ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 91

Document Title:

Idaho National Laboratory – Site Description

Document Number: ORAUT-TKBS-0007-2 

Revision: 03

Effective Date: 08/17/2007

Type of Document TBD

Supersedes: Revision 02

Subject Expert(s): Jo Ann M. Jenkins and Earl Graham

Site Expert(s): Don Marshall, Frank Hinckley, Boyd Leavitt, Norman D. Rohrig

Approval: Signature on File Approval Date: 08/07/2007

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 08/06/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 08/06/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 08/15/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 08/17/2007

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-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 2 of 91

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

11/07/2003 00 New technical basis document for the Idaho National Engineering 

and Environmental Laboratory – site description. First approved 

issue. Initiated by Norman D. Rohrig.

07/28/2004 01 Corrects dates of operation in Section 2.4.11 and adds OGC required 

paragraphs. Approved issue of Revision 01. Initiated by Norman D. 

Rohrig.

07/29/2005 02 Adds additional descriptive material on ICPP in response to union 

outreach comments. Incorporates formal internal and NIOSH review 

comments. Approved issue of Revision 02. Retraining is not 

required. Initiated by Norman D. Rohrig.

08/17/2007 03 Approved Revision 03 initiated to add Attributions and Annotations 

section. Added references. 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-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 3 of 91

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

7

2.1 Introduction ............................................................................................................................11

2.1.1 Purpose......................................................................................................................12

2.1.2 Scope.........................................................................................................................12

2.2 Test Area North......................................................................................................................15

2.2.1 Technical Support Facility........................................................................................... 15

2.2.1.1 Hot Shop/Warm Shop/Hot Cells Annex, 1955 to 2007...................................15

2.2.1.2 TAN 607 Storage Pool, 1955 to 2007 ............................................................16

2.2.1.3 Storage Pads TAN 790 and TAN 791, 1986 to 2005 .....................................16

2.2.1.4 Radioactive Liquid Waste Disposal System, 1959 to 2007 ............................17

2.2.1.5 Radioactive Parts Service and Storage Area, 1955 to 2004 ..........................17

2.2.1.6 TAN 607 Radiography Facility .......................................................................18

2.2.2 Initial Engine Test Facility, 1955 to 1966 .................................................................... 18

2.2.2.1 Heat Transfer Reactor Experiment No. 1, December 27, 1955, to 

January 3, 1959.............................................................................................18

2.2.2.2 Heat Transfer Reactor Experiment No. 2, July 1957 to March 28, 1961 ........19

2.2.2.3 Heat Transfer Experiment No. 3, 1958 to December 1960 ............................19

2.2.2.4 Systems for Nuclear Auxiliary Power 10A Transient No. 1, Early 1960s ........20

2.2.2.5 Systems for Nuclear Auxiliary Power 10A Transient No. 3, April 1, 1964.......20

2.2.2.6 Systems for Nuclear Auxiliary Power 10A Transient No. 2, 1965 to 

January 11, 1966...........................................................................................20

2.2.3 Water Reactor Research Test Facility, 1958 to 1973.................................................. 21

2.2.3.1 Shield Test Pool Facility ................................................................................21

2.2.3.2 Low Power Test Facility, 1958 to 1973 ..........................................................21

2.2.4 Loss-of-Fluid Test Facility, 1973 to July 9, 1985 ......................................................... 23

2.2.5 Specific Manufacturing Capability, 1985 to 2007 ........................................................ 24

2.3 Idaho Nuclear Technology and Engineering Center, November 1951 to 2007 ....................... 24

2.3.1 Fuel Processing Facility, CPP 601/602, February 1953 to 1992 ................................. 25

2.3.2 Ancillary Facilities ....................................................................................................... 27

2.3.3 INTEC Fuel Storage Facility (CPP-603), 1950 to 2007 ............................................... 29

2.3.4 High Level Liquid Waste Underground Storage Tanks (Tank Farm), 1951 to 2007..... 29

2.3.5 Waste Calcining Facility 1 and New Waste Calcining Facility, December 1963 to 

2002 ...........................................................................................................................30

2.3.6 Fluorinel Dissolution Process and Fuel Storage Facility, 1984 to 2007....................... 31

2.3.7 INL Comprehensive Environmental Response Compensation and Liability Act 

Disposal Facility Complex, July 2003 to 2007............................................................. 31

