 years of operation, much bioassay data have been collected on a 

large number of individuals. The overall elimination patterns are consistent with type M but probably 

are a mixture of all types. It could be too simplistic to assume a pure absorption type when the 

chemical form is not known for certain. The dose reconstructor should assume either type M or type 

S uranium to maximize the dose to the organ of concern. Exposure to significant quantities of type F 

uranium at the SMC Project is not considered credible [27].

Particle Size: Detailed particle size analyses of representative samples from the various operations 

indicate that an activity median aerodynamic diameter (AMAD) of 2.4 μm is appropriate for typical 

SMC Project operations. This site-specific value of 2.4 μm AMAD is used for assessments of intakes 

at the SMC Project and is the default particle size distribution (INEEL 2001).

5.5.6 Intakes for RWMC Workers

The RWMC has supported INEL operations as a waste management complex since 1952 and has 

received large quantities of TRU waste from the RFP and other DOE facilities. Improved operations 

have resulted in a decrease in internal dose potential. The original disposal techniques (dumping 

waste in open trenches) were relatively vulnerable to airborne release in comparison to current totalcontainment practices. The four major areas in the RWMC facility are: (1) the Subsurface Disposal 

Area (SDA) for permanent disposal of low-level waste and some early TRU waste (which will 

eventually be exhumed and repackaged); (2) the 58-acre Transuranic Storage Area (TSA) for 

temporary storage, examination, and certification before shipment to the Waste Isolation Pilot Plant; 

(3) the operations area; and (4) an administrative area where no radioactive waste is permitted.

The comprehensive radiation protection program for RWMC includes extensive air monitoring, 

personnel monitoring, and surface contamination surveillance. Although infrequent, there have been 

instances of inadvertent intakes (based on Table 5-1, there were two in 1992 and one in 1996). 

Therefore, bioassay is conducted randomly at the present time (INEEL 2001).

Because of the variety of radiological materials that were present at the RWMC, more than one 

mixture of radioactive materials may need to be considered for the intake and internal dose 

calculations for RWMC workers. For RWMC workers who were likely exposed to materials bearing 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 47 of 63

irradiated fuels or for instances when the material that the worker was exposed to was unknown, the 

guidance information in Sections 5.5.1 and 5.5.2 should be followed. 

Tables 5-29 and 5-30 summarize the major radionuclides in the RWMC waste inventory for the SDA

and TSA, respectively. TRU radionuclides are the primary contaminants in the TSA waste; all but 

Pu-241 are alpha emitters. Because the materials are not homogeneous, it should not be assumed 

that failing to detect one radionuclide in the inventory invalidates detection of other radionuclides. 

Table 5-29. Radioactive low-level waste inventory in the active pits in the SDA (INEEL 2001). 

Waste type Volume (m3) Total Ci Radionuclide Concentration (Ci/m3) Percentage

Low-level waste 75,600 3.35E+05 Co-60 4.1E+0 92.00%

Ni-63 3.3E-1 7.40%

Sr-90 9.7E-3 0.22%

Cs-137 9.7E-3 0.22%

H-3 5.8E-3 0.13%

C-14 8.9E-4 0.02%

Table 5-30. Radioactive waste inventory in the TSA (INEEL 2001). 

Waste type Volume (m3) Total Ci Radionuclide Concentration (Ci/m3) Percentage

Stored contacthandled TRU waste65,000 4.06E+5 Pu-241 2.5E+00 44.1%

Am-241 1.4E+00 24.7%

Pu-238 9.7E-01 17.1%

Pu-239 6.3E-01 11.1%

Pu-240 1.5E-01 2.6%

U-233 1.4E-02 0.2%

Cm-244 0.8E-02 0.1%

5.5.7 Intakes for Waste Reduction Operations Complex Workers

The Waste Reduction Operations Complex (WROC) includes several reactor facilities that operated 

from the 1950s to the late 1960s and the Power Burst Facility (PBF) reactor, which operated from 

1972 to 1985. These currently inactive facilities are in a common control area. In addition, a low-level 

waste incinerator called the Waste Experimental Reduction Facility (WERF) burned waste from all 

INEL facilities from 1982 to 2001. The WERF, which is undergoing decontamination and 

decommissioning, was a low-level waste incinerator, and its operations included some mixed waste 

treatment (Stacy 2000). 

