ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 60

Document Title:

Idaho National Laboratory – 

Occupational Internal Dose

Document Number: ORAUT-TKBS-0007-5 

Revision: 02

Effective Date: 07/30/2007

Type of Document TBD

Supersedes: Revision 01 

Subject Expert(s): Jo Ann M. Jenkins 

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

Approval: Signature on File Approval Date: 07/05/2007

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 07/03/2007

John M. Byrne, Task 3 Manager 

Concurrence: Signature on File Concurrence Date: 07/06/2007

Edward F. Maher, Task 5 Manager 

Concurrence: Signature on File Concurrence Date: 07/27/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 07/30/2007

James W. Neton, Associate Director for Science 

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 2 of 60

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:

Norman D. Rohrig, Document Owner

Signature on File 09/30/2004

Judson L. Kenoyer, Task 3 Manager

Signature on File 10/01/2004

Richard E. Toohey, Project Director

Signature on File 10/01/2004

James W. Neton, Associate Director for Science

Signature on File 10/12/2004

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 PC1. 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:

John M. Byrne, Document Owner

Signature on File 05/23/2006

John M. Byrne, Task 3 Manager

Signature on File 05/23/2006

Edward F. Maher, Task 5 Manager

Signature on File 05/22/2006

Kate Kimpan, Project Director

Signature on File 05/24/2006

James W. Neton, Associate Director for Science

Signature on File 06/05/2006

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. 02 Effective Date: 07/30/2007 Page 3 of 60

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

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

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

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

5.2 Historical Overview .....................................................................................................................9

5.2.1 Test Reactors............................................................................................................... 10

5.2.2 Highly Enriched Spent Fuel Chemical Processing....................................................... 11

5.2.3 Other Nuclear Facilities and Processes ....................................................................... 11

5.2.4 Radionuclides of Concern and Solubility...................................................................... 11

5.3 Radiological Protection Program Management and Support....................................................13

5.3.1 Internal Dosimetry Issues Related to Contractor Changes .......................................... 13

5.3.2 Bioassay Programs ...................................................................................................... 14

5.3.3 Internal Dose Records ................................................................................................. 14

5.3.4 Internal Dose Regulations, Investigation Levels, and Data Codes .............................. 16

5.4 Internal Dose Control................................................................................................................19

5.4.1 Air Monitoring............................................................................................................... 20

5.4.2 Early Technical and Analytical Capabilities at the National Reactor 

Testing Station ............................................................................................................. 21

5.5 Minimum Detectable Activities ..................................................................................................24

5.6 Development and Impact of Whole-Body Counting ..................................................................24

5.7 Specific Facilities ......................................................................................................................28

5.7.1 Test Area North............................................................................................................ 28

5.7.1.1 Specific Manufacturing Capability Project ......................................................29

5.7.2 Idaho Nuclear Technology and Engineering Center .................................................... 31

5.7.2.1 Highly Enriched Spent Fuel Storage ..............................................................32

5.7.2.2 High-Level Wastes .........................................................................................33

5.7.2.3 High-Level Waste Calcination ........................................................................33

5.7.2.4 Process Analytical Facilities ...........................................................................33

5.7.2.5 Spent Fuel Processing and INTEC Most-Limiting 

Radionuclides................................................................................................. 33

5.7.3 Argonne National Laboratory–West ............................................................................. 35

5.7.4 Radioactive Waste Management Complex .................................................................. 35

5.7.5 Waste Reduction Operations Complex ........................................................................ 36

5.7.6 Test Reactor Area........................................................................................................ 37

5.7.6.1 Test Reactor Area Reactors...........................................................................37

5.7.6.2 Test Reactor Area Laboratories .....................................................................38

5.8 Default for Missed Dose............................................................................................................39

5.9 Unmonitored Workers ...............................................................................................................41
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 4 of 60

5.10 Attributions and Annotations .....................................................................................................41

References ...........................................................................................................................................50

Glossary................................................................................................................................................56

LIST OF TABLES

TABLE TITLE

5-1 Recorded internal doses for 1992 to 2000................................................................................

PAGE

12

5-2 Primary radionuclides of concern ..............................................................................................13

5-3 Routine bioassay history summary ...........................................................................................15

5-4 Internal dose assessment information after 1989.....................................................................17

5-5 Internal dose assessment information before 1989..................................................................17

5-6 Analytical information that could be in worker dose files ..........................................................18

5-7 Area codes that could be in worker dose files ..........................................................................18

5-8 Derived investigation levels in 1977 for acute exposures .........................................................19

5-9 Surface contamination control and MDAs.................................................................................20

5-10 Urinalysis results in 1959, 1960, and 1961 ...............................................................................23

5-11 Detection limits applicable to environmental sample analyses, 1953 to 1965 ..........................24

5-12 MDAs for urine samples by period............................................................................................25

5-13 MDAs for fecal samples by period ............................................................................................26

5-14 In vivo MDAs by period.............................................................................................................27

5-15 Summary statistics from the 1963 WBC program.....................................................................29

5-16 Mass and activity ratios of SMC DU isotopes ...........................................................................30

5-17 Notable airborne incidents at INTEC ........................................................................................31

5-18 Radiologically significant radionuclides for INTEC-processed fuels .........................................32

5-19 Radionuclides of concern and MDAs for ANL-W locations.......................................................35

5-20 Radioactive waste inventory in the TSA ...................................................................................36

5-21 Radioactive low-level waste inventory in the active pits in the SDA .........................................36

5-22 Radiologically significant radionuclides for ATR fuel gaseous releases ...................................38

5-23 Major airborne incidents at TRA ...............................................................................................38

5-24 Default table for missed dose ...................................................................................................39

LIST OF FIGURES

FIGURE TITLE

5-1 History of urine bioassay...........................................................................................................

PAGE

22
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 5 of 60

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

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

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

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 6 of 60

IDO Idaho Operations Office

IET initial engine test

IMBA Integrated Modules for Bioassay Analysis

in. inch

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

mCi millicurie

MDA minimum detectable activity

MDL minimum detectable level

MeV megavolt-electron, 1 million electron-volts

MFP mixed fission product

mg milligram

mi mile

min minute

mL milliliter

mo month

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPCa MPC for airborne activity

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

PBF Power Burst Facility

pCi picocurie

PCS primary coolant system

POC probability of causation

RaLa radioactive lanthanum

RAM radiation (or remote) area monitor

RCGa radioactive concentration guide [for airborne activity]

RCIMS Radiation Control Information Management System

RDR Radiation Dosimetry and Records

RESL Radiological Environmental Sciences Laboratory

RWMC Radioactive Waste Management Complex
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 7 of 60

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

TLV threshold limit value

TMI Three Mile Island

TRA Test Reactor Area

TREAT Transient Reactor Experiment and Test

TRU transuranic

TSA Transuranic Storage Area

U.S.C. United States Code

WBC whole-body counting

WERF Waste Experimental Reduction Facility

yr year

ZPPR Zero Power Plutonium (later Physics) Reactor 

α alpha particle

β beta particle 

γ gamma

σ standard deviation

μCi microcurie 

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 8 of 60

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 for particular sites or categories of sites. They will be revised in the event 

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

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

derived:

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

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

• Radiation from naturally occurring radon present in conventional structures

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 9 of 60

5.1.1

This TBD provides specific information concerning documentation of historical practices at the Idaho 

National Laboratory (INL).

Purpose

5.1.2

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

through increased general understanding, data interpretation, defaults, and so forth. Sections 5.2.1 to 

5.2.3 provide facility descriptions, and Section 5.2.4 details the radionuclides of concern. Section 5.3 

describes the INL radiological protection program as it evolved over the years, Section 5.4 discusses 

internal dose control, and Sections 5.5 and 5.6 describe minimum detectable activities (MDAs) and 

whole-body counting (WBC), respectively. Section 5.7 contains specific information necessary for 

dose reconstruction when the facility or facilities where the employee worked are known. Sections 5.8 

and 5.9 describe the treatment of missed dose and unmonitored workers.

Scope

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

justification, or clarification of the associated information, are presented in Section 5.10. 

5.2 HISTORICAL OVERVIEW

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

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

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

and then the Idaho National Engineering and Environmental Laboratory. In 2005, DOE shortened the 

name to the Idaho National Laboratory. For convenience, this TBD uses INL where it is unnecessary 

to distinguish.

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 INL that had a 

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

• The original mission of the NRTS was (as the name implied) highly enriched (over 50% and 

mostly over 90%) uranium reactor concept development, materials testing through high-flux 

test reactor operation, and chemical processing of those highly enriched (valuable) uranium 

fuels. The production of weapons-grade nuclear materials was not a mission.

• The NRTS began operations 8 to 10 yr after Oak Ridge National Laboratory 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 Oak Ridge (ACC 1952) were adopted at the startup of the 

NRTS facilities. 

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

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

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

program development. There were thus three Federal organizations utilizing the NRTS and its 

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

personnel received exposure from AEC operations. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 10 of 60

• 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 NRTS among a large variety of 

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

Laboratory at NRTS 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 AEC HSL and each 

contractor [2].

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

or minimum detectable levels (MDLs), missed dose potential, and so forth are relatively consistent 

across the years [3]. There were differences in the available nuclear materials from facility to facility, 

but as early as 1955 or 1956 gamma spectral analysis capabilities at the NRTS allowed the significant 

bioassay results (those which would result in reportable internal dose) to be defined in terms of the 

specific radionuclides. The practice in the case of a higher urine sample result was to attempt 

radionuclide identification through gamma spectral analysis and chemical separation. This document 

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

the radionuclide analyses and only record gross beta or gross alpha results [4].

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

involved in the incident [5]. 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.

5.2.1

NRTS, which was the primary nuclear reactor development laboratory in the United States, tested or

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

constructed, and operated (including operation-to-destruction tests) on the NRTS. The INL site 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 Argonne National Laboratory-West (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). 

Test Reactors
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 11 of 60

The largest internal exposures at INL 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.

5.2.2

In addition to being the site for experimental test reactors, INL is the home of the Idaho Nuclear 

Technology and Engineering Center (INTEC), formerly known as the Idaho Chemical Processing 

Plant (ICPP). INTEC reprocessed highly enriched reactor fuel (235U enrichments of 50% to 93%) for 

39 yr from 1952 to 1991. Aged mixed fission products (MFPs) were the predominant internal hazard, 

although enriched uranium isotopes and plutonium isotopes (238Pu enhanced) were limiting in specific 

process locations. INTEC 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. 

Highly Enriched Spent Fuel Chemical Processing

5.2.3

Other nuclear facilities at INL that resulted in internal dose potential or experience include:

Other Nuclear Facilities and Processes

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

generated by nuclear facilities on INL 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) 

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 uranium 

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

occurred.

• At the Naval Reactors Facility (NRF), the U.S. Navy used the INL for operating reactors and as 

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

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

However, through the years NRF has participated in limited coordination of radiological 

protection programs and site support services. It is possible that some workers’ internal dose 

could have resulted from their support work at the NRF [7].

5.2.4

INL facilities and activities have related primarily to experimental reactor design and development, 

spent fuel processing, DU parts production, and low- and high-level radioactive waste treatment and 

disposal. The latest revision of the INL Site Description TBD (ORAUT 2005) describes these activities 

in more detail. Table 5-1, which lists the internal doses received at the INL from 1992 to 2000, 

demonstrates that the radionuclides of more recent concern to internal dose are MFPs (primarily 

aged), uranium and its decay products, and TRU wastes and their decay products. The table lists 

internal doses as committed effective dose equivalents (CEDEs). This information is from 

Radionuclides of Concern and Solubility
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 12 of 60

INEEL (2001) and can be useful to the reconstruction effort primarily as an indicator of intake 

radionuclide identity and quantities for this recent 10-yr period. Before 1990, workers received higher 

doses from the radionuclides in Table 5-1, as well as from MAPs (primarily longer-lived, i.e., 110mAg 

and others) [8]. Most internal doses have been identified following an incident rather than as a result 

of routine bioassay measurements (INEEL 2001).