2.3.8 TMI-2 Independent Spent Fuel Storage Installation, CPP-1774, 1998 to 2007 ........... 32

2.4 Argonne National Laboratories–West, February 1951 to Present........................................... 32

2.4.1 Experimental Breeder Reactor No. I, April 1951 to December 30, 1963 ..................... 32

2.4.2 Boiling Water Reactor Experiment No. 1, Late 1953 to July 22, 1954......................... 33

2.4.3 Boiling Water Reactor Experiment No. 2, October 19, 1954, to March 1955............... 33

2.4.4 Boiling Water Reactor Experiment No. 3, June 9, 1955, to December 1956 ............... 34

2.4.5 Boiling Water Reactor Experiment No. 4, December 3, 1956, to June 1958 ............... 34
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 4 of 91

2.4.6 Boiling Water Reactor Experiment No. 5, February 9, 1962, to September 1964........ 35

2.4.7 Zero Power Reactor No. 3, October 1955 to November 1970..................................... 35

2.4.8 Argonne Fast Source Reactor, October 29, 1959, to Late 1970.................................. 36

2.4.9 Transient Reactor Test Facility, February 23, 1958, to April 1994............................... 36

2.4.10 Experimental Breeder Reactor No. II, May 1961 to September 30, 1994.................... 37

2.4.11 Hot Fuel Examination Facility, 1964 to 2007............................................................... 37

2.4.12 Zero Power Physics (Plutonium) Reactor, April 18, 1969 to April 1992 (Standby)....... 38

2.4.13 Neutron Radiography Facility, October 1, 1977 to 2006.............................................. 38

2.4.14 Fuel Assembly and Storage Building, 1970 to 2007.................................................... 39

2.4.15 Other Argonne National Laboratory–West Support Facilities ...................................... 39

2.5 Radioactive Waste Management Complex, May 1952 to 2007 .............................................. 40

2.5.1 Subsurface Disposal Area .......................................................................................... 40

2.5.2 Intermediate Level Transuranic Storage Facility ......................................................... 41

2.5.3 Transuranic Storage Area........................................................................................... 41

2.5.4 Stored Waste Examination Pilot Plant ........................................................................ 41

2.5.5 Transuranic Package Transporter Loading Station..................................................... 41

2.5.6 Advanced Mixed Waste Treatment Project................................................................. 41

2.6 Central Facilities Area, August 2, 1943, 2007......................................................................... 42

2.6.1 Hot Laundry................................................................................................................42

2.6.2 Health Physics Instrument Laboratory ........................................................................ 42

2.6.3 DOE Laboratory Accreditation Procedure Irradiation Facility ...................................... 43

2.6.4 Radiological and Environmental Sciences Laboratory ................................................ 43

2.6.5 CF-674 Building.......................................................................................................... 43

2.7 Test Reactor Area, March 31, 1952, to 2007.......................................................................... 43

2.7.1 Materials Test Reactor, March 31, 1952, to April 23, 1970 ......................................... 43

2.7.2 Engineering Test Reactor, September 19, 1957, to December 1981 .......................... 44

2.7.3 Advanced Test Reactor, July 7, 1967, to 2007 ........................................................... 45

2.7.4 Reactivity Measurement Facility, February 11, 1954, to April 10, 1962....................... 45

2.7.5 Advanced Reactivity Measurement Facility No. 1, October 10, 1960, to 1974 ............ 45

2.7.6 Advanced Reactivity Measurement Facility No. 2, December 14, 1962, to 1968 ........ 46

2.7.7 Coupled Fast Reactivity Measurement Facility, 1968 to 1991..................................... 46

2.7.8 Engineering Test Reactor Critical Facility, May 20, 1957 to 1982 ............................... 46

2.7.9 Advanced Test Reactor Critical Facility, May 19, 1964, to 2007 ................................. 47

2.7.10 TRA Hot Cell Facility, 1954 to 2007............................................................................ 47

2.7.11 TRA Gamma Facility, 1955 to Unknown ..................................................................... 47

2.7.12 Radiation Measurements Laboratory, 1952 to 2007 ................................................... 48

2.7.13 Radiochemistry Laboratories ...................................................................................... 48

2.7.14 Liquid Waste Disposal Ponds ..................................................................................... 48

2.7.15 High-Level Liquid Waste Tanks and Transfer Facility ................................................. 49

2.8 Auxiliary Reactor Area, April 1958 to Late 1990s ................................................................... 49

2.8.1 ARA-I ..........................................................................................................................49

2.8.2 ARA-II/SL-1, February 23, 1960, to April 6, 1961........................................................ 49

2.8.3 ARA-III........................................................................................................................50

2.8.4 ARA-IV, March 30, 1961 to May 29, 1964................................................................... 50

2.9 Waste Reduction Operations Complex/Power Burst Facility/Special Power Excursion 

Reactor Test Areas ................................................................................................................51
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 5 of 91