The waste at WERF was in the form of burnable containers and the resultant high-fired and solidified 

ash. The radioactive wastes at the mixed waste storage facility and the reactors were the sources of 

the radioactivity inventory. The ashes were removed remotely to a glovebox and solidified in 55-gal 

drums. 

The radiological protection program included CAMs, fixed air-sampling systems, RAMs, surface and 

personnel contamination surveillance, and effluent monitors.

To assess potential exposures to activation and fission products for WROC workers, the 

recommendations in Section 5.5.1 should be followed. Because the types of radioactive materials 

processed at WROC varied depending on the area shipping the waste to WROC, the assumption 

favorable to claimants is that potential actinide exposures were associated with zirconium fuel as 

processed at the ICPP. Therefore, the recommendations in Section 5.5.2 for zirconium fuels should 

be used to assess potential actinide exposures at the WROC.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 48 of 63

5.5.8 Sodium-24 Intakes at Liquid-Sodium-Cooled Reactors

Because liquid sodium metal was used as a coolant for the INEL’s fast reactors, significant quantities 

of Na-24 were produced by neutron reactions in the coolant. However, unmonitored Na-24 intakes at 

the fast reactor sites are unlikely for the following reasons.

Sodium metal would not have been in areas routinely accessed by personnel until most of the Na-24 

had decayed due to the intense gamma radiation from the Na-24. Given that the half-life of Na-24 is 

only 15 hours, a decay period of only about 6.25 days would be required to eliminate nearly all of the 

Na-24 in the sodium metal. Significant intakes of Na-24 from sodium metal in the solid or liquid forms 

are also unlikely, due to its reduced dispersion potential in those forms. It is also unlikely that 

unknown accidental exposures to Na-24 would have occurred, because sodium metal is highly 

reactive and reacts violently with water. Because of the significant hazards associated with sodium 

metal, any exposures to the metal would have been minimized and would have also been known 

when they occurred. When an exposure was known to have occurred, the radiological protection 

practices being performed throughout the INEL Site would have likely required that bioassay 

measurements capable of detecting Na-24 intakes be performed on the exposed workers. Therefore, 

no special assessments need be performed for workers who were not monitored for Na-24 exposures 

at the fast reactors. 

If it can be determined that a worker received an intake of Na-24 or was monitored for a Na-24 

exposure, fitted and/or missed Na-24 intakes and doses may need to be assessed. When bioassay 

measurements were used to monitor a worker for Na-24 exposures, the potential Na-24 intakes and 

doses are assessed in accordance with the OTIB-0060 (ORAUT 2007b).

5.6 INTAKE AND INTERNAL DOSE ASSESSMENT FOR UNMONITORED WORKERS

This section is intended to apply to the periods of employment that an INEL worker was not monitored 

for internal dose. 

For most of the history of the INEL, personnel dosimeters were issued to all workers who entered the 

security access control points at each facility, regardless of their work assignment. For example, 

administrative and clerical personnel were required to wear dosimeters even though they were not 

exposed to external or internal doses that were above the onsite ambient levels. Dose reconstructors 

should determine the appropriate unmonitored dose assignment in accordance with the following 

recommendations or the recommendations for unmonitored workers in the Technical Information 

Bulletin: Application of Internal Doses Based on Claimant-Favorable Assumptions for Processing as 

Best Estimates, also known as OTIB-0033 (ORAUT 2005a). 