Table 5-1. Recorded internal doses for 1992 to 2000 (INEEL 2001, p 37).a

Year of dose 

assignment Facility CEDE (mrem) Radionuclides of concern

2000 SMC 10, 31, 20, 10, 13, & 20 U-238

2000 INTEC 11 Pu-238, Pu-239/240

1999 SMC 15, 48, 13, & 12 U-238

1999 INTEC 16 Pu-239/240

1998 SMC 16 U-238

1997 TRA 10a & 10 a Eu-152 and Eu-154

1997 INTEC 24 Pu-238 and Pu-239/240

1997 TAN 13 a Am-241, Cs-137, Sr-90, U-233, U-238, Pu238, Pu-239/240

1997 SMC 16 & 20 U-238

1996 RWMC 43 Pu-239

1996 INTEC 15, 87, 136, 652a, 655a, 677a, & 678a Am-241, Cs-137, Sr-90, Pu-238, 

Pu-239/240

1996 SMC 10, 10, 12, 16, 17, 18, 20, 20, & 23 U-238

1995 INTEC 10, 13, 14, 15, 23, 28, 29, 42, 45, & 53 Am-241, Cs-137, Sr-90, Pu-238, 

Pu-239/240

1995 SMC 10, 12, 14, 15, 19, & 26 U-238

1994 INTEC 14, 20, 25, & 29 Am-241, Cs-137, Pu-238, Pu-239/240

1994 SMC 10, 10, 12, & 15 U-238

1993 INTEC 14, 35, 36, 39, 50, & 53 Am-241, Pu-238, Pu-239/240

1993 SMC 11 U-238

1992 RWMC 20 & 205 Am-241, Pu-238, Pu-239/240

1992 SMC 11, 12, 12, 14, 15, 15, 16, 20, 32, & 52 U-238

a. Internal doses determined after a “significant” event when intake occurred. However, even the other intakes were 

defined as a result of contamination incidents or high airborne measurements.

Table 5-2 lists radionuclides of concern at INL 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 INL program follows the DOE 

guidance and uses the International Commission on Radiological Protection (ICRP) Publication 30 

nomenclature for solubility (ICRP 1979) as noted in the table. In addition, Table 5-2 provides the 

ICRP Publication 68 (ICRP 1995) recommended absorption type.

Through the years at INL, plotting urine and fecal elimination curves has shown that an effective 

solubility of moderate (M) to slow (S) per the ICRP Publication 66 system (ICRP 1994) is to be 

expected for the radionuclides listed in Table 5-2, with the exceptions of strontium, cesium, iodine, 

mercury, and tritium. Elimination curves for 3

H, 90Sr, 131I, 137Cs, and 203Hg show they are typically of a 

fast absorption type (F). The chemical explanation is that radioactive materials in oxygen 

atmospheres oxidize rapidly, which forms less soluble compounds. The default assumption of type M 

or S would be appropriate when based on the result to the organ in question that is more favorable to 

the claimant.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 13 of 60

Table 5-2. Primary radionuclides of concern. 

Element Radionuclides

ICRP 30a

retention 

class

ICRP 68 

absorption 

type

Preferred 

analysis technique

Hydrogen H-3 [assume tritiated water vapor (HTO)] D F Urine

Chromium Cr-51 F,M,S WBC

Manganese Mn-54 F,M WBC

Iron Fe-59 F,M WBC

Cobalt Co-58, Co-60 W,Y M,S WBC

Zinc Zn-65 S WBC

Strontium Sr-89, Sr-90/Y-90b D F Urine

Zirconium/niobium Zr/Nb-95 W,Y F,M,S WBC

Molybdenum Mo-99 F,S Urine

Technetium Tc-99 F,M Urine

Ruthenium Ru-103, Ru-106 F,M,S Urine

Silver Ag-110m F,M,S WBC

Antimony Sb-122, Sb-125 F,M WBC

Tellurium Te-132 F,M Urine, WBC

Iodine I-131, I-133 (assume elemental I2) F Urine, thyroid counter

Cesium Cs-134, Cs-137 D F WBC

Barium/lanthanum Ba/La-140 F/F,M WBC

Cerium Ce-141, Ce-144 W,Y M,S WBC

Europium Eu-152, Eu-154, Eu-155 W,Y M WBC

Gadolinium Gd-153 F,M WBC

Tantalum Ta-182 M,S WBC

Mercury Hg-203 (assume inorganic) F Urine, WBC

Protactinium Pa-233 M, S WBC

Uranium U-233, U-234, U-235, U-236, U-238 D,W,Y F,M, S Urine, fecal

Neptunium Np-237 M Urine, fecal

Plutonium Pu-238, Pu-239/240 W,Y M, S Urine, fecal

Americium Am-241 W M Urine, fecal, lung count

a. Assigned in the INL internal dosimetry technical basis document (INEEL 2001); ICRP 30 = ICRP 1979; ICRP 68 = ICRP 

1995.

b. Yttrium has an absorption type of M or S; however, in the case of short-lived Y-90, it tracks its parent Sr-90, which is 

normally a type F at the INL.

5.3 RADIOLOGICAL PROTECTION PROGRAM MANAGEMENT AND SUPPORT

5.3.1

The changes in contractors at INL during its 54-yr history [listed in Tables 2-1 and 2-2 of the Site 

Description TBD of this Site Profile (ORAUT 2005)] resulted in relatively frequent management 

changes at most of the facilities. (The contract with the University of Chicago to operate the ANL-W 

facilities has not changed in this 54-yr period.) 

Internal Dosimetry Issues Related to Contractor Changes

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

initially assigned to the AEC Idaho Operations Office (IDO) [9]. 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 [10]. The Chicago Operations Office (COO) has provided oversight for 

ANL-W (contracted by the University of Chicago) programs and facilities, while the U.S. Navy has 

provided oversight for the NRF program [11]. ANL-W, which is a DOE program, is included in this 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 14 of 60

TBD. The ANL-W program uses site support services, including internal dosimetry support, but the 

bioassay results are reported through COO [12]. 

The INL personnel dosimetry records have been and are documented and permanently maintained by 

the various organizations over time [13]. 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 report. 

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 INL, particularly the bioassay analytical techniques and calculation processes [14]. 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 [15]. 

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 [16]. 

Internal dose reconstruction for personnel who worked at a number of INL 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 [17]. 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). In the cases where only gross beta or gross alpha bioassay 

results were available, the results were normally within 2σ [18]. If it is necessary to evaluate intakes 

from the gross beta or gross alpha results, the radionuclide defaults should be 90Sr/90Y and 238Pu, 

respectively. 

5.3.2

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 mo in 1953 [19]. 

Table 5-3 lists the reconstructed history of routine bioassay frequency.

Bioassay Programs

5.3.3

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 INL 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 

Internal Dose Records
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 15 of 60

oversight of INL internal dosimetry program functions and provides quality assurance. The RDR unit 

functions include documentation and records custodial responsibilities. In 1999, the Radiation Control 

Table 5-3. 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

Annual DE >10% quarterly 

standard in ERDA Manual 

Chapter 0524 (ERDA 1975).

Reporting 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 50-yr CDE >10% quarterly 

standard in ERDA Manual 

Chapter 0524 (ERDA 1975).

Reporting 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.

1988–

1989 

1 to 6 mo In vivo All radiation workers.

Lung, 50-yr CDE 

>0.5–1.0 rem.

Investigating

Bone surface, 50-yr CDE 

>1.0–2.0 rem.

Other organs, 50-yr CDE 

>0.5-1.0 rem.

Staggered so that 

worker receives 

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.

In accordance with DOE 

Order 5480.11 (DOE 1988). 

Reporting

Workers that could receive 

0.1 rem AEDE or 5 rem ADE 

organ or tissue dose. 

Bioassay requested 

when workplace 

monitoring program 

indicates >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.

AEDE ≥0.01 rem. 

Investigating

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.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 16 of 60

Year

Typical 

frequency Type Groups analyzed/sampled Investigating level Comments Reference

1995 Appropriate to 

the facility 

mission, 

potential 

uptakes.

In vivo All radiation workers that 

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. 

In accordance with DOE 

5480.11 (DOE 1988) and 

10 CFR Part 835: 

Reporting

Workers that could receive 

0.1 rem CEDE. 

Declared pregnant workers 

when embryo/fetus could 

receive 0.05 rem DE. 

Each facility has a 

specific Technical 

Basis Document for 

Internal Dosimetry. 

Andersen, 

Perry, and 

Ruhter 1995 

In vitro urine; 

fecal

When 

workplace 

monitoring 

indicated 

significant 

potential for 

intakes. 

Follow-up for any 

suspected intake of 

radionuclides and to 

more accurately 

identify and 

characterize the 

amount of intake 

and excretion 

pattern. 

Internal doses resulting from 

all confirmed intakes are to 

be evaluated. 

Investigating

Random sampling is 

performed to demonstrate the 

adequacy of the radiological 

controls in limiting the internal 

intake of radionuclides. 

Employees are selected at 

random from both 

nonradiation workers and a 

radiation worker population.

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 that 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. 

In accordance with DOE 

5480.11 (DOE 1988) and 10 

CFR Part 835. 

Reporting

Workers that 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 that was on a 

scheduled monitoring 

program.

In vitro 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.

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).

5.3.4

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.

Internal Dose Regulations, Investigation Levels, and Data Codes
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 17 of 60

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

years. Table 5-4 describes coded information that could appear in records after 1989. Table 5-5 

describes internal dose information that could appear in pre-1989 records. Table 5-6 contains 

analytical nomenclature, and Table 5-7 contains INL codes for various site areas.

Table 5-4. 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.

Table 5-5. 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.

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

U.S. Energy Research and Development Administration (ERDA), and the DOE changed periodically 

during the site’s history. While the regulations influenced the level of calculated internal dose that 

would be reported, these changes did not significantly affect the analytical programs for the detection 

of internal intake [20]. For example, the monitoring and analytical programs were designed to initiate, 

through in vitro and/or in vivo bioassay analysis, an investigation of any potential internal intake as 

indicated by positive air sampling, personnel contamination, etc. Most of these recorded analyses did 

not result in detectable radionuclides [21]. 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) 

Regulations 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, 

1963b, 1968, 1975; ERDA 1975). Changes in the reporting levels did not generally result in changes 

to the air-monitoring or bioassay programs [22]. Each individual analytical result was documented 

and placed in individual exposure files regardless of the formal reporting requirements.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 18 of 60

Table 5-6. Analytical information that could be in worker dose files. 

Analytical information Description

Sample no. Sample log number.

Date and time Generally clear interpretation.

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

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

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

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

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

column.

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

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

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

either case.

Total count Total number of counts recorded.

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

Bkgd., cpm Background cpm recorded.

Net count, cpm Gross cpm minus background cpm.

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

product; ignore for an internal dose reconstruction

Foreign activity, cpm 

and dpm

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

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

counting statistics.

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

Result in μg/L These results are for uranium whether stated or not.

Table 5-7. Area codes that could be in worker dose files. 

Area code Descriptiona Area code Descriptiona

1 AEC Headquarters Building 20, 261, 264 TREAT

2 EBR-I 21 LX

3, 034, 035 CFA 22 GCRE

4, 042, 045 MTR, TRA 23 OX

5, 053, 055 INTEC (ICPP) 24 ARHG

6 NRF 25 No information available

7 TAN (General Electric) 26, 263, 265 EBR-II

8 Services 27 ML-1

9 NX (X is construction) at NRF 28 On-site site survey

10 AX at TAN 29 Off-site site survey

11, 113 CX at CPP 30 ANP program at SL-1

12 EX at EBR 31 STPF

13, 133, 135 SPERT, PBF 65 ECF

14 OMRE 66 Non-security

15 SX at SPERT 67 Division of Compliance

16 SL-1 68 STEP

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

18, 814, 815 WP, RWMC 71 CADRE (guard force)

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

a. See the acronyms and abbreviations list.

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 [23]. Dickson (1977) established official investigation levels (Table 5-8) for acute 

uptakes of radionuclides corresponding to one-tenth of the quarterly radiation standard. Later 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 19 of 60

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). These 

procedural limits did not materially affect the bioassay sampling frequency or the recording of even 

undetectable radioactivity in bioassay samples, although the request for and number of follow-up 

samples and analyses could have been different as a function of the formal regulations in effect [24].

Table 5-8. Derived investigation levels (μCi) in 1977 for acute 

exposures (Dickson 1977). 

Radionuclide

Inhalation 

(lung burden)

Ingestion 

(total activity)

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

5.4 INTERNAL DOSE CONTROL 

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 [25]. 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 INL for internal 

exposures has been preventive in nature [26]. 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 

materials of low specific activity. The consistent INL 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).

In a related matter, 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 5-9 is 

a summary of control limits primarily from the CPP [Chemical Processing Plant] Health Physics 

Manual (ACC 1952) and current operating procedures.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 20 of 60

5.4.1

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

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

beginning of all facility and program operations in routinely occupied areas. With the exception of the 

1985 SMC Project, the primary contaminant radionuclides by activity were either MFPs or MAPs, 

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

Air Monitoring

Table 5-9. Surface contamination control and MDAs. 

Period

Surface 

location

Detection 

technique Control levels Typical MDA

1952–

1960s

Plant/equipment Smears 500 dpm β & 

20 dpm α per 100 cm2 

150 dpm β & 

10 dpm α per 100 cm2

Personal 

clothing

Portable survey 

instruments

1500 dpm β & 

500 dpm α per 100 cm2

1,000 dpm β & 

500 dpm α per 100 cm2

Personal skin Portable survey 

instruments

Any detectable reported, e.g. 