2.9.1 SPERT-I, June 11, 1955, to 1964 ............................................................................... 51

2.9.2 SPERT-II, March 11, 1960, to October 1964 .............................................................. 51

2.9.3 SPERT-III, December 19, 1958, to June 1968............................................................ 52

2.9.4 SPERT-IV, July 24, 1962, to August 1970 .................................................................. 52

2.9.5 Power Burst Facility, September 22, 1972, to 1985 .................................................... 52

2.9.6 Lead Storage Facility.................................................................................................. 53

2.9.7 Waste Experimental Reduction Facility, 1982 to 2001 ................................................ 53

2.9.8 Mixed Waste Storage Facility ..................................................................................... 53

2.10 Organic Moderated Reactor Experiment, September 17, 1957, to April 1963......................... 53

2.11 Test Grid III, 1957 to April 24, 1970........................................................................................ 54

2.11.1 Fuel Element Burn Tests ............................................................................................ 54

2.11.2 Fission Products Field Release Tests, July 25, 1958, to September 26, 1958 ............ 54

2.11.3 Relative Diffusion Tests, November 30, 1967, to October 1, 1969.............................. 55

2.11.4 Experimental Cloud Exposure Study, May 3, 1968, to April 24, 1970 ......................... 55

2.12 Experimental Field Station/Experimental Dairy Farm ............................................................. 55

2.12.1 Controlled Environmental Radioiodine (Release) Tests, May 27, 1963, to 

December 1977.......................................................................................................... 55

2.13 INL Research Center, 1984 to 2007....................................................................................... 56

2.14 Army Reentry Vehicle Facility Site (or Station), 1965 to May 1996......................................... 56

2.15 Attributions and Annotations .................................................................................................. 57

References ........................................................................................................................................73

Glossary ............................................................................................................................................80
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 6 of 91

LIST OF TABLES

TABLE TITLE

2-1 INL prime site contractors ......................................................................................................

PAGE

13

2-2 Other operating facility technical area contractors.................................................................. 13

2-3 Radionuclides of concern for all reactors................................................................................ 15

2-4 ICPP 601/602 process cell information................................................................................... 25

LIST OF FIGURES

FIGURE TITLE

2-1 Idaho National Laboratory......................................................................................................

PAGE

14
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 7 of 91

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AFSR Argonne Fast Source Reactor

ANL Argonne National Laboratory

ANL-E Argonne National Laboratory–East

ANL-W Argonne National Laboratory–West

ANP aircraft nuclear propulsion

ARA Army (later Auxiliary) Reactor Area

ARMF Advanced Reactivity Measurement Facility

ARVFS Army Reentry Vehicle Facility Site

ATR Advanced Test Reactor

ATRC Advanced Test Reactor Critical (facility) 

BNFL British Nuclear Fuels Limited

BORAX Boiling-Water Reactor Experiment 

BWR boiling-water reactor

CAM continuous air monitor

CDC capsule drive core

CE Critical Experiment

CERCLA Comprehensive Environmental Response, Compensation, and Liability Act

CERT Controlled Environmental Radioiodine (later Release) Test

CET Critical Experiment Tank

CFA Central Facilities Area

CFRMF Coupled Fast Reactivity Measurement Facility 

Ci curie

CPP Chemical Processing Plant

CRCE Cavity Reactor Critical Experiment

CTF Core Test Facility

d day

D&D decontamination and decommissioning

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DU depleted uranium

EBR Experimental Breeder Reactor

EBOR Experimental Beryllium Oxide Reactor

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EFS Experimental Field Station

EOCR Experimental Organic-Cooled Reactor

ETR Engineering Test Reactor

ETRC Engineering Test Reactor Critical (facility)

FASB Fuel Assembly and Storage Building

FAST Fluorinel Dissolution Process and Fuel Storage (facility) 

FDF Fluorinel Dissolution Facility

FEBT Fuel Element Burn Test

FMF Fuel Manufacturing Facility
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 8 of 91