• If the worker’s file includes positive external dosimeter readings during a given calendar year 

that the worker was not monitored for internal dose, the worker should be treated as a 

radiation worker and a default missed internal dose (i.e., internal missed doses based on 

hypothetical bioassay data) should be assessed. The default missed internal dose is 

assessed by assuming that the most likely type of bioassay measurement was performed on 

the last day of the calendar year with a positive external dose and that bioassay measurement 

was negative. For the years from INEL’s startup through 1960 the most likely type of bioassay 

method was a urine sample analyzed for gross beta radioactivity; after 1960 the most likely 

type of bioassay method was a WBC. The missed intake and dose calculations are the same 

as described in Section 5.5 above. For readily non-compensable claims, a simpler but 

favorable to claimant alternative to assigning a default missed internal dose is to assign 

OTIB-0018 (ORAUT 2005b) based doses in accordance with the recommendations in 

OTIB-0033 (ORAUT 2005a) for unmonitored workers who were routinely exposed.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 49 of 63

• If a worker was not monitored for internal dose and had no detectable external doses reported 

for a given calendar year, only the environmental internal doses need to be assessed for that 

calendar year. 

The approaches outlined in this section are based on the determination that significant 

unmonitored intakes of radioactive material are unlikely at the INEL. This determination was 

based on evaluations of the most common types of bioassay that were employed throughout the 

INEL’s history and an evaluation of the results for those bioassay measurements. The INEL’s 

bioassay records indicate that over 90% of those bioassay results were below their respective 

detection limits (ORAUT 2010a, 2010b).

5.7 ATTRIBUTIONS AND ANNOTATIONS 

Where appropriate in this TBD, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction.

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this TBD, link data, quotations, and other information to documents available for 

review on the Project’s Site Research Database.

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

employed at the INEL and his work involved management, direction, or implementation of radiation 

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

activities at the site. This revision has been overseen by a new Document Owner, who is fully 

responsible for the content of this TBD, including all findings and conclusions. In all cases where such 

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

materials are fully attributed to the source.

Bryce Rich served as one of the initial Subject Experts for this document. Mr. Rich was previously 

employed at the INEL and his work involved management, direction, or implementation of radiation 

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

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

responsible for the content, including all findings and conclusions. Mr. Rich continues to serve as a 

Site Expert for this document because he possesses or is aware of information relevant for 

reconstructing radiation doses experienced by claimants who worked at the site. In all cases where 

such information or prior studies or writings are included or relied upon by Mr. Rich, those materials

are fully attributed to the source. Mr. Rich’s Disclosure Statement is available at www.oraucoc.org.

Boyd Levitt served as one of the initial Subject Experts for this document. Mr. Levitt was previously 

employed at the INEL and his work involved management, direction or implementation of radiation 

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

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

responsible for the content, including all findings and conclusions. In all cases where such 

information or prior studies or writings are included or relied upon by Mr. Levitt, those materials are 

fully attributed to the source.

Douglas Wenzel served as one of the initial Subject Experts for this document. Mr. Wenzel was 

previously employed at the INEL and his work involved management, direction or implementation of 

radiation protection and/or health physics program policies, procedures or practices related to atomic 

weapons activities at the site. This revision has been overseen by a new Document Owner who is 

fully responsible for the content, including all findings and conclusions. In all cases where such 

information or prior studies or writings are included or relied upon by Mr. Wenzel, those materials are 

fully attributed to the source.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 50 of 63

Henry Peterson served as one of the initial Subject Experts for this document. Mr. Peterson was 

previously employed at the INEL and his work involved management, direction or implementation of 

radiation protection and/or health physics program policies, procedures or practices related to atomic 

weapons activities at the site. This revision has been overseen by a new Document Owner who is 

fully responsible for the content, including all findings and conclusions. In all cases where such 

information or prior studies or writings are included or relied upon by Mr. Peterson, those materials 

are fully attributed to the source.

[1] Rich, Bryce. Principal Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of the referenced AEC annual reports. 

[2] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

The existence of the RESL, also known as the H&S Laboratory, HSL, and HSD, allowed for a 

consistency in assumptions about the internal dosimetry program because RESL was a 

constant even though there were a large number of varying facilities and continually changing 

contractors at the INEL. This is supported by the referenced AEC annual reports. 

[3] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement is based on the fact that the Na-24 is not liberated from the aluminum cladding 

and that Na-24 has a short half-life.