1,000 dpm β & 

500 dpm α per 100 cm2 

1,000 dpm β & 

500 dpm α per 100 cm2

Shipments Smears 500 dpm β

& 20 dpm α per 100 cm2

150 dpm β & 10 dpm α per 

100 cm2 

Portable survey 

instruments

0.1 mrep/hr β & 

500 dpm α per 100 cm2

0.01 mrep/hr β

& 500 dpm α per 100 cm2

1970s–

present

Plant/equipment

surfaces

Smears 300 dpm β & 

20 dpm α per 100 cm2

30 dpm β & 

10 dpm α per 100 cm2

Personnel Portable survey 

instruments

Any detectable reported, e.g. 

300 dpm β & 

200 dpm α per 80-100 cm2 

300 dpm β & 

200 dpm α per 80–100 cm2

concentrations (MPCs/DACs) above 1 × 10-9 μCi/cm3. TRU materials and uranium were available at 

some of the INL facilities, but they were nearly always well-tagged with beta/gamma activity that 

allowed beta/gamma-detecting CAMs to be used to warn of possible alpha contamination or internal 

exposures.

ACC (1952) describes a CAM and three other air-sampling systems. The manual required use of a 

filter-type respirator when airborne activity exceeded 1 × 10-8 μCi/cm3 for beta/gamma activity or 

1 × 10-11 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, p IX: 4-1) (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 [27]. The fixed air filter 

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

facilities where alpha contaminants could be controlling [28]. 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 [29]. 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 [30].
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 21 of 60

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

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

need for bioassay if elevated air sample results were detected [31]. 

5.4.2

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

documented the analytical detection capability of NRTS 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.

Early Technical and Analytical Capabilities at the National Reactor Testing Station

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). 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 3- by 3-in. sodium iodide 

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

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

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

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

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

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

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

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

at risk (radiation workers) every 2 yr 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 8 × 10-8 μCi/mL 

of 90Sr. 

As part of the EEOICPA coworker data program for the INL, images of all bioassay results from 

before 1986 were provided to the Oak Ridge Associated Universities (ORAU) Team, and the results 

were entered into a database. This database has over 150,000 bioassay results. The nearly 135,000 

urine bioassay results fall into over 80 categories of analytes, a few of which are the result of minor 

typographical differences. Figure 5-1 shows the time course of the gross beta, gross gamma, and 

strontium analyses that were performed most often. It appears that the group beta/gamma protocol 

seems to have followed the beta protocol in the early years and the gamma protocol after 1964. The 

early 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 [32]. In the early 1970s, in 

vitro bioassay was largely replaced by WBC [33]. In the 1950s and 1960s, several thousand assays 

were performed each year, well over one per monitored worker per year. The early results were 

reported in dpm per sample except for certain assays, such as uranium reported in grams U per liter 

and tritium in dpm per milliliter [34]. About September 1973 the activity units changed to microcuries 

with a large negative exponent so that picocuries would be a more natural unit. There are some 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 22 of 60

INEEL Urine Bioassay History

10

100

1000

10000

50 55 60 65 70 75

Year

Number Samples Analyzed

beta

Gamma

beta gamma

Strontium

Figure 5-1. History of urine bioassay [35]. 

confusions in the database record between dpm and microcuries and between per-sample or permilliliter results, so if the results or uncertainties appear anomalous, that is a likely explanation. The 

sample size for alpha spectrometry went from the urine sample size (about a liter) to 1 (meaning all of 

the sample, not 1 mL) in mid-1979. 

Table 5-10 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 

workers provided urine samples multiple times during the year.

Table 5-11 summarizes early detection limits for various environmental water and milk samples 

between 1953 and 1965. These analytical sensitivities are applicable to analysis of body fluids or 

substances for the purpose of bioassay (Dodd 1964).

The special and routine bioassay sample analyses 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 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).
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 23 of 60

Table 5-10. 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. If only gross γ analyses are available, the default should be Cs-137.

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. 02 Effective Date: 07/30/2007 Page 24 of 60

Table 5-11. Detection limits applicable to 

environmental sample analyses, 1953 to 1965 (Dodd

1964, p. 71). 

Type of sample Radiation Detection limit

Water Alpha 3E-9 μCi/mL

Beta 6E-9 μCi/mLa

Tritium 4E-6 μCi/mL

Milk Iodine-131 10 pCi/Lb

Strontium-90 1.5 pCi/L

a. Reduced in October 1962 from 1.5E-7 μCi/mL.

b. Reduced in September 1962 from 50 pCi/L.

The individual analytical data were recorded, as were the specific doses for which formal evaluations 

were required. Aoki (1979) states that before 1977, internal doses were assigned if the internal dose 

or committed dose was greater than 50% of the yearly body or organ doses allowable per ERDA 

Manual Chapter 0524 (ERDA 1977), which was the controlling regulation at the time. In 1977, the 

policy was changed to assign internal doses when the CDE to an organ exceeded one-tenth of the 

quarterly radiation protection standard (Aoki 1979). Replies to requests for radiation exposure history 

before 1979 stated that there was “no positive exposure reported” when the dose was below the 

reporting levels noted above (Aoki 1979). Because individuals remembered receiving some internal 

exposure, the statement, “no positive exposure reported,” was determined to be misleading and was 

changed in 1977 to “no reportable levels recorded” (Aoki 1979). In all cases, copies of the bioassay 

results were placed in the individual’s radiation exposure file and should be in the individual’s dose file 

in the NIOSH Office of Compensation and Support Claims Tracking System (NOCTS) database.

5.5 MINIMUM DETECTABLE ACTIVITIES

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 N 13.30, Performance Criteria for Radiobioassay (HPS 1996), both the in vitro and 

in vivo radiobioassay laboratories at INL received DOELAP accreditation in February 1998 (INEEL 

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

level (2σ) are performed. Tables 5-12 and 5-13 list current MDAs for urine and fecal sample analysis, 

respectively, along with values gleaned from historical documents. A large majority of the urine 

samples taken at the site were single voidings; 24-hr samples were used for special sampling 

purposes (i.e., follow-up samples, primarily to extend the sensitivity). The MDAs listed are those for 

the primary samples; the tables include the recommended periods for the MDA values. Table 5-14 

lists the current in vivo MDAs along with values from historical documents and the recommended 

periods for use with the MDA values.

5.6 DEVELOPMENT AND IMPACT OF WHOLE-BODY COUNTING

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

experience at the NRTS with in vivo and in vitro internal dosimetry analytical techniques was that a 

large proportion of the internal exposures to NRTS workers was to insoluble materials. Radionuclides 

(e.g. 125Sb, 110mAg, 65Zn, and 95Zr/Nb) 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-hr specimen (Horan 1962; Sill, 

Anderson, and Percival 1964). WBC was demonstrated to detect activity as low as 0.01 μCi in a 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 25 of 60

Table 5-12. MDAs for urine samples 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 93 26,000 Ebersole and Flygare 1957b

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

1965–1971 75 205 3,800 Data Sheet

H-3 1972–1994 3 0.5 dpm/mL 700 AEC 1972b , AEC 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 1.7 32 AEC 1972b, AEC 1974b

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 0.1 0.14 AEC 1974b

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

U (FP) 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)d 1985–present 0.2 μg/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 0.93 1.3 Dodd 1964b

1971–1973 1,000 1.03 1.4 AEC 1972b

1974–1979 1,000 0.47 0.66 AEC 1974b

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 0.16 0.22 AEC 1974b

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 0.155 0.22 AEC 1974b

a. Based on 1,400-mL daily volume and typical sample size. May need adjustment for larger or smaller sample.

b. MDA calculated from inferred 2σ uncertainty.

c. Smallest reported value. Not MDA.

d. KPA = kinetic phosphorescence analysis.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 26 of 60

Table 5-13. MDAs for fecal samples 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 0.4 Dodd 1964b

1974–1994 0.02 AEC 1974

1995–present 0.03 Andersen, Perry, and Ruhter 1995; Bechtel BWXT 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 inferred 2σ uncertainty.

10-min count (Horan 1962). This detection level was several orders of magnitude more sensitive than 

the maximum permissible body burdens (MPBBs) for most beta/gamma fission and activation 

products.

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

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

intake, and (3) selected groups of individuals that were scheduled for semiroutine analyses by health 

physics supervisors (Sommers 1961). In 1963 approximately 1,650 counts were performed; only 

those activities greater than 0.1 μCi were further quantified. This level was determined to be less than 

one-tenth of the MPBB for most of the gamma-emitting isotopes.

The 1963 annual summary report (Dodd 1964) describes the year’s follow-up analyses for the WBC 

program. As shown in Table 5-15, many of the individuals were counted multiple times. The 

maximum activity detected provides an upper bound on how large an activity could be found in 

someone in earlier years before the whole-body counter was operational. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 27 of 60

Table 5-14. In vivo MDAs by period. 

Radiation/

radionuclide Period

In vivo 

MDA (nCi)

Count 

time 

(min) Reference

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

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 1962–2000 12 10 Percival and Anderson 1962a

2001–present 2.5 5 INEEL 2001

2001–present 1.1 10 INEEL 2001

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

1971–1988 5 10 AEC 1972a; AEC 1974

1989 7 10 Martin 1989

1990–1992 2 (lung) Rich 1990

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

2001–present 2.5 5 INEEL 2001

2001–present 1.1 10 INEEL 2001

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

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

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

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

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

1989–2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

2001–present 2.6 5 INEEL 2001

Ru-106 2001–present 27 5 INEEL 2001

2001–present 7.6 10 INEEL 2001

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

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

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

1990–1992 2 (thyroid) 10 Rich 1990

1993–2000 0.3 (thyroid) 10 Grothaus 1993

2001–present 3.8 5 INEEL 2001

2001–present 0.13 (thyroid) INEEL 2001

Cs-134 1989–2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

1990–present 2(lung) Rich 1990

2001–present 3 5 INEEL 2001

2001–present 0.96 10 INEEL 2001

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

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

Perry, and Ruhter 1995

1999–2000 2. (lung) 10 Rich 1990

2001–present 3.1 5 INEEL 2001

2001–present 1.9 10 INEEL 2001

Ba/La-140 1962–present 12 5 INEEL 2001

Ce-141 1962–present 9.9 5 INEEL 2001

2001–present 3.2 10 INEEL 2001

2001–present 0.11 (lung) 60 INEEL 2001

Ce-144 1962–2000 50 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

2001–present 44 5 INEEL 2001

2001–present 15 10 INEEL 2001

2001–present 0.44 (lung) 60 INEEL 2001

Eu-152 1962–present 4 10 INEEL 2001

2001–present 0.18 (lung) 60 INEEL 2001
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 28 of 60

Radiation/

radionuclide Period

In vivo 

MDA (nCi)

Count 

time 

(min) Reference

Eu-154 1962–present 2 10 INEEL 2001

Eu-155 1962–present 1 10 INEEL 2001

Ga-153 1962–present 6.5 10 INEEL 2001

2001–present 0.096 (lung) 60 INEEL 2001

Hf-181 1962–present 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

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

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

Th-230 1974–present 1,000 AEC 1974

Th-234 2001–present 1.4 (lung) 60 INEEL 2001

Np-237 1974–present AEC 1974

U-235 1993 0.2 (wound) 20 Grothaus 1993

1962–present 0.2 (lung) Rich 1990

2001 0.11 (lung) 60 INEEL 2001

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

Pu-238 1989–1998 26 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

1993 1 (wound) 20 Grothaus 1993

1999–2000 30 (lung) Rich 1990

2001 54 (lung) 60 INEEL 2001

Pu-239/240 1971–1993 30 100 AEC 1972a

1974–1988 74 (lung) 100 AEC 1974

1989–1988 80 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

1993 2 (wound) 20 Grothaus 1993

1990–2000 30 (lung) Rich 1990

2001–present 140 (lung) 60 INEEL 2001

Am-241 1989–1999 0.6 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 1995

1993 0.1 (wound) 20 Grothaus 1993

1990–2000 0.2 (lung) Rich 1990

2001–present 0.14 (lung) 60 INEEL 2001

a. MDA calculated from inferred 2σ uncertainty.

5.7 SPECIFIC FACILITIES

Each of the facilities was responsible for its internal dosimetry monitoring program, which was 

designed 1) to prevent and mitigate internal exposure and 2) to evaluate and document internal dose 

above the detectable limits [36]. The following descriptions provide insight and default instructions for 

dose reconstruction if the specific operational facility is known.