FPF Fuel Processing Facility

FPR fuel processing restoration

ft foot

g gram

gal gallon

GCRE Gas-Cooled Reactor Experiment

GE General Electric Company

HEU highly enriched uranium

HFEF Hot Fuel Examination Facility

HLLW high-level liquid waste

HOTCE Hot Critical Experiment

HP health physics

HPIL Health Physics Instrument Laboratory

HPP Hot Pilot Plant (later Headend Processing Plant) 

hr hour

HTRE Heat Transfer Reactor Experiment

IC Initial Criticality

ICDF INL CERCLA Disposal Facility

ICP Idaho Cleanup Project

ICPP Idaho Chemical Processing Plant

ID, IDO Idaho Operations Office

IET Initial Engine Test

IFSF Irradiated Fuel Storage Facility

IMBA Integrated Modules for Bioassay Analysis

in. inch

INEEL Idaho National Engineering and Environmental Laboratory

INL Idaho National Laboratory

INTEC Idaho Nuclear Technology and Engineering Complex

IRC INL Research Center

IREP Interactive RadioEpidemiological Program

ISFSI independent spent fuel storage installation

IWP Industrial Waste Pond

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

kVp kilovolts-peak, applied kilovoltage

kW kilowatt

L liter

LOCA loss-of-coolant accident

LOFT Loss-of-Fluid Test

LPTF Low Power Test Facility

m meter

MAP mixed activation product

mCi millicurie

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission product
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 9 of 91

min minute

mL milliliter

ML-1 Mobile Low-Power Reactor No. 1

mo month

mph miles per hour

mR milliroentgen

mrem millirem

ms millisecond

MTR Materials Test Reactor

MW megawatt

NASA National Aeronautics and Space Administration

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NRAD Neutron Radiography Facility

NRC U.S. Nuclear Regulatory Commission

NRF Naval Reactors Facility

NRTS National Reactor Testing Station

NWCF New Waste Calcining Facility

OMRE Organic-Moderated Reactor Experiment

ORAU Oak Ridge Associated Universities

PBF Power Burst Facility

PIC pocket ionization chamber

PIF Process Improvement Facility 

POC probability of causation

psi pounds per square inch

psig pounds per square inch gauge

PWR pressurized-water reactor

R roentgen

R&D research and development

RAL Remote Analytical Laboratory

RaLa radioactive lanthanum

RAM remote area monitor

RCRA Resource Recovery and Conservation Act

RESL Radiological and Environmental Sciences Laboratory

RFP Rocky Flats Plant

RMF Reactivity Measurement Facility

RML Radiation Measurements Laboratory

RPSSA Radioactive Parts Service and Storage Area

RWMC Radioactive Waste Management Complex

s second

SCRCE Spherical Cavity Reactor Critical Experiment

SL-1 Stationary Low-Power Reactor No. 1

SMC Specific Manufacturing Capability

SNAP System for Nuclear Auxiliary Power

SNAPTRAN System for Nuclear Auxiliary Power Transients 

SPERT Special Power Excursion Reactor Test
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 10 of 91

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

STP Sewage Treatment Pond

STPF Shield Test Pool Facility

STR Submarine Thermal Reactor (S1W)

SWEPP Stored Waste Examination Pilot Plant

t thermal

TAN Test Area North

TBD technical basis document

TLD thermoluminescent dosimeter

TMI Three Mile Island

TRA Test Reactor Area

TREAT Transient Reactor Experiment and Test (facility)

TRIGA Training, Research, Isotope General Atomics (reactor)

TSA Transuranic Storage Area

TRU transuranic

TSF Technical Support Facility

U.S.C. United States Code

W watt

WCF Waste Calcining Facility

WERF Waste Experimental Reduction Facility

WIPP Waste Isolation Pilot Plant

wk week

WRRTF Water Reactor Research Test Facility

yd yard

ZPR Zero Power Reactor 

ZPPR Zero Power Plutonium (later Physics) Reactor

α alpha particle

β beta particle

γ gamma

μCi microcurie

°C degrees Celsius

°F degrees Fahrenheit

§ section or sections
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 11 of 91

2.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

derived:

] guidelines established under 

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

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

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

• Radiation from naturally occurring radon present in conventional structures

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 12 of 91

2.1.1

The purpose of this technical basis document (TBD) is to provide a description of the Idaho National 

Laboratory (INL) that contains technical basis information to evaluate the total individual occupational 

dose for claimants under EEOICPA. 

Purpose

INL played a major role in early reactor research and development. The site has operated 52 reactors 

plus fuel handling and reprocessing and radioactive waste storage and disposal facilities since it 

began operations in 1949. The purpose of this TBD is to assist in the evaluation of worker dose from 

INL processes using the methodology in the External Dose Reconstruction Implementation Guideline

(NIOSH 2006) and the Internal Dose Reconstruction Implementation Guideline (NIOSH 2002). 