[4] Peterson, Henry. Principal Health Physicist. Intrepid Technology and Resources. November 

2004.

This statement is based on Mr. Peterson’s experience as a radiological engineer at TRA.

[5] Jenkins, JoAnn M. Senior Health Physicist. Dade Moeller & Associates. March 2007.

Because noble gases are inert, they do not interact with the lung when they are inhaled. The 

inhalation dose due to noble gases is due to the amount of gas that is inhaled and then 

exhaled. Because the volume of noble gases in the semi-infinite cloud around an individual is 

much greater than the inhaled volume, the external dose from noble gases is generally 

greater.

[6] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of dosimetry and incident records. Typically, 

large internal exposures can be related to unplanned events or planned releases, both of 

which are usually well documented.

[7] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on information from King (2001). It is important to use 

caution in considering worker controls based solely on radiological limits because these limits 

might not provide adequate protection from an industrial hygiene perspective. 

[8] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

The fuel at ICPP contained long-lived fission products with both alpha- and beta-emitting 

radionuclides. This allowed for the use of beta/gamma CAMs. It was understood that if 

beta/gamma airborne radioactivity was detected by the CAMs, the possibility of airborne alpha 

activity also existed. This is supported by the data in Table 5-18 that demonstrate that most of 

the isotopes were beta and gamma emitters.

[9] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

The delay before processing fuel allowed the short-lived radioisotopes to decay significantly, 

which left only the longer-lived radioisotopes as significant contributors to radiation dose.
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 51 of 63

[10] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on Hayden (1958).

[11] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on knowledge of the NRF from working at INEL. Because 

the NRF had operating reactors, the potential for internal exposure existed and, therefore, 

individuals who worked at the facility could have received internal doses.

[12] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on his experience as a member of the H&S management 

team that supported the INEL internal dosimetry program and is supported by the referenced 

AEC annual reports and internal dosimetry TBDs. It is common industry practice to take a 

proactive approach to the detection of ventilation failure. By monitoring the work area with 

continuous and retrospective air monitors, and using personnel and contamination monitoring, 

the facility H&S team provided real-time notification of ventilation failures and a means to 

assess personnel exposures. 

[13] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on his experience as a member of the H&S management 

team that supported the INEL internal dosimetry program and is supported by the referenced 

AEC annual reports and internal dosimetry TBDs.

[14] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of the referenced AEC annual reports. It is 

further supported by Stacy (2000). 

[15] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of the INEL dosimetry and bioassay records. 

It is further supported by the creation of the RCIMS database, which has a bioassay tracking 

function that went into effect in June 1999 and is described in Bhatt (2002).

[16] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This conclusion is based on a review of Andersen, Perry, and Ruhter (1995), INEEL (2001), 

King (1990) and Puphal (1994).

[17] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on his experience as a member of the H&S management 

team that supported the INEL internal dosimetry program and is supported by Puphal (1994). 

[18] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

Certain crafts were required site-wide. Examples of these are maintenance and construction 

workers such as pipefitters, plumbers, electricians, and health physics technicians. When a 

single contractor managed all or most of the INEL site, it was common for these types of 

personnel to perform work all over the site instead of at a specific facility. In these cases, 

personnel could have been exposed to a variety of radioactive material due to their varying 

work locations and situations.

[19] Rich, Bryce, and Jenkins, JoAnn M. Consulting Health Physicist and Senior Health Physicist. 

M. H. Chew & Associates and Dade Moeller & Associates. November 2004.

Internal doses are most accurately calculated when they are based on specific knowledge of 

exposure conditions such as specific radionuclides and quantities. When this information is 

available, it should be used in the dose reconstruction. When an employee worked at various 

locations, the specific bioassay data for each location should be used to calculate their internal 
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 52 of 63

dose. The statement by Mr. Rich about the equivalency of the bioassay programs at the 

different facilities is based on his professional experience as a member of the H&S 

management team that supported the INEL internal dosimetry program and is supported by 

the referenced internal dosimetry TBDs.