5.7.1

General Electric built the first facilities at TAN in 1952 for the ANP Program, which was active during 

the 1950s and early 1960s before ANP was determined to be impractical [37]. The initial mission was 

to develop reactors for aircraft propulsion. The large facilities for this program have been used for a 

number of subsequent INL projects. Approximately 25 different reactor concepts and experiments 

have been conducted at this location, which features large hot cells, maintenance shops, water 

storage pools, and waste management areas (Stacy 2000). The IETs in the 1950s were conducted 

under area controls and radiological monitoring surveillance. The reactors operated in the open, and 

each test involved the release of large quantities of short-lived radioactive fission product gasses and 

volatiles. However, workers were protected by enclosures (e.g., control point buildings) and constant 

monitoring for identification of unanticipated exposures [38].

Test Area North
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 29 of 60

After termination of the IET programs, the TAN facilities were used to handle, inspect, store, and 

prepare for disposal the materials from unplanned reactor excursions. Major reactor components of 

the damaged SL-1 reactor were examined in the large hot cell and prepared for ultimate disposal. 

Table 5-15. 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-103-106b 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-141-144 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.

Fuel from the damaged Three Mile Island (TMI) reactor was brought to and stored in the large water 

pool facility, where it was examined and prepared for permanent dry storage. 

The operation of the hot cells, storage basins, and waste treatment facilities involved aged MFPs 

(primarily 137Cs and 90Sr/Y) with periods when 95Zr/Nb and 144Ce were present. Activation products 

were also encountered, primarily 60Co. Alpha emitters (uranium isotopes and TRU radionuclides) 

were present, but their ratios were at least 50:1 beta:alpha.

Operation of the 15 experimental reactor facilities in this area of the site resulted in short-lived fission 

products in addition to longer-lived MFPs and some MAPs. A few reactor experiments involved 

operating the reactor to destruction, with the attendant breach of containment and potential internal 

exposures.

5.7.1.1 Specific Manufacturing Capability Project

The SMC Project began late in the site history (1985) at the ANP Program site in the large hangar 

facility. The program uses DU to produce armor packages for the U.S. Army M1-A1 and M1-A2 tanks 

(Stacy 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). 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 30 of 60

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 radionuclides of concern at SMC are the isotopes of DU listed in Table 5-16. The mass 

percentages, relative activities in picocuries per microgram, and the total picocuries per microgram 

are based on Integrated Modules for Bioassay Analysis (IMBA) default values.

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 yr 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 

Table 5-16. Mass and activity ratios of SMC DU isotopes [39].a 

Isotope Mass % Activity %

Relative activity 

(pCi/μg)

Total 

(pCi/μg)

U-238 99.8 83.42 0.3354 0.402

U-236 0.0031 0.05 0.0002

U-235 0.20 1.07 0.004

U-234 0.001 15.46 0.062

a. In addition to the uranium isotopes, DU contains two beta-emitting 

radionuclides. Due to short half-lives, Th-234 and Pa-234m reach equilibrium 

with the U-238 parent within about 6 mo of billet casting. Beta monitoring is 

an important part of radiological survey and measurements, even though 

these radionuclides do not contribute significantly to internal doses from 

intakes of DU.

S 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 [40].

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).

Chemical Toxicity: The threshold limit value (TLV) for inhalation of airborne concentrations of 

uranium and its compounds (independent of isotopic composition), as reported by the American 

Conference of Governmental Industrial Hygienists, is 0.2 mg/m3

. The TLV is the time-weighted 

average concentration for a normal 8-hr workday and a 40-hr workweek, to which nearly all workers 

can be repeatedly exposed, day after day, without adverse effect. Conversion of the mass 

concentration using the specific activity of DU of 3.81 × 10-7 mCi/mg results in a radioactivity 

concentration of 7.6 × 10-11 μCi/cm3

. The DAC for Type M uranium is 3 × 10-10 μCi/cm3, which is a 

factor of 4 larger than the chemical TLV. The SMC Project staff has always been aware of the need 

to consider the chemical toxicity of SMC Project DU exposure in addition to the radiological limit [41].

Natural Background Uranium Excretion: Urine samples submitted by SMC Project nonradiation 

workers in 1987, 1994, and 1998 were assumed to represent the nonoccupational excretion levels of 

the SMC Project worker population. The results ranged from 0.04 to 0.33 μg/L with wide fluctuations 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 31 of 60

in individual measurement, some as high as 1.0 μg/L (King 2001). The average reported uranium 

concentration was 0.157 ±0.109 μg/L at 1σ uncertainty. Therefore, 0.16 μg/L is used as the 

nonoccupational component of uranium excretion for SMC Project workers, and is subtracted from 

each urine result before assessment of occupational internal dose. The bioassay results in the worker 

files reflect the subtraction of 0.16 μg/L from the value determined in the laboratory bioassay result 

(King 2001).

5.7.2

INTEC, formerly the ICPP, consists of a complex of highly enriched spent fuel storage basins, fuel 

dissolution and uranium extraction processing facilities, a high-level liquid waste storage tank farm, 

high-level waste calcining processes, and associated analytical and support capabilities. INTEC was 

a process facility for the recovery of highly enriched uranium from spent fuels from a variety of 

national and a few foreign reactors. Because highly enriched uranium was the product, the process 

vessels had to be small for criticality control. Rather than being a plant with large canyon construction 

and complete remote control and maintenance, INTEC 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 [42]. 

Idaho Nuclear Technology and Engineering Center

Most internal doses experienced at INTEC were from accidental releases. Table 5-17 lists unusual 

and episodic events that have occurred at INTEC.

Table 5-17. Notable airborne incidents at INTEC. 

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 at INTEC I-131 Radioactive iodine spread through makeup 

area to operating corridor. Thyroid intake to 

several health physics technicians and 

operators in the 40-μCi range.

Rich 1958; 

Hayden 1958

10/16/59 Criticality accident - in 

shielded process 

system 

Short-lived 

noble gases 

and I-131, -132, 

-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 & I131, -132, -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 INTEC main 

stack

Ru-106 No internal doses detected. ERDA 1977

11/17/72 INTEC mass 

spectrometry Pu 

contamination incident

Pu-238, -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 

Henry and 

Slagle 1985
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 32 of 60

the radiological control program.

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

5.7.2.1 Highly Enriched Spent Fuel Storage

The original spent fuel storage facility (CPP-603) was a 1.5-million gal, three-basin, 20-ft-deep unlined 

concrete water pit that operated from 1950 until 1984. Because the basin was unlined, the use of 

demineralized water was not feasible due to corrosion of the bare concrete. In the late 1970s, ionexchange columns and a sand filter were installed for water cleanup.

The basin was used for the storage of aluminum-clad, stainless-steel-clad, zirconium-clad, and 

sodium-bonded stainless-steel fuels. During 34 yr of operation, cases of cladding leaks resulted in 

MFP contamination of the basin water. Breaks in the stainless-steel cladding of sodium-bonded 

elements allowed a sodium–water reaction that increased the dispersal of fission products to the pool 

water. Pool water contamination up to 0.15 μCi/mL (Rich et al. 1974) existed for periods during which 

cleanup methods were developed. As a result, during the late 1950s through the 1970s, air activity 

above the pool area was routinely measured at 10% to 25% (Rich et al. 1974) of the radioactive 

concentration guide for airborne activity (RCGa) [1 × 10-9 μCi/mL for soluble 90Sr (AEC 1968)] and was 

one of the very few operations at INL in which operators were allowed to work in fractional MPCa 

levels without respiratory protection for several hours a day. Routine bioassay sampling was 

increased for those personnel. The primary contaminants were aged MFPs, primarily 90Sr/90Y and 

137Cs. For workers with established work histories in the Building 603 storage facility for extended 

periods, and if specific bioassay analyses are either not available or insufficient, a default intake of 

250 RCGa-hr/yr should be assumed (i.e., 1,000 hr/yr at 0.25 RCGa), which is favorable to the claimant 

[43]: 

20,000 cm3

/min × 60 min/hr × 1,000 hr/yr × 0.25 × 1 × 10-9 μCi/cm3 × 1 × 106 pCi/μCi 

= 3 × 105 pCi/yr.

Based on the activity mix for stainless-steel fuel after 3 yr of decay provided in Table 5-18, an annual 

intake of 54,000 pCi of 147Pm, 48,000 pCi of 144Ce, 30,000 pCi of 137Cs, 26,000 pCi each of 90Sr and 

90Y, 9,300 pCi of 106Ru, 90 pCi of 239Pu, and 33 pCi of 234U can be assumed to account for over 95% 

of the dose. The remaining ~56,000 pCi are spread over many nuclides and cause less than 5% of 

the dose [44].

Table 5-18. Radiologically significant radionuclides for INTEC-processed fuelsa

 [45].

Nuclide Half-life

Absorption

type

Aluminum-clad fuel

(decayed 1 yr)

Stainless steel-clad fuel 

(decayed 3 yr)

Zirconium-clad fuel

(decayed 5 yr)

Relative 

activity

% inhalation 

dose

Relative 

activity

% inhalation 

dose

Relative 

activity

% inhalation 

dose

Sr-90 28.78 yr F 2.4E-02 13.1 8.6E-2 14.1 2.0E-1 6.3

Y-91 58.51 d S 2.6E-02 1.1 9.7E-6 0.0 8.0E-10 0.0

Zr-95 63.98 d S 4.0E-02 1.2 3.3E-5 0.0 5.6E-9 0.0

Ru-106 368.2 d S 2.3E-02 5.5 3.1E-2 5.7 4.5E-3 0.2

Cs-137 30.07 yr F 2.5E-02 1.2 9.3E-2 3.2 2.1E-1 1.4

Ce-144 284.3 d M 3.0E-01 47.4 1.6E-1 18.4 2.2E-2 0.5

Pm-147 2.623 yr M 5.6E-02 1.3 1.8E-1 3.2 8.4E-2 0.3

U-234 245,500 yr S 2.9E-09 0.0 1.1E-4 4.0 7.2E-8 0.0

Pu-238 87.71 yr M 1.3E-04 26.1 7.6E-6 1.2 3.0E-3 90.0

Pu-239 24,110 yr M 4.1E-07 0.1 3.0E-4 50.0 1.6E-6 0.1

Totalb 97.0 99.0 98.4

Mass fraction U-234 to total U 3.4E-5 5.3E-5 1.4E-4
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 33 of 60

a. Assume exposure to aluminum-clad fuel activity from the beginning through 1970, zirconium-clad fuel activity from 1971 to present, and 

stainless-steel-clad fuel activity when there is indication that the Pu-239 activity exceeds the Pu-238 activity.

b. The total percent of the inhalation dose is less than 100% because other radionuclides not included in the table contribute small 

amounts of dose. 

In 1984, the Fluorinel Dissolution Process and Fuel Storage Facility was completed. It features a 

large stainless-steel-lined pool with ion-exchange cleanup systems and other features for improved 

contamination control and the reduction of chronic internal exposure potential.

5.7.2.2 High-Level Wastes

The high-level waste storage tank farm consists of a series of 500,000-gal underground stainlesssteel tanks, each of which is in a reinforced concrete bunker with alarmed sumps. Several major 

underground spills occurred in the tank farm, primarily at stainless-steel to mild-steel connections. 

These spills resulted in extremely high levels of contaminated soil that were removed with remote 

equipment. These operations had high release potential, and were planned and executed with 

personal protective equipment and monitoring. The contaminated soil was kept wet when being 

handled to minimize airborne contamination. Both the in vitro and in vivo data are documented in the 

personnel dosimetry files [46].

5.7.2.3 High-Level Waste Calcination

High-level waste calcination operations, which began in 1963, consisted of the production of a hightemperature calcine by spraying high-level liquid waste into a fluidized bed of calcine. The nitric 

oxides were vented up the stack following high-efficiency filtering, which left the radionuclides high-fire 

coated on calcine granules. The potential for release of high-level MFP, MAP, and TRU particulate 

activity was significant, but it was recognized and monitored with an extensive array of air and 

radiation (or remote) area monitors (RAMs) [47]. Facility operators were placed on routine as well as 

special bioassay schedules [48].

5.7.2.4 Process Analytical Facilities

The Remote Analytical Facilities and lower-level Process Sample Analytical Laboratories analyzed 

samples critical to process controls. The potential for radioactive material release and internal 

exposure was significant in these facilities. Some internal exposure incidents occurred in the 

laboratories from loss of control of process samples in hoods and other causes. These incidents 

frequently involved loss of control of TRU materials during analytical procedures in hoods or benchtop 

confinement-type operations [49].