2.1.2

This TBD provides supporting technical data with assumptions that are favorable to claimants to 

evaluate the total INL occupational dose that can be reasonably associated with worker radiation 

exposure covered under the EEOICPA legislation. The TBD addresses evaluation of monitored and 

unmonitored worker exposure and missed dose. In addition, it presents the technical basis of 

methods used to prepare INL worker dose records for input to the NIOSH Interactive

RadioEpidemiological Program (IREP) and the Integrated Modules for Bioassay Analysis (IMBA) 

computer codes used to evaluate worker dose. It also presents information on the uncertainty for 

recorded INL exposure and dose values.

Scope

This section describes the facilities and processes and historical information related to worker internal 

and external exposures for use when actual monitoring data might be unavailable. 

Proving the Principle, A History of the Idaho National Engineering and Environmental Laboratory, 

1949-1999 (Stacy 2000) was used as a resource in the development of facility and process 

descriptions.

INL is an 890-mi2 reservation with a maximum distance of about 39 mi from north to south and 36 mi 

from east to west. It is 30 to 60 mi west of Idaho Falls, Idaho. Major site-related research facilities 

and offices are in Idaho Falls. The site, situated on the Snake River Plain of southeastern Idaho at an 

elevation of about 5,000 ft, is above the Snake River Plain Aquifer. 

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

(NRTS) in Idaho as a Federal reservation to build, test, and operate nuclear reactors. The site utilized 

a variety of support facilities and equipment. In 1974, NRTS became the Idaho National Engineering 

Laboratory, which in 1997 became the Idaho National Engineering and Environmental Laboratory 

(INEEL). On February 1, 2005, the site became the Idaho National Laboratory, combining the 

research side of the INEEL and Argonne National Laboratory–West (ANL-W) and the Idaho Cleanup 

Project (ICP) working on closure of inactive portions of the site. (For convenience, this TBD refers to 

the site as INL.)

INL is unique among DOE facilities because it is a large complex site with many independent 

technical areas, contractors, goals, and missions. Table 2-1 lists historical prime site contractors and 

the years during which they operated. The prime contractor provided some services to the other 

contractors and operated most of the facilities. Some of the specific technical areas were operated in 

part by other contractors during the tenure of the prime contractor. Table 2-2 lists those facilities and 

technical areas and the years during which contractors other than the prime contractor operated them.
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 13 of 91

Table 2-1. INL prime site contractors (Stacy 2000). 

Operating years Prime contractor

1950–1966 Phillips Petroleum Company

1966–1972 Idaho Nuclear Corporation (Allied Chemical Corporation, Aerojet General Corporation, 

and Phillips Petroleum Company)

1972–1976 Aerojet Nuclear Corporation

1976–1994 EG&G Idaho

1994–1999 Lockheed Martin Idaho Technologies Company

1999–2004 Bechtel BWXT Idaho, LLC

2005 Battelle Energy Alliance for INL, CH2M Washington Group for ICP

Table 2-2. Other operating facility technical area contractors (Stacey 2000). 

Argonne National Laboratory–West

1949–2004 University of Chicago

Idaho Chemical Processing Plant (Idaho Nuclear Technology and Engineering 

Center)

1950–1953 American Cyanamid Company

1953–1966 Phillips Petroleum Company

1966–1971 Idaho Nuclear Corporation

1971–1979 Allied Chemical Corporation

1979–1984 Exxon Nuclear Idaho Company

1984–1994 Westinghouse Idaho Nuclear Company

Aircraft Nuclear Propulsion Program

1952–1961 General Electric Company

Specific Manufacturing Capability

1983–1986 Exxon Nuclear Idaho Company

1986–1991 Rockwell INEL

1991–1994 Babcock & Wilcox Idaho, Inc.

Dosimetry services at INL were unique among DOE facilities in that DOE Idaho Operations Office (ID) 

personnel operated and provided internal and external dosimetry services. The DOE-ID dosimetry 

branch provided and analyzed external dosimetry badges, counted workers in the whole-body 

counter, and analyzed bioassay samples. DOE-ID also provided portable radiation survey 

instruments and maintained and calibrated them. Originally, DOE-ID personnel at the Radiological 

and Environmental Sciences Laboratory (RESL) in Building CF-690 in Central Facilities Area (CFA) 

maintained exposure histories of personnel based on dosimetry records, including bioassay data. 

With the advent of the DOE Laboratory Accreditation Program (DOELAP), dosimetry responsibility 

transferred to the prime contractor on January 2, 1989, to eliminate a conflict of interest on the part of 

DOE-ID. 

Figure 2-1 shows the relative location of each facility or technical area discussed in this TBD. The 

following sections describe each facility and its processes, with the exception of the Naval Reactors 

Facility (NRF), a naval propulsion facility exempted under EEOICPA. The subsections for the facilities 

and technical areas and processes contain information on the particular area. 