[20] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of the referenced AEC annual reports and 

internal dosimetry TBDs. CAMs measure the amount of radioactivity in the air and an alarm 

occurs when a preset level is reached. Retrospective air monitors also provide information on 

the levels of airborne activity, but the information is obtained after the fact when the air 

samples are analyzed. 

[21] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of the referenced AEC annual reports and 

internal dosimetry TBDs. 

[22] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of facility bioassay data and is supported by 

the referenced AEC annual reports and internal dosimetry TBDs.

[23] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of facility bioassay data and is supported by 

Puphal (1994). 

[24] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on information from an extensive review of INEL bioassay 

records.

[25] Rohrig, Norman, Consulting Health Physicist. Intrepid Technology and Resources. 

November 2004.

This plot was generated from noting the number of samples of different types in the database 

derived by the ORAU Team from copies of the bioassay results. 

[26] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of dosimetry records. It is also supported by 

the INEL incident reports. It is common industry practice to conduct incident investigations in 

a timely manner.

[27] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement is based on a review of the document titled INEEL M&O Contractor Technical 

Basis for Internal Dosimetry (INEEL 2001).
Document No. ORAUT-TKBS-0007-5 Revision No. 03 Effective Date: 03/02/2010 Page 53 of 63

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GLOSSARY

absorption 

In external dosimetry, process in which radiation energy is imparted to material. In internal 

dosimetry, movement of material to blood regardless of mechanism. 

absorption type 

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively in the 

respiratory tract (slow solubilization). Also called solubility type. See inhalation class.

activation 

Creation of a radioisotope by interaction of a stable (nonradioactive) element with neutrons, 

protons, or other types of radiation. 

beta radiation 

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

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

beta particle is a positron. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

breeder reactor 

Nuclear reactor that makes more new fissionable material than it consumes. 

calcine 

(1) Dry solid (grainy or granular) product of a chemical process that removes liquids from a 

solution. (2) Process for creating the chemical reaction that removes liquids from a solution.

cladding 

The outer layer of metal that encases a reactor fuel element or fissile material of the pit of a 

nuclear weapon, often made with aluminum or zirconium. In a reactor, cladding promotes the 

transfer of heat from the fuel to the coolant, and it builds up fission and activation products 

over time from the fission of the fuel. 

contamination 

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

persons.

core 

Central region of a nuclear reactor where fission of the fuel takes place. 

criticality 

State of a radioactive mass (e.g., the core of a nuclear reactor) when the fission reaction 

becomes self-sustaining. Nuclear reactors go critical when started.
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curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7E+10) becquerels, which is approximately 

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

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 a period of time.

depleted uranium (DU) 

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

dosimeter 

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

received by an individual. 

dosimetry 

Measurement and calculation of internal and external radiation doses.

enriched uranium

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7%. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade uranium 

contains greater than 90% 235U.

fission 

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons.

fission product

(1) Radionuclides produced by fission or by the subsequent radioactive decay of 

radionuclides. (2) Fragments other than neutrons that result from the splitting of an atomic 

nucleus. 

gamma radiation 

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are very 

penetrating, but dense materials such as lead or uranium or thick structures can stop them. 

Gamma photons are identical to X-ray photons of high energy; the difference is that X-rays do 

not originate in the nucleus. 

half-life

Time in which half of a given quantity of a particular radionuclide disintegrates (decays) into 

another nuclear form. During one half-life, the number of atoms of a particular radionuclide 

decreases by one half. Each radionuclide has a unique half-life ranging from millionths of a 

second to billions of years. 

hot cell 

Shielded laboratory for handling of radioactive materials with the aid of remotely operated 

manipulators. The walls and windows are made of materials that protect workers from 

radiation. 
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in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay

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

detects radiation emitted from the radioactive material in the body. 

ionizing radiation 

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

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

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

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

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

irradiated fuel

Nuclear reactor fuel that has been exposed to neutron radiation in a nuclear reactor and has 

undergone fission reactions. Irradiated fuel often contains hazardous levels of activation and 

fission products that make handling it difficult. Nuclear reactor fuel that has been irradiated to 

the point where it is no longer useful in sustaining a nuclear reaction is typically referred to as 

spent fuel. See spent fuel. 