5.7.2.5 Spent Fuel Processing and INTEC Most-Limiting Radionuclides

Highly enriched fuel processing, which began in 1953, consisted of nitric acid dissolution of aluminumclad elements, electrolytic nitric acid dissolution of stainless-steel-clad elements, hydrofluoric acid 

dissolution of zirconium-clad elements, and nitric acid dissolution of graphite fuel following graphite 

burning. Uranium was extracted from the dissolved elements (using variations of the plutonium–

uranium extraction process, which produced a uranyl nitrate product). The first-, second-, and thirdcycle raffinates contained high levels of MAPs and MFPs along with TRU radionuclides. These 

products, in highly corrosive matrices, were difficult to contain in the confinement barriers of the 

process piping, process off-gas, etc., and they resulted in routine leaks and spills. Most of the leaks 

were confined to process cells, but they occasionally occurred in occupied process control and 

equipment areas [50]. For this reason, the extensive network of CAMs and RAMs was essential for 

timely detection of loss of confinement. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 34 of 60

Source terms for the various INTEC operations are based on the types of fuel that were reprocessed. 

All of the reprocessed fuels were highly enriched with 235U enrichments ranging from 50% to 93%. 

Most of the INTEC reprocessing involved aluminum-clad fuels from test reactors, stainless-steel clad 

fuels from the Experimental Breeder Reactor II (EBR-II), and zirconium-clad fuels from various 

reactors. There were other minor campaigns for processing minor amounts of fuel such as graphite 

fuel from the ROVER (space nuclear propulsion) program. In addition, some fuels that had no burnup 

or very little burnup were reprocessed in a hands-on operation known as Custom Processing. 

To minimize the radiological safety hazard of the relatively volatile halogens, the fuel was normally 

decayed a minimum of 120 d before shipment to INTEC for processing. Processing of fuels often did 

not occur until years later. 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 [51]. To give an indication of the radionuclide inventory in 

fuels that were processed at INTEC, the ORIGEN2 software (Croff 1980) was used to determine the 

decay of previously calculated fuel inventories (Wenzel 2000) to typical decay times before 

processing. Aluminum fuel based on fuel from the Advanced Test Reactor (ATR) was decayed 1 yr. 

Stainless-steel fuel based on fuel from EBR-II was decayed 3 yr. Zirconium fuel based on various 

reactors was decayed 5 yr. These decay times represented the minimum decay times that occurred 

before processing of the fuels. Table 5-18 lists the radiologically significant nuclides and the percent 

of the inhalation dose from inhaling activity from these three types of fuel. Several short-lived progeny 

(

90Y, 106Rh, 137mBa, and 144Pr) have over one-third of the activity, which will be followed in IMBA 

calculations and are not shown. The inhalation dose percentages were calculated using ICRP 

Publication 68 dose conversion factors for AMADs of 5 μm (ICRP 1995). 

Aluminum fuels were processed between 1953 and 1986, stainless-steel fuels primarily between 1977 

and 1981, and zirconium fuels between 1972 and 1988 (Staiger 2003). There are relatively long halflife fission products that persist for INTEC source terms. In most cases, the source terms were well 

tagged with beta-emitting radionuclides, which allowed beta/gamma-detecting CAMs to be used at 

INTEC with the realization that they would also warn of possible alpha contamination or internal 

exposures [52]. 

Table 5-18 contains too many radionuclides for efficient dose reconstruction. Rather than include all 

of the radionuclides in the default summary table for missed dose (Table 5-24 later in this document), 

only 90Sr, 137Cs, 144Ce, and 238Pu are included for aluminum and zirconium fuels. For stainless-steel 

fuels, the 238Pu is replaced by 239Pu [53]. Cesium-137 was selected because it is most commonly 

reported in the in vivo results rather than for its dose contribution. For the pre-1960 period, 91Y and 

95Zr are included, but after that the fuel being processed had decayed sufficiently so that these 58- 

and 64-d half-life radionuclides would have decayed away.

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 Materials Test Reactor (MTR) with as 

little decay as manageable (less than 2 d). Fuel was removed from the MTR, transported about 2 mi 

to INTEC in a heavily shielded transport container by a straddle carrier, immediately dissolved, and 

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

National Laboratory. This process released large quantities of volatile radioactive iodines, which have 

a higher potential for escaping confinement systems. Several significant internal exposure incidents 

occurred in which 131I, 132I, and 133I thyroid intakes occurred before personnel could respond to CAM 

alarms and take protective or corrective actions [54].
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 35 of 60

5.7.3

ANL-W, based at INL since 1951, continues to conduct nuclear programs. Although ANL-W receives 

internal dosimetry support from the INL service laboratories, its radiological safety programs operate 

under the DOE COO. However, over the 52 yr of ANL-W experience, informal program coordination 

has occurred to foster technical consistency with the other INL facility internal dosimetry programs.

Argonne National Laboratory–West

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

locations, one on the southwest side of the site near RWMC and the others at the current location on 

the southeast side of the site. Early reactor operations included physics critical experiments; power 

production; routine unmoderated operation, uranium-fueled, plutonium-fueled, and breeder reactor 

designs; and self-destruct experiments. As listed in Table 5-19, the radionuclides of concern ranged 

the spectrum of MFP, MAP, uranium, and TRU nuclides. 

Table 5-19. Radionuclides of concern and MDAs for ANL-W locations (Nielsen 1996). 

Radionuclide

Absorption 

types Sources/characteristicsa

MDAs

Urine 

(pCi/mL)

In vivob,

5-min count

(nCi)

In vivoc,

10-min count

(nCi)

H-3 (HTO) F EBR-II Reactor Facility & Sodium 

Components Maintenance Shop

2.0 N/Ad N/A

Mn-54 F, M EBR-II primary source. Levels low 

and decreasing

N/A 2.6 1.3

Fe-59 N/A 4.5 2.4

Co-58 N/A 2.5 1.3

Co-60 N/A 2.5 1.4

Sr/Y-90 F All facilities handling fission products 0.02e N/A N/A

Cs-134 F All facilities handling fission products N/A 2.7 1.3

Cs-137 N/A 3.0 1.5

U-235 F, M, S Hot cell, hoods, glove boxes, waste, 

reactor fuel, research areas

0.02e 0.1 N/A

U-238 0.02e 1.6 N/A

Pu-238 M, S FCF, HFEF, ZPPR, Analytical 

Laboratory

0.02e 60.0 N/A

Pu-239 0.02e 161.0 N/A

Am-241 M FCF, HFEF, ZPPR, Analytical 

Laboratory

N/A 0.1 N/A

a. See the acronyms and abbreviations list.

b. Using a Canberra WBC with a 5-min count. 

c. Using a large NaI WBC detector with a 10-min count.

d. N/A = not applicable.

e. Assumes a 100-mL sample.

5.7.4

The RWMC has supported INL operations as a waste management complex since 1952 and has 

received large quantities of TRU waste from Rocky Flats 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 total-containment practices. The four major areas in the RWMC facility are the Subsurface 

Disposal Area (SDA) for permanent disposal of low-level waste and some early TRU waste (which will 

eventually be exhumed and repackaged); the 58-acre Transuranic Storage Area (TSA) for temporary 

storage, examination, and certification before shipment to the Waste Isolation Pilot Plant; the 

operations area; and an administrative area where no radioactive waste is permitted.

Radioactive Waste Management Complex
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 36 of 60

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 (there were two in 1992 and one in 1996) [55]. Therefore, bioassay is 

conducted randomly at the current time (INEEL 2001).

Tables 5-20 and 5-21 summarize the major radionuclides in the RWMC waste inventory. TRU 

radionuclides are the primary contaminants in the TSA waste; all but 241Pu 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. 

Because the waste has all chemical and physical characteristics by nature, the absorption type most 

favorable to claimants should be assumed unless data indicate that another type fits the data better. 

Exposures in later years are most likely from contaminated soil or corrosion products, and a default 

particle size distribution of 5-μm AMAD should be assumed as recommended by ICRP Publication 68 

(ICRP 1995).

Table 5-20. 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

Table 5-21. 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

Ni-63 3.3E-1 7.4

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

5.7.5

The Waste Reduction Operations Complex 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 INL

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). 

Waste Reduction Operations Complex

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.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 37 of 60

Because the types of radioactive materials processed at WERF varied depending on the area 

shipping the waste to WERF, the assumption favorable to claimants is that radioactive materials came 

from zirconium fuel as processed at INTEC (see Table 5-18, columns 7 and 8).

5.7.6

5.7.6.1 Test Reactor Area Reactors

Test Reactor Area

The MTR was the second operating reactor at INL and ran from March 1952 to 1970 in the Test 

Reactor Area (TRA) (Stacy 2000); 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 TRA complex includes hot cells, a gamma irradiation pool facility, 

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

INL, 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 DOE reactors. 

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

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

aluminum. Sodium-24 has a half-life of 15 hr and emits a high-energy gamma ray (2.75 MeV). The 

inhalation dose to personnel from 24Na is insignificant in comparison to that from the fission products 

in the fuel [56]. There are minor levels of activation products of stainless steel (58Co, 60Co, 51Cr, 56Mn, 

etc.) in the primary coolant system (PCS) due to corrosion of the stainless-steel PCS components 

[57]. To be favorable to claimants, this TBD recommends direct use of any in vivo counting data in 

the worker files. If applicable data are absent, the in vivo MDLs can be assumed consistent with 

information in Tables 5-14 and 5-24 (see Section 5.8). 

Several factors contributed to unusual amounts of fission products in the PCSs 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 [58]. Another factor was tramp fuel, which was a 

contaminant on the outside of the cladding. During reactor operation, fission products in tramp 

uranium were released directly to the PCS [59]. Reactor operators and the fuel manufacturer 

resolved these deficiencies over time; by the time the ATR became operational, the PCS of the ATR 

was considerably less contaminated than that of the MTR or the ETR during their early years of 

operation. The radiologically significant radionuclides for aluminum-clad fuel processed at INTEC in 

Table 5-18 are applicable to this PCS contamination [60].

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

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

personnel from releases of noble gases was direct radiation rather than inhalation [61]. The direct 

radiation caused instruments to alarm, which resulted in immediate evacuation of the affected areas. 

The position favorable to claimants is that there could have been some halogens and particulate 

radionuclides released along with the noble gases. The ORIGEN2 code (Croff 1980) was used to 

calculate a radionuclide inventory for ATR fuel at the time of reactor shutdown. The radionuclide 

inventory was then fractionated according to the release percentages often used to assess 

hypothetical reactor accidents (DiNunno et al. 1962) as 100% noble gases, 50% halogens, and 1% 

solids. A halogen:solid ratio of 50:1 was therefore used to determine the radionuclides that could 

significantly contribute to inhalation dose. Table 5-22 contains the relative amounts of halogens and 

particulate radionuclides from ATR fuel that contribute significantly to inhalation dose, but there are 

too many for efficient dose reconstruction. The iodine isotopes contribute the large majority (90.3%) 

of the dose from a test reactor gaseous release. Therefore, 131I was selected to be the representative 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 38 of 60

radionuclide for missed dose (in Table 5-24), and the dose from 131I was weighted by a factor of 1.6 to 

account for the total iodine dose [62]. When there is measured 131I for any facility, the dose 

reconstructor can choose to apply this factor of 1.6 to the 131I dose to account for short half-life iodines 

and can choose to ratio 144Ce, 89Sr, 91Y, and 95Zr activities to the 131I activity using the values in 

Table 5-22.

Another potential source of inhalation dose was the pressurized-loop experiments in the TRA 

reactors. These experiments sometimes contained silver and tantalum, which became activated and 

produced 110mAg and 182Ta. When a depressurizing incident occurred in the loop, these activation 

products were sometimes released to work areas (Nertney et al. 1967). One of the major airborne 

incidents at TRA listed in Table 5-23 involved the release of 182Ta. The position favorable to claimants 

is to assign a missed dose based on the MDA for in vivo counting (see Table 5-14) for 110mAg and 

182Ta.

Table 5-22. Radiologically significant radionuclides for ATR fuel gaseous 

releases [63]. 

Nuclide Half- life

Absorption 

type

Fraction 

activity

Percent

inhalation 

dose

Ratio activity 

to I-131 

activity

Sr-89 50.5 d S 0.0028 1.0 3.5E-02

Y-91 58.51 d M/S 0.0033 1.3 4.1E-02

Zr-95 63.98 d S 0.0036 0.9 4.4E-02

I-131 8.041 d F 0.0816 55.7 Not applicable

I-132 2.3 h F 0.1298 1.6 Not applicable

I-133 20.8 h F 0.2040 26.6 Not applicable

I-134 0.876 h F 0.2225 1.1 Not applicable

I-135 6.61 h F 0.1855 5.3 Not applicable

Ce-144 284.3 d S 0.0012 1.7 1.5E-02

Total 0.834 95.2 Not applicable

Dose percentage from all iodine isotopes 90.3

Weighting factor for I-131 to account for iodine dose 1.6

Table 5-23. Major airborne incidents at TRA. 