Personnel working at INL in designated radiological areas were typically required to wear state-of-theart dosimetry [film badges, thermoluminescent dosimeters (TLDs), pocket ion chambers (PICs), 

respiratory protection, anticontamination clothing, etc.] [1]. Facilities and radiological areas were 

monitored by remote and portable remote area monitors (RAMs) and continuous air monitors (CAMs). 

Portable RAMs, CAMs, air samplers, etc., were used for work where fixed units were not available [2]. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 14 of 91

Figure 2-1. Idaho National Laboratory (Hoff 1985, 1986). Facilities include ANL-W, 

where the Experimental Breeder Reactor No. 2 (EBR-II), Transient Reactor 

Experiment and Test (TREAT) Facility, and Zero Power Plutonium (later Physics) 

Reactor (ZPPR) are located; Army (later Auxiliary) Reactor Area (ARA); CFA; Idaho 

Chemical Processing Plant (ICPP) [now the Idaho Nuclear Technology and 

Engineering Complex (INTEC)]; EBR-I; Boiling-Water Reactor Experiment (BORAX); 

Radioactive Waste Management Complex (RWMC); Special Power Excursion 

Reactor Test (SPERT) area; Grid III, the test grid where the Fuel Element Burn Tests 

(FEBTs) A and B occurred; Test Reactor Area (TRA); the Experimental Field Station 

(EFS); NRF; Test Area North (TAN), where the Initial Engine Tests (IETs) occurred; 

and the Core Test Facility at TAN.

In cases where airborne radioactivity might be present or internal exposure was possible, respirators 

were provided to prevent/reduce internal exposure [3]. 

Engineered systems were incorporated as practicable to minimize the potential for airborne 

radioactivity and direct radiation exposure. Bioassay programs were instituted to monitor and assess 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 15 of 91

potential internal exposures. Environmental monitoring systems were placed around the site at 

multiple locations to measure direct radiation, fallout, effluent discharges, or releases to onsite 

personnel and to members of the public. Each facility has had film or TLD badges in specific building 

areas and around perimeter fences to measure direct environmental radiation accumulation at the 

location as a check and balance on source terms in site locations [4]. 

In addition to the nuclear reactor experiments discussed below in relation to the TRA, there have been 

other such experiments at INL. Table 2-3 lists common radionuclides for reactors.

Table 2-3. Radionuclides of concern for all reactors. 241Am 60Co 131I 103Ru

140Ba 51Cr 133I 106Ru

141Ce 134Cs 140La 90Sr/90Y

143Ce 137Cs 54Mn 91Sr

144Ce 152Eu 95Nb 92Sr

242Cm 154Eu 239Np 234U

244Cm 59Fe 238Pu 65Zn

58Co 3

H 239Pu 95Zr

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

justification, or clarification of the associated information, are presented in Section 2.15. 

2.2 TEST AREA NORTH

TAN is 30 mi northeast of the CFA. General Electric Company (GE) opened TAN in 1952 for the 

Aircraft Nuclear Propulsion (ANP) Program, which operated during the 1950s and early 1960s. 

Additional facilities built to support the program were the IET Facility, the Technical Support Facility 

(TSF), and the Water Reactor Research Test Facility (WRRTF). 

During the 1970s, the Loss of Fluid Test (LOFT) Facility was built at the west end of TAN next to the 

aircraft hanger built to support the ANP Program. LOFT was a smaller version of a commercial 

pressurized-water reactor (PWR) designed to allow, create, or re-create loss-of-fluid accidents. In 

1983, construction started on the classified Project X, which later became the Specific Manufacturing 

Capability (SMC). The SMC, which was inside the aircraft hanger, manufactured depleted uranium 

(DU) armor for the M1-A1 Abrams Main Battle Tank.

2.2.1

TSF was built to provide technical and administrative support for the ANP Program. The facilities 

included a large Hot Shop, hot cells, storage pool, cafeteria, machine shop, office space, etc. A fourrail railroad system, with a shielded locomotive and two turntables, connected the IET Facility to the 

Hot Shop, the Warm Shop, and eventually the LOFT Facility. 

Technical Support Facility

2.2.1.1 Hot Shop/Warm Shop/Hot Cells Annex, 1955 to 2007

The TAN Hot Shop is in the TAN-607 building. It is 51 ft wide by 165 ft long by 55 ft high. The walls 

are 7 ft thick and the windows are 6 ft thick to provide protection to personnel involved in the 

examination, handling, analysis, or disassembly of radioactive components.
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The Warm Shop, immediately adjacent to the Hot Shop, is one large open room measuring 51 ft wide 

by 80 ft long by 50 ft high. A four-rail track system connects the Warm Shop to the Hot Shop and 

supports TAN projects. Experiments, projects, or equipment with relatively low-level contamination or 

direct radiation were brought to the Warm Shop for modification or repair.