isotope 

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., U-234, U-235, and 

U-238). Isotopes have very nearly the same chemical properties but often have different 

physical properties. 

microPrefix that divides a unit by 1 million (multiplies by 1E-06).

milliPrefix that divides a unit by 1,000 (multiplies by 1E-03).

mixed waste 

Unwanted material containing both radioactive and hazardous components. 

natural uranium (U, U-nat, NU) 

Uranium as found in nature, approximately 99.27% U-238, 0.72% U-235, and 0.0054% U-234

by weight. The specific activity of this mixture is 2.6E+07 becquerel per kilogram (0.7 

picocuries per gram). 

neutron (n) 

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen. 

nucleus 

Central core of an atom, which consists of positively charged protons and, with the exception 

of ordinary hydrogen, electrically neutral neutrons. The number of protons (atomic number) 

uniquely defines a chemical element, and the number of protons and neutrons is the mass 

number of a nuclide. The plural is nuclei.
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nuclide 

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

photon 

Basic unit of electromagnetic radiation. Photons are massless “packages” of light energy that 

range from low-energy microwave photons to high-energy gamma rays. Photons have 

energies between 10 and 100 kiloelectron-volts. 

proton 

Basic nucleic particle with a positive electrical charge and mass slightly less than that of a 

neutron. There are protons in the nuclei of every atom, and the number of protons is the 

atomic number, which determines the chemical element. 

radiation 

Subatomic particles and electromagnetic rays (photons) that travel from one point to another, 

some of which can pass through or partly through solid materials including the human body. 

See ionizing radiation.

radioactivity 

A property possessed by some isotopes of some elements (e.g. K-40, C-14, Cs-137, etc...). 

The process of undergoing spontaneous transformation of the nucleus, generally with the 

emission of alpha or beta particles, often accompanied by gamma rays. This term is also used 

to designate radioactive materials. See radionuclide.

radionuclide 

Radioactive nuclide. See radioactivity and nuclide. 

radioactive lanthanum (RaLa)

Lanthanum-140 was used in diagnostic tests of the implosion mechanism of a nuclear fission 

bomb. Inside the high-explosive shell, a core of La-140 replaced the usual plutonium core, 

and gamma radiation from the lanthanum provided information on the course of the implosion.

radioactive waste 

Radioactive solid, liquid, and gaseous materials for which there is no further use. Wastes are 

generally classified as high-level (with radioactivity as high as hundreds of thousands of curies 

per gallon or cubic foot), low-level (in the range of 1 microcurie per gallon or cubic foot), 

intermediate level (between these extremes), mixed (also contains hazardous waste), and 

transuranic. 

rem 

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

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

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural.

reprocessing 

Mechanical and chemical processing of irradiated nuclear fuel to separate useable fissionable 

products (i.e., uranium and plutonium) from waste material. Reprocessing was discontinued in 

the United States in 1992.
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shielding 

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

spent fuel

Nuclear reactor fuel that has been irradiated in a nuclear reactor to the point where it is no 

longer useful in sustaining a nuclear reaction. See irradiated fuel. 

spent fuel storage basin 

Pool or pit of reinforced concrete filled with water for storage of irradiated nuclear fuel. The 

water is shielding and coolant. 

transuranic (TRU) waste 

Radioactive waste that contains transuranic elements and has radioactivity of 100 or more 

nanocuries per gram.

U.S. Atomic Energy Commission (AEC) 

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

X-ray radiation 

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

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

shell transitions within atoms (characteristic X-rays). The only difference between X-rays and 

gamma rays is their point of origin. X-rays originate in the electron shells of an atom, whereas, 

gamma rays originate in the nucleus of an atom. 

zirconium 

Metallic element with atomic number 40. Zirconium is highly resistant to corrosion, and it is 

alloyed with aluminum to make cladding for nuclear fuel and sometimes in small amounts with 

the fuel itself.