Date Incident

Radionuclide(s) 

released Reference

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

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

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

01/06/1967 Noble gas release at ATR Noble gas Sommers 1977

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 radiologically significant radionuclides listed in Table 5-18 for aluminum 

fuel processed at INTEC are applicable to this TRA contamination.

5.7.6.2 Test Reactor Area Laboratories

The wing buildings of the MTR at TRA house chemistry and other laboratories. Over the years, 

experiments in the laboratories have resulted in contamination and airborne activity incidents. These 

experiments have involved various isotopes of plutonium and uranium along with other radionuclides. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 39 of 60

The radiologically significant radionuclides listed in Table 5-18 for aluminum-clad fuel processed at 

INTEC are applicable for the experiments in the TRA laboratories.

Beginning in 1980, many of the laboratories were involved in various studies of TMI reactor fuel from 

the accident in March 1979. Because the TMI core had only the equivalent of 100 effective fullpower d at the time of the accident, the TRU content would not be as high as that for a typical power 

reactor core, which is routinely operated for a year or so before the fuel is changed. Use of the 

radiologically significant nuclides for zirconium fuel processed at INTEC in Table 5-18 is favorable to 

claimants for evaluating inhalation dose from exposure to contamination from the TMI fuel.

5.8 DEFAULT FOR MISSED DOSE 

Based on the INL characteristics and circumstances, a number of missed dose default assumptions 

have been derived. Table 5-24 is a summary of these recommended defaults. 

Table 5-24. Default table for missed dose.

Period

Based 

on Site area Recommendation Basis

Startup–

1960

Urine

gross β

All Calculate chronic Sr-90 intake that 

results in a urine activity of 

0.4 × gross β.

Typical β activity is 0.33 Sr-90, 

0.33 Y-90 & 0.33 Cs-137. Use of 

0.4 is favorable to claimants.

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

Cs-137 & Sr-90 are approximately 

equal.

Pu-238 intake = 0.005 × Sr-90 

intake

Pu:Sr-90 ratio of 0.005 in Al-clad 

fuel.

Ce-144 intake = 12.4 × Sr-90 

intake

Ce-144:Sr-90 ratio of 12.4 in Alclad fuel. 

Y-91 intake = 1.1 × Sr-90 intake Y-91:Sr-90 ratio of 1.1 in Al-clad 

fuel.

Zr-95 intake = 1.2 × Sr-90 intake Zr-clad-95:Sr-90 ratio of 1.2 in Alclad fuel.

TRAa Annual I-131 intake = 25 μCi Assumes I-131 in thyroid at 3 d is 5 

μCi Vapor Type F, the maximum in 

vivo measured in 1963. 

1961–1970 In vivo

Cs-137

All Calculate chronic Cs-137 intake 

that results in the in vivo

measurement.

Sr-90 intake = Cs-137 intake when 

no in vitro measurement.

Half-lives and fission yields are 

approximately equal.

Pu-238 intake = 0.005 × Cs-137 

intake when no in vitro

measurement.

Pu:Cs-137 ratio of 0.005 in Al-clad 

fuel. 

Ce-144 intake = 12.4 × Cs-137 

intake when no measurement.

Ce-144:Sr-90 ratio of 12.4 in Alclad fuel.

TRA Annual I-131 intake= 60 nCi when 

no measurement. 

Based on MDA for I-131 of 12 nCi 

at 3 d in thyroid Vapor Type F.

One acute intake of Ag-110m = 

71 nCi when no measurement. 

Based on MDA for Ag-110m of 

12 nCi at 3 d type M. Known 

incident involving Ag-110m.

One acute intake of Ta-182 = 

80 nCi when no measurement. 

Based on MDA for Ta-182 of 

12 nCi at 3 d type M. Known 

incident involving Ta-182.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 40 of 60

Period

Based 

on Site area Recommendation Basis

1971–1980 In vivo

Cs-137

All Sr-90 intake = Cs-137 intake when 

no in vitro measurement. 

Half-lives and fission yields of 

Cs-137 & Sr-90 are approximately 

equal.

All but

ANL-W 

Pu-238 intake = 0.02 × Cs-137 

intake when no in vitro

measurement

Pu:Sr-90 ratio of 0.02 in Zr-clad 

fuel. See Table 

5-18, column 9.

Ce-144 intake = 1.05 × Cs-137 

intake when no measurement

Ce-144:Sr-90 ratio of 1.05 in Zrclad fuel.

ANL-W Pu-239 intake = 0.003 × Cs-137 

intake when no in vitro

measurement

Pu:Cs-137 ratio of 0.003 in 

stainless-steel-clad fuel. 

Ce-144 intake = 2 × Cs-137 

assigned intake when no

measurement

Ce-144:Cs-137 ratio of 2 in 

stainless-steel-clad fuel.

TRA Annual I-131 intake = 94 nCi when 

no measurement. 

Based on MDA for I-131 of 12 nCi 

at 3 d thyroid Vapor Type F 

weighted by a factor of 1.6.

1981–

present

Bioassay All except 

ANL-W, 

INTECb

, & 

SMC

Calculate associated radionuclide 

intakes based on Al-clad fuel 

distribution

Distribution calculated for Al-clad 

fuel. See Table 5-18, column 3.

ANL-W Calculate associated radionuclide 

intakes based on stainless-steelclad fuel distribution

Distribution calculated for stainless 

steel fuel. See Table 5-18, column 

4.

INTECb & 

unknownc 

Calculate associated radionuclide 

intakes based on Zr-clad fuel 

distribution

Distribution calculated for Zr-clad 

fuel. See Table 5-18, column 5.

SMC Calculate associated intakes based 

on known isotope distribution

Isotope distribution, see 

Table 5-16.

a. The dose recommendations for specific areas (TRA, INTEC, ANL-W) are to be added to recommendations for “All”.

b. Formerly known as ICPP.

c. When the worker’s area is not known, or if the employee worked in many different areas, use the dose 

recommendations for INTEC.

For most of the history of the INL, personnel dosimeters were issued to all workers, regardless of 

work assignment, who entered the security access control points at each facility. For example, 

administrative and clerical personnel were required to wear dosimeters even though they were not 

exposed to elevated backgrounds or internal dose potential. Dose reconstructors should determine 

the appropriate default categories from Table 5-24 as follows: 

• If the worker’s file includes positive external dosimeter readings, the worker should be treated 

as a radiation worker and the default internal missed dose should be applied as outlined in the 

table. 

• If no detectable external or internal dose information is recorded, only the environmental dose 

should be included. For example, a worker at TRA in 1975 should be assigned the intakes for 

“All,” for “All but ANL-W,” and for “TRA” [64].
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 41 of 60

5.9 UNMONITORED WORKERS

As noted above, INL personnel dosimeters were issued to all workers at facilities that handled 

radioactive material. Many of these workers, due to the nature of their work, would not have been 

exposed to internal activity and would not have been subject to routine bioassay [65]. 

The INL radiation protection program to detect the presence and spread of radioactive contamination 

included: 

• Areas with potential airborne contamination were monitored with alarming CAMs. 

• The spread of contamination was monitored through the use of smears and monitoring 

instrumentation. 

INL policy was to screen radiation workers using urine samples and WBC for internal contamination 

many (often four or more) times a year.

Most of the encountered activity was well tagged with beta/gamma activity, which would have 

produced measurable direct radiation doses on a personnel dosimeter [66]. Therefore, the probability 

that a worker received a significant unmonitored internal intake of radioactive material is very low. It is 

recommended that individuals who were not issued a personnel dosimeter and have no record of 

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

INL often used construction workers. Each construction job was evaluated to determine if radiation 

exposure or internal dose could be received [67]. When construction work occurred in an area with 

potential radiation exposure, including internal dose exposure, construction workers were monitored in 

the same fashion as radiation workers. Construction workers who were issued personnel dosimeters 

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

Construction workers who were not issued a personnel dosimeter should be assigned the 

environmental dose for the facility. 

5.10 ATTRIBUTIONS AND ANNOTATIONS 

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

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

employed at INL 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 document, including all findings and conclusions. Mr. Rohrig 

continues to serve as a Site Expert for this document because he possesses or is aware of 

information relevant 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 the 

Document Owner, those materials are fully attributed to the source. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 42 of 60

Bryce Rich served as one of the initial Subject Experts for this document. Mr. Rich was previously 

employed at INL 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 INL 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 INL 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. 

Henry Peterson served as one of the initial Subject Experts for this document. Mr. Peterson was 

previously employed at INL 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.

This statement by Mr. Rich is based on a review of the referenced AEC annual reports. 

[3] 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 the HSD, allowed 

for a consistency in assumptions about the internal dosimetry program because RESL was a 

constant even though there was a large number of varying facilities and continually changing 

contractors at INL. This is supported by the referenced AEC annual reports. 

[4] Rich, Bryce, and Jenkins, JoAnn M. Consulting Health Physicist and Senior Health Physicist. 

M. H. Chew & Associates and Dade Moeller & Associates. November 2004.

To assign an internal dose from a radionuclide intake, the specific radionuclide must be 

known. In some cases worker bioassay records only provide results for overall beta- and 

alpha-emitting radionuclides and do not identify the specific radionuclides. In these cases, it is 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 43 of 60

necessary to provide guidance to the dose reconstructor on what the radionuclides that are 

represented by the gross beta and alpha analyses are likely to be so that internal doses can 

be assigned. These default assumptions are based on reviews of bioassay and facility data by 

Mr. Rich.

[5] 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 INL incident reports. It is common industry practice to conduct incident investigations in a 

timely manner.

[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 knowledge of the NRF from working at INL. 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.

[8] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement is based on a review of the referenced AEC annual reports and employee 

bioassay and dosimetry data.

[9] 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). 

[10] 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). 

[11] 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). 

[12] 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). 

[13] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of the INL 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). 

[14] 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). 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 44 of 60

[15] 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 INL internal dosimetry program and is supported by Puphal (1994).

[16] 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 INL 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. 

[17] 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 

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 INL internal dosimetry program and is supported by the 

referenced internal dosimetry TBDs. 

[18] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This conclusion is based on the fact that confirmatory analyses that identified specific 

radionuclides were generally performed on bioassay samples whose initial results were in 

excess of 2σ. It can be surmised that if additional analyses were not performed on a sample, 

then the initial gross alpha and gross beta results were within 2σ. This conclusion is 

supported by facility bioassay data and Puphal (1994). 

[19] 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).

[20] JoAnn M. Jenkins, Senior Health Physicist, Dade Moeller & Associates. March 2007.

Federal regulations provide reporting requirements for internal dose. Internal dose is typically 

calculated based on bioassay monitoring results. Changes in federal regulations about the 

threshold at which dose is required to be reported and the annual dose limits affect the level at 

which calculated dose is reported; they do not generally affect the bioassay methodology and 

analytical programs for collection of the raw data for dose calculations. 

[21] 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.

[22] JoAnn M. Jenkins, Senior Health Physicist, Dade Moeller & Associates. March 2007.

Federal regulations provide reporting requirements for internal dose. Internal dose is typically 

calculated based on bioassay monitoring results. Changes in federal regulations about the 

threshold at which dose is required to be reported and the annual dose limits affect the level at 

which calculated dose is reported, they do not generally affect the bioassay methodology and 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 45 of 60

analytical programs for collection the raw data for dose calculations or the facility air 

monitoring programs. 

[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 

the referenced AEC annual reports and internal dosimetry TBDs.

[24] JoAnn M. Jenkins, Senior Health Physicist, Dade Moeller & Associates. March 2007.

The DOE set guidelines for positive bioassay results that require follow-up and reporting. 

These guidelines changed over time, but the changes in guidelines did not affect bioassay 

sampling frequency and the recording of results because the changes affected reporting 

requirements and not the collection of raw data for calculation of internal doses. The number 

of bioassay samples that required follow-up analyses varied over time as a result of the 

changing regulatory guidance about the need for follow-up sampling. 

[25] 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 INL 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. 

[26] 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 INL internal dosimetry program and is supported by the referenced 

AEC annual reports and internal dosimetry TBDs. 

[27] 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.

[28] 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. 

[29] 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. 

[30] 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. 

[31] 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. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 46 of 60

[32] 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 INL bioassay 

records.

[33] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on information from a review of the referenced AEC 

annual reports.

[34] 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 INL bioassay 

records.

[35] 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.

[36] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of the referenced internal dosimetry TBDs. 

[37] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on his knowledge of the INL and work experience at the 

facility and is supported by Stacy (2000). 

[38] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on information in Stacy (2000). 

[39] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

The information in the table is from King (2001).

[40] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement is based on a review of INEEL (2001). 

[41] 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. 