The Hot Cell Annex is on the south side of the Hot Shop and consists of a hot cell and control 

galleries. The interior of the Hot Cell is 10 ft wide by 35 ft long. The Hot Cell Annex (TAN-633) is a 

one-story building north of and adjoining the storage pool. It consists of four shielded cells with 

interlocking sliding drawers for transferring samples. 

The Hot Shop and Hot Cells are equipped with cranes, manipulators, and other equipment for remote 

handling and work on experiments of all types. The Hot Shop was used extensively for refueling and 

repairs on the Heat Transfer Reactor Experiment (HTRE) reactors. Many other activities involving 

radiation levels measured up to 50 rad/hr for brief periods have occurred in the facilities, including 

disassembling the Stationary Low-Power Reactor (SL-1) and receiving, examining, and storing Three 

Mile Island (TMI) fuel and debris [5]. 

Internal exposure potential exists from airborne radioactivity of mixed fission products and activation 

products associated with reactor operation (Table 2-3) [6]. 

External exposure

2.2.1.2 TAN 607 Storage Pool, 1955 to 2007

 exists from mixed fission products and mixed activation products (MAPs). 

Radiation levels varied from background levels to measured levels greater than 50 rad/hr, principally 

gamma rays with energy greater than 250 keV [7]. 

The TAN Storage Pool is adjacent to the north side of the Hot Shop. It is 70 ft long, 48 ft wide, and 

24 ft deep. An underground passageway (19 ft by 24 ft) under the Hot Shop north wall connects the 

main pool to the Hot Shop vestibule. The vestibule, in the northeast corner of the Hot Shop, is 25 ft

long, 24 ft wide, and 24 ft deep. The top of the passageway under the shield wall is 5 ft under water 

to protect the main pool area from radiation sources in the Hot Shop.

The TAN Pool was used for storage of reactor fuel, including some commercial fuel and TMI core 

debris. In mid-2002, the TMI fuel and debris were moved to INTEC for storage. Radioactive 

materials being put in the pool or removed are generally transferred in shielded casks. The casks can 

be transferred via the vestibule in the Hot Shop or loaded directly on a truck in the north end of the 

pool. The pool is not lined and does not meet current criteria for radioactive storage pools (Bonney et 

al. 1995). 

The water in the Storage Pool is contaminated with mixed fission products (MFPs), activation 

products, and transuranic (TRU) materials. The major isotopes in the pool are 137Cs, 147Pm, 241Pu, 

241Am, and 90Sr, with an average water concentration of 4.6 × 10-4 μCi/mL (Bonney et al. 1995).

Internal exposure potential exists from airborne radioactivity, primarily from the nuclides listed above. 

External exposure

2.2.1.3 Storage Pads TAN 790 and TAN 791, 1986 to 2005

exists from the nuclides listed above. 

The Storage Pads are near the main east-west railroad west of the Hot Shop and east of the 

turntable. The TAN 790 pad was used to store three “Abnormal Waste Casks.” Two of the casks are 
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empty; the third contains filter elements with activity greater than or equal to 100 nCi/g of TRU 

materials (Bonney et al. 1995). External radiation levels are less than or equal to 0.1 mrem/hr [8]. 

Storage Pad TAN 791 is west of TAN 607 and contains the “Spent Fuel Storage Casks.” These casks 

were designed to investigate the feasibility of storing spent nuclear fuel from commercial reactors in a 

dry state without external cooling requirements. The casks have fuel from a Virginia Electric & Power 

Company reactor and from LOFT. The casks, which are filled with helium gas and sealed, are 

monitored for temperature and pressure, and alarm if either deviates from accepted limits. Measured 

radiation levels near the casks are 25 to 30 mrem/hr gamma and about 40 mrem/hr neutron. The 

casks are periodically surveyed and assessed for radioactive contamination [9]. 

Internal exposure potential exists from potential airborne leaks [10]. 

External exposure

2.2.1.4 Radioactive Liquid Waste Disposal System, 1959 to 2007

exists from activities near the casks. Not all personnel wear neutron dosimetry

[11]. 