[42] 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 INL internal dosimetry program. Although the maintenance 

operations that required entry into the process cells were planned extensively before execution 

and the cells were decontaminated, the potential for internal exposures was still high because 

not all of the contaminants could be cleaned out of the cells.

[43] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

The assumption of 1,000 hr/yr working in the pool area is based on the fact that individuals 

worked in various areas throughout the facility and not at a single location. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 47 of 60

[44] Rohrig, Norman. Consulting Health Physicist. Intrepid Technology and Resources. 

November 2004. 

These numbers are the product of the relative activity and the total activity (3 × 105 pCi/yr) to 

obtain the activity of each radionuclide.

[45] Wenzel, Douglas, and Rohrig, Norman. Principal Health Physicist and Consulting Health 

Physicist. Intrepid Technology and Resources. November 2004.

This table was compiled by Mr. Wenzel from his knowledge of the fuel history he obtained 

from his work experience at the INL and of ORIGEN calculations. Percent inhalation dose 

calculations were performed by Mr. Rohrig using the ORIGEN output and ICRP Publication 68 

dose conversion factors (ICRP 1995). 

[46] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on a review of INL bioassay records and the previously 

mentioned bioassay database.

[47] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on information from a review of the referenced internal 

dosimetry TBDs.

[48] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on information from a review of the referenced internal 

dosimetry TBDs.

[49] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on information from a review of the referenced internal 

dosimetry TBDs.

[50] Levitt, Boyd. Senior Health Physicist. Intrepid Technology and Resources. November 2004.

An example is a leak in the process control corridor of Building 601 that resulted in radiation 

levels of over 100 rad/hr.

[51] 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. 

[52] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

The fuel at INTEC 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. 

[53] Wenzel, Douglas. Principal Health Physicist. Intrepid Technology and Resources. November 

2004.

Table 5-18 identifies the activity of both plutonium isotopes. The importance of the different 

plutonium isotopes varies with the type of fuel. These radionuclides were selected by Mr. 

Wenzel from information from his work experience at the INL and from an extensive review of 

the referenced AEC annual reports and internal dosimetry TBDs. To simplify missed dose 

calculations, the following four representative radionuclides were selected for the stated 

reasons: 90Sr based on the availability of radioanalytical techniques, 137Cs based on 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 48 of 60

confidence that it was easily detectable, 144Ce due to its large dose contribution, and 238Pu to 

represent the TRU elements. In the case of stainless-steel fuel, 239Pu should be substituted 

for 238Pu because it is more prevalent. Together these radionuclides make up 87.8% of the 

committed effective inhalation dose. Percent committed effective inhalation dose values were 

taken from Table 5-18. This table was compiled by Mr. Wenzel from his knowledge of both the 

fuel history he obtained from his work experience at the INL and ORIGEN calculations. 

Percent committed effective inhalation dose calculations were performed by Mr. Rohrig using 

the ORIGEN output and ICRP Publication 68 dose conversion factors (ICRP 1995).

[54] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement by Mr. Rich is based on Hayden (1958). 

[55] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

See Table 5-1.

[56] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement is based on the fact that the 24Na is not liberated from the aluminum cladding 

and that 24Na has a short half-life.

[57] 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. 

[58] 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. 

[59] 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. 

[60] 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. 

[61] 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. 

[62] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

The dose from 131I makes up 55.7% of the total dose. It is the largest contributor to dose in the 

iodine family. The total committed effective dose contribution due to iodines is 90.3 percent. 

Applying a factor of 1.6 to the 131I dose accounts for the dose from the additional iodine 

isotopes. Doses were taken from Table 5-22. This table was compiled by Mr. Wenzel from 

his knowledge of the fuel history from his work experience at INL and of ORIGEN calculations. 

Ratio of activity to 131I activity was performed by Mr. Rohrig using the ORIGEN output.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 49 of 60

[63] Wenzel, Douglas, and Rohrig, Norman. Principal Health Physicist and Consulting Health 

Physicist. Intrepid Technology and Resources. November 2004.

This table was compiled by Mr. Wenzel from his knowledge of the fuel history from his work 

experience at INL and of ORIGEN calculations. Ratio of activity to 131I activity was performed 

by Mr. Rohrig using the ORIGEN output. 

[64] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement is an example of the use of Table 5-24.

[65] 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 INL internal dosimetry program.

[66] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement is based on a review of the referenced internal dosimetry TBDs and AEC 

annual reports.

[67] Rich, Bryce. Consulting Health Physicist. M. H. Chew & Associates. November 2004.

This statement is based on a review of the referenced internal dosimetry TBDs.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 50 of 60

REFERENCES 

ACC (American Cyanamid Company), 1952, CPP Health Physics Manual, IDO 14055, Idaho Falls, 

Idaho, April. [SRDB Ref ID: 8095]

AEC (U.S. Atomic Energy Commission), 1958, "Permissible Levels of Radiation Exposure," Chapter 

0524, AEC Manual, 0524-01, Washington D.C., February 1. [SRDB Ref ID: 10678]

AEC (U.S. Atomic Energy Commission), 1960, Annual Report of Health and Safety Division, 1959, 

Idaho Operations Office, IDO-12014, Idaho Operations Office, Idaho Falls, Idaho, October. 

[SRDB Ref ID: 8080]

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

IDO-12019, Idaho Operations Office, Idaho Falls, Idaho, September. [SRDB Ref ID: 1230]

AEC (U.S. Atomic Energy Commission), 1963, "Standards for Radiation Protection," Chapter 0524, 

AEC Manual, Washington D.C. August 12. [SRDB Ref ID: 13037]

AEC (U.S. Atomic Energy Commission), 1968, "Chapter 0524, Standards for Radiation Protection," 

AEC Manual, 0524-01, Washington, D.C., November 8. [SRDB Ref ID: 10668]

AEC (U.S. Atomic Energy Commission), 1972, 1971 Annual Report of the Health Services Laboratory, 

IDO-12076, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1431]

AEC (U.S. Atomic Energy Commission), 1974, Detection Limits, Idaho Operations Office, Idaho Falls, 

Idaho, December. [SRDB Ref ID: 13448]

AEC (U.S. Atomic Energy Commission), 1975, "Occupational Radiation Exposure Information," 

Chapter 0525, AEC Manual, 0525-01, Washington, D.C., January 7. [SRDB Ref ID: 11594]

Andersen, B. L., O. R. Perry, and P. E. Ruhter, 1995, Technical Basis for Internal Dosimetry at 

LITCO, EDF# RDRU-95-3.0, EG&G Idaho, Idaho National Engineering Laboratory, Idaho 

Falls, Idaho, January 31, 1995. [SRDB Ref ID: 13453]

Aoki, I., 1979, "Reporting of Dose or Dose Commitment from Internal Sources," memorandum to 

Dosimetry File, July 9. [SRDB Ref ID: 13455] 

Author unknown, 1981, "Appendix I," ICPP Internal Dose Assessment Manual, Idaho Falls, Idaho, 

March. [SRDB Ref ID: 1223]

BBI (Bechtel BWXT Idaho), 2000, Quality Assurance Project Plan for The Analytical Laboratories 

Department Radioanalytical Section, PLN-153, Rev. 1, Idaho National Engineering and 

Environmental Laboratory, Idaho Falls, Idaho, August 17. [SRDB Ref ID: 8540]

Bhatt, A. R., 2002, Bioassay Sample Analysis Technical Basis Document, EDF-In-Vitro-1, Rev. 0, 

Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho, August 21. 

[SRDB Ref ID: 7981]

Bhatt, A. R., 2003, In Vitro Analysis Measurement Techniques, INEEL/INT-03-00702 Rev. 0, Bechtel 

BWXT Idaho, Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho, 

September. [SRDB Ref ID: 8414]
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 51 of 60

Bodnar, L. Z., and D. R. Percival, editors, 1982, RESL Analytical Chemistry Branch Procedures 

Manual, IDO-12096, U.S. Department of Energy, Idaho Operations Office, Idaho Falls, Idaho. 

[SRDB Ref ID: 16103]

Casto, W. R., 1980, “ICPP Criticality Event of October 17, 1978,” Nuclear Safety, volume 21, number 

5, pp. 648–653. [SRDB Ref ID: 1218]

Croff, A. G., 1980, ORIGEN2—A Revised and Updated Version of the Oak Ridge Isotope Generation 

and Depletion Code, ORNL-5621, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 

July. [SRDB Ref ID: 23569]

Dickson, R. L., 1977, "Reporting Levels for Whole-Body Counting," memorandum to L. Z. Bodnar and 

D. G. Olson, U.S. Energy Research and Development Administration, Idaho Operations Office, 

Idaho Falls, Idaho, July 15.

DiNunno, J. J., F. D. Anderson, R. E. Baker, and R. L. Waterfield, 1962, Calculation of Distance 

Factors for Power and Test Reactor Sites, TID-14844, U.S. Atomic Energy Commission, 

March, 1962.

Dodd, A. O., editor, 1963, Health & Safety Division Annual Progress Report 1962, IDO-12033, U.S. 

Atomic Energy Commission, Idaho Operations Office, National Reactor Testing Station, Idaho 

Falls, Idaho. [SRDB Ref ID: 8075]

Dodd, A. O., editor, 1964, Health & Safety Division Annual Progress Report 1963, IDO-12037, U.S. 

Atomic Energy Commission, Idaho Operations Office, National Reactor Testing Station, Idaho 

Falls, Idaho. [SRDB Ref ID: 13449]

DOE (U.S. Department of Energy), 1988, Radiation Protection for Occupational Workers, Order 

5480.11, Office of Environment, Safety and Health, Washington D.C., December 21. [SRDB 

Ref ID: 8073]

DOE (U.S. Department of Energy), 1991, Idaho National Engineering Laboratory Historical Dose 

Evaluation, DOE/ID-12119, Volumes 1 and 2, Idaho Operations Office, Idaho National 

Engineering Laboratory, Idaho Falls, Idaho, August. [SRDB Ref ID: 1107, 1109]

DOE (U.S. Department of Energy), 1994, U.S. Department of Energy Radiological Control Manual, 

DOE/EH-0256T, Rev. 01, Office of Assistant Secretary for Environment, Safety and Health, 

Washington, D.C., April. [SRDB Ref ID: 14856]

DOE (U.S. Department of Energy), 1997, Linking Legacies: Connecting the Cold War Nuclear 

Weapons Production Processes to Their Environmental Consequences, DOE/EM-0319, Office 

of Environmental Management, Washington, D.C., January. [SRDB Ref ID: 11930]

Ebersole, E. R., 1956, Analysis of Stack and Field Samples during IET Operations, June 1956, Idaho 

Operations Office, IDO-12009, U.S. Atomic Energy Commission, Idaho Operations Office, 

Idaho Falls, Idaho, July. [SRDB Ref ID: 8393]

Ebersole, E. R., and J. K. Flygare Jr., 1957, "Determination of Gross Beta Activity in Urine with K-40 

Correction," Proceedings of the Health Physics Society First Annual Meeting, June 25-27, 

1957, University of Michigan, Ann Arbor, Michigan, pp. 72–78. [SRDB Ref ID: 8388]
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 52 of 60

ERDA (U.S. Energy Research and Development Administration), 1975, "Standards for Radiation 

Protection," Chapter 0524, ERDA Manual, 0524-01, Washington, D.C., April 8. [SRDB Ref ID: 

8083]

ERDA (U.S. Energy Research and Development Administration), 1977, Final Environmental Impact 

Statement, Waste Management Operations, Idaho National Engineering Laboratory, ERDA1536, Assistant Administrator for Environment and Safety, Washington, D.C., September. 

[SRDB Ref ID: 1143]

Ginkel, W. L., C. W. Bills, A. O. Dodd, K. K. Kennedy, and F. H. Tingey, 1960, Nuclear Incident at the 

Idaho Chemical Processing Plant on October 16, 1959, Report of the Investigating Committee, 

IDO-10035, Phillips Petroleum Company, National Reactor Testing Station, Idaho Falls, Idaho, 

February 15. [SRDB Ref ID: 12784]

Grothaus, G. E., 1993, “MDAs For In-Vivo Counting,” memorandum to C. Bloom (EG&G ES&Q), U.S. 

Department of Energy, RESL, February 19. [SRDB Ref ID: 7978]

Hayden, R. E., 1958, Incident Report, Ha-46-58A, memorandum to J. P. Lyon, Phillips Petroleum 

Company, Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho, 

March 26. [SRDB Ref ID: 8410]

Henry, R. N., and W. L. Slagle, 1985, "Report of MORT Investigation of N-Cell Incident," Hen-21-

85/Slag-2-85, memorandum to W. C. Mallory, Westinghouse Idaho Nuclear Company, Idaho 

National Engineering Laboratory, Idaho Falls, Idaho, April 30. [SRDB Ref ID: 8424]

Hill-Ellis, M., 2003, untitled facsimile of code information to F. Hinkley (Intrepid), Bechtel BWXT Idaho, 

Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho, October 1. 

[SRDB Ref ID: 29240]

Hoff, D. L., R. G. Mitchell, and R. Moore, 1989, The Idaho National Engineering Laboratory Site 

Environmental Report for Calendar Year 1988, IDO-12082(88), U.S. Department of Energy, 

Idaho Operations Office, Idaho Falls, Idaho, June. [SRDB Ref ID: 1686]

Horan, J. R., editor, 1959, Annual Report of the Health and Safety Division, 1958, Idaho Operations 

Office, IDO-12012, U.S. Atomic Energy Commission, Idaho Operations Office, Idaho Falls, 

Idaho. [SRDB Ref ID: 8079]

Horan, J. R., 1962, Health and Safety Division Annual Report 1961, IDO-12021, U.S. Atomic Energy 

Commission, Idaho Operations Office, Idaho Falls, Idaho, May 21. [SRDB Ref ID: 8076]

HPS (Health Physics Society), 1996, An American National Standard, Performance Criteria for 

Radiobioassay, ANSI N13.30-1996, American National Standards Institute, New York, New 

York. 

ICRP (International Commission on Radiological Protection), 1979, Limits for the Intake of 

Radionuclides by Workers, Part 1, Publication 30, Part 1, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 53 of 60

ICRP (International Commission on Radiological Protection), 1995, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England.

INEEL (Idaho National Engineering and Environmental Laboratory), 2001, "General Technical Basis 

and Facility Specific Documents," Attachment A, INEEL M&O Contractor Technical Basis for 

Internal Dosimetry, Idaho Falls, Idaho, May 30. [SRDB Ref ID: 8479]

INEEL (Idaho National Engineering and Environmental Laboratory), 2002, Analytical Chemistry 

Methods Manual, Volume VI (ACMM-VI), Rev. 13, Idaho Falls, Idaho, September 11. [SRDB 

Ref ID: 13456]

King, E. A., 1990, WINCO Internal Dose Technical Basis Document, Westinghouse Idaho Nuclear 

Company, Idaho National Engineering Laboratory, Idaho Falls, Idaho, December. [SRDB Ref 

ID: 26111]

King, V. A., 2001, Technical Basis for Internal Dosimetry at SMC, EDF No. SMC-2001-02, Rev. 3,

Idaho National Environmental and Engineering Laboratory, Idaho Falls, Idaho, May 31. 

[SRDB Ref ID: 8479, p. 67]

Martin, D. B., 1989, "MDAs For In-Vivo Counting," memorandum to C. Bloom (EG&G Idaho), U.S. 

Department of Energy, Idaho Operations Office, Idaho Falls, Idaho, November 16. [SRDB 

Ref ID: 7977]

Nertney, R. J., E. S. Brown, J. A. Buckham, L. J. Weber, and D. K. Jenson, 1967, The GA-18-1 

Contamination Incident at the TRA, June 13, 1967, NE-312-67, Idaho Nuclear Corporation, 

National Reactor Testing Station, Idaho Falls, Idaho, July 10. [SRDB Ref ID: 8381]

ORAUT (Oak Ridge Associated Universities Team), 2005, Idaho National Engineering and 

Environmental Laboratory (INEEL) – Site Description, ORAUT-TKBS-0007-2, Rev. 02, Oak 

Ridge, Tennessee, July 29. 

Paulus, R. C., A. O. Dodd, K. K. Kennedy, F. H. Tingey, and F. M. Warzel, 1961, Nuclear Incident at 

the Idaho Chemical Processing Plant, January 25, 1961, Final Report of the Investigating 

Committee, Phillips Petroleum Company, National Reactor Testing Station, Idaho Falls, Idaho, 

June 1. [SRDB Ref ID: 12735]

Percival, D. R., and J. I. Anderson, 1962, Whole-Body Counting at the National Reactor Testing 

Station, U.S. Atomic Energy Commission, Idaho Operations Office, Idaho Falls, Idaho. [SRDB 

Ref ID: 8399]

Puphal, K., 1994, Work Assessment History of INEL Bioassay, Final Report, Westat, Rockville, 

Maryland. [SRDB Ref ID: 14664]

Rich, B. L., 1958, "Thyroid Exposures From March 1958 Incident at ICPP," memorandum to 

J. W. McCaslin, BLR-83-58A, Phillips Petroleum Company, National Reactor Testing Station, 

Idaho Falls, Idaho, July 21. [SRDB Ref ID: 8013]

Rich, B. L., 1990, Internal Dosimetry and Control, EGG-ESQ-8808, EG&G Idaho, Idaho National 

Engineering Laboratory, Idaho Falls, Idaho, May. [SRDB Ref ID: 8412]
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 54 of 60

Rich, B. L., W. L. Slagle, C. L. Willis, H. D. Christiansen, C. E. Nichols, D. K. Jensen, D. R. Wenzel, C. 

W. Ison, and D. E. Black, 1974, Preliminary ICPP Health Physics Upgrade Program, ACI-167, 

Allied Chemical Corporation, Idaho National Engineering Laboratory, Idaho Falls, Idaho, 

October. [SRDB Ref ID: 14667]

Sill, C. W., J. I. Anderson, and D. R. Percival, 1964, “Comparison of Excretion Analysis with WholeBody Counting for Assessment of Internal Radioactive Contaminants,” Proceedings of

Symposium on the Assessment of Radioactive Body Burdens in Man, Heidelberg, 11-16 May 

1964, pp. 217–229. [SRDB Ref ID: 8003]

Sommers, J. F., 1954, "Radiation Incident involving GE-ANP-1 on 3/28/54," memorandum to J. P. 

Lyon, Som-24-54A, Phillips Petroleum Company, National Reactor Testing Station, Idaho 

Falls, Idaho, April 1. [SRDB Ref ID: 8386]

Sommers, J. F., 1958, "Incident Report on GEH-4 Rupture on 12-17-58," memorandum to J. P. Lyon, 

Som-126-58A, Phillips Petroleum Company, National Reactor Testing Station, Idaho Falls, 

Idaho, December 23. [SRDB Ref ID: 12711]

Sommers, J. F., 1961, "Whole Body Counting Procedures for MTR-ETR HP’s," memorandum to 

MTR/ETR Health and Safety Personnel, Som-106-61A-M, Phillips Petroleum Company, Idaho 

Falls, Idaho, September 27. [SRDB Ref ID: 8545]

Sommers, J. F., 1977, "Xenon Dose Determinations," memorandum to D. W. Marshall, TRA Safety 

Manager, Som-20-77, EG&G Idaho, Idaho National Engineering Laboratory, Idaho Falls, 

Idaho, April 26. [SRDB Ref ID: 8411]

Stacy, S. M., 2000, Proving the Principle, A History of the Idaho National Engineering and 

Environmental Laboratory 1949 - 1999, DOE/ID-10799, U.S. Department of Energy, Idaho 

Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1152, 1155, 1158, 1163, 1165, 1168, 

1171, 1173, 1177, 1179, 1182, 1184, 1186, 1188, 1191, 1192, 1193, 1195, 1196, 1199, 1200, 

1201, 1206, 1209, 1211, 1212, 1215]

Staiger, M. D., 2003, Calcined Waste Storage at the Idaho Nuclear Technology and Engineering 

Center, INEEL/EXT-98-00455, Rev. 1, Bechtel BWXT Idaho, Idaho National Engineering and 

Environmental Laboratory, Idaho Falls, Idaho, June. [SRDB Ref ID: 13458]

Till, J. E., K. R. Meyer, H. J. Mohler, and J. W. Aanenson 2002, Final Report, Identification and 

Prioritization of Radionuclide Releases from the Idaho National Engineering and 

Environmental Laboratory, 3-CDCTask Order 5-2000-FINAL, Radiological Assessment 

Corporation, Atlanta, Georgia. [SRDB Ref ID: 1141]

Tschaeche, A. N., 1988, “WINCO Internal Dosimetry Procedures Manual,” memorandum with 

attachment to Users of the WINCO Internal Dosimetry Procedures Manual, ANT-105-88, 

Westinghouse Idaho Nuclear Company, Idaho National Engineering Laboratory, Idaho Falls, 

Idaho, July 14. [SRDB Ref ID: 13435]

Vance, F. P., 1957, "Iodine-131 Incident," VA-59-57A, memorandum to J. P. Lyon, Phillips Petroleum 

Company, Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho. 

[SRDB Ref ID: 8539]
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 55 of 60

Voelz, G. L., 1969, 1968 Annual Report of the Health Services Laboratory, George L. Voelz, IDO12071, U.S. Atomic Energy Commission, Idaho Operations Office, Idaho Falls, Idaho, April. 

[SRDB Ref ID: 1359]

Wellus, M. K., and S. E. Stoven, 1994, "History of the INEL Bioassay Programs," Work Assignment 

200-93-2637-007, National Institute for Occupational Safety and Health, Idaho Falls, Idaho, 

May 2. [SRDB Ref ID: 14664]

Wenzel, D. R., 1973, "Internal PU Doses," Wen-2-73, memorandum to D. E. Black, Allied Chemical 

Corporation. Idaho Falls, Idaho, July 24. [SRDB Ref ID: 8014]

Wenzel, D. R., 1974, "Lung Doses for Mass Spect. Lab Pu Incident," letter to F. W. Cipperley (U.S. 

Atomic Energy Commission), DRW-2-74, Allied Chemical Corporation. Idaho Falls, Idaho, July 

3. [SRDB Ref ID: 7997]

Wenzel, D. R., 2000, Determination of the Normalized Mass of Individual Radionuclides in the 

Dissolver Product for Aluminum, Zirconium and Stainless Steel Fuels Previously Processed at 

INTEC, EDF INTEC-2000-006, Idaho National Engineering and Environmental Laboratory, 

Idaho Falls, Idaho. [SRDB Ref ID: 7973]

Wong, M. A., and R. S. Roberts, 1979, Description of the Stack Monitoring System and Effluent 

Measuring Techniques at The Idaho Chemical Processing Plant, ENICO-1005, Exxon Nuclear 

Idaho Company, Idaho National Engineering Laboratory, Idaho Falls, Idaho, August. [SRDB 

Ref ID: 13432]
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 56 of 60

GLOSSARY

absorption 

In relation to health physics, process in which radiation energy is imparted to and 

radionuclides are transported to bodily fluids, tissues, and organs. 

absorption type 

Categories for materials according to their speed of absorption in the body, which replaced the 

earlier inhalation clearance classes. Defined by the International Commission on Radiological 

Protection, the absorption types are F for fast absorption (formerly inhalation class D), M for 

moderate absorption (formerly inhalation class W), and S for slow absorption (formerly 

inhalation class Y). Also called solubility type. 

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.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 57 of 60

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 x 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

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. As 

examples, spent (used) fuel elements, byproduct tails, residues from uranium isotope 

separation, and some weapons materials contain DU. DU can be blended with highly 

enriched uranium to make reactor fuel or used as a raw material to produce plutonium.

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. 
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 58 of 60

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. 

in vitro 

Of or relating to a process that takes place under artificial conditions or outside a living 

organism (e.g., in the laboratory). From Latin meaning in glass. 

in vivo 

Of or relating to a process that takes place in a living organism. From Latin meaning in life. 

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. 

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., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties but often have different physical 

properties. 

milliPrefix that divides a unit by 1,000 (multiplies by 1 × 10-3). 

microPrefix that divides a unit by 1 million (multiplies by 1 × 10-6). 

mixed waste 

Unwanted material containing both radioactive and hazardous components. 

natural uranium (U, U-nat, NU) 

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

weight. The specific activity of this mixture is 2.6 × 107 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.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 59 of 60

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 

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei. See 

radionuclide. 

radionuclide 

Radioactive nuclide. See radioactivity and nuclide. 

radioactive lanthanum (RaLa)

(1) Isotope 140La (a fission product). (2) A lanthanum recovery process at INTEC for 

development of weapons.

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 spent nuclear fuel to separate useable fissionable 

products (i.e., uranium and plutonium) from waste material. Reprocessing was discontinued in 

the United States in 1992.
Document No. ORAUT-TKBS-0007-5 Revision No. 02 Effective Date: 07/30/2007 Page 60 of 60

shielding 

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

spent nuclear fuel

Fuel that has been in a reactor long enough to become ineffective because the proportion of 

fissile material has dropped below a certain level. Spent nuclear fuel contains fission and 

activation products.

spent fuel storage basin 

Pool or pit of reinforced concrete filled with water for storage of spent 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 

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei. Once formed, there is no difference between X-rays and 

gamma rays, but gamma photons originate inside 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.