A radioactive liquid waste system (TAN 666), which was built in 1959, collected and processed 

intermediate-level liquid wastes generated in TAN and transferred them to one of three underground 

10,000-gal, stainless-steel collection tanks (V1, V2, and V3). The liquid wastes were concentrated in 

an evaporator in TAN 616, and the concentrated solution was pumped to one of two 50,000-gal 

underground liquid waste feed tanks (V9 and V10). Solids were separated and sent to the RWMC. 

The liquid effluent was originally combined with low-level radioactive liquid waste and discharged to a 

disposal well. In 1972, INL replaced the disposal well with a disposal pond. The pond was an unlined 

diked area encompassing about 35 acres that could receive about 33 million gal per year (ERDA 

1977). From 1959 to 1974, TAN reported liquid effluent releases to the disposal well or pond of 58 Ci, 

with highest releases in 1959, 1968, and 1969 (ERDA 1977, p 143). 

Internal exposure exists from potential airborne MFPs and activation products (Bonney et al. 1995). 

External exposure

2.2.1.5 Radioactive Parts Service and Storage Area, 1955 to 2004

exists from activities around the liquid waste system [12]. 

Storage buildings (TAN 647/TAN 648) and the adjacent storage pads are known as the Radioactive 

Parts Service and Storage Area (RPSSA). The facility is in the northwest portion of the TAN TSF 

area. The area has residual contamination from earlier projects, including the HTRE-2 and HTRE-3 

reactors. The buildings and contaminated areas are marked with perimeter fencing and warning 

signs. Some of the radiologically contaminated soil in the area resulted from movement of SL-1 

debris into the Hot Shop, and some particles were measured as high as 50 rad/hr beta near contact 

(

90Sr/Y). The gates and buildings are locked when access is not required [13]. 

The RPSSA buildings and areas on and around the pads contain casks, boxes, and equipment of all 

types that is either contaminated or radioactive and has to be in a controlled storage area.

Internal exposure potential exists from airborne radioactivity from leaking packages or disturbing 

contaminated soil areas containing MFPs and/or activation products [14]. 

External exposure comes from several items in storage that have had radiation levels in the roentgenper-hour range from MFPs and activation products [15]. 
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2.2.1.6 TAN 607 Radiography Facility

The TAN 607 Radiography Facility was in TAN 607, south of the TAN Warm Shop. The facility used 

60Co and 192Ir sources and X-ray units for nondestructive examination of metal welds, parts, or 

equipment. Some of the isotopic sources were in the 100-Ci range when first purchased [16]. 

Radiography was performed by trained and certified technicians according to approved procedures. 

Periodic radiography would be required on a reactor or piece of equipment that could not be brought 

to the facility. Applicable procedures were required to evaluate radiological hazards and establish 

necessary controls. The Radiography Facility is not operational at present.

Internal exposure is negligible for radiographic activities in the Radiography Facility [17]. 

External exposure

2.2.2

occurred from activities associated with radiographic sources [18]. 

The IET (TAN 620) was approximately 1 mi north of the TAN TSF area. It was built as a shielded test 

facility to prove that heat from a nuclear energy source could run a turbojet engine. The IET had highdensity reinforced concrete walls 2 ft thick and ceilings 3 ft thick. The floor of the facility was 15 ft

below grade with 14 ft of dirt over the top. After a HTRE test, a locomotive driver in a shielded cab 

would hook onto the reactor and return it to the TAN TSF Hot Shop. There were a total of 26 IET runs 

involving three separate reactor assemblies – HTRE 1, 2, and 3 (DOE 1991a). Jet engines were fitted 

to a HTRE reactor at TSF and transported by the four-rail system to the IET facility.

Initial Engine Test Facility, 1955 to 1966

During the early 1960s, the AEC initiated another nuclear safety program at IET, called Systems for 

Nuclear Auxiliary Power (SNAP) Transients (SNAPTRAN). The program evaluated the hazards 

associated with using nuclear reactors for aerospace auxiliary power systems. In addition, it was 

designed to investigate the consequences of a nuclear accident. Three SNAPTRAN tests were 

conducted, with the last two ending in destruction of the SNAP 10A/2 reactors.

The test series involved a number of power reactor operations that resulted in the release of 

radionuclides to the environment. The dates and times of reactor operations and consequent 

releases varied throughout the duration of the project. Therefore, the meteorological conditions that 

existed depended on the time of the tests. All operations were under strict meteorological control, and 

permissible wind directions seriously limited operations. On many days, it was impossible to operate 

at all, and most of the time operation was possible only a few hours each day (Thornton and Rothstein

1962).

Internal exposure was possible from airborne MFPs and activation products. It was minimized by a 

filtered intake air system that kept the air pressure in occupied areas positive in relation to outside air. 

The exhaust from the HTRE jet engines was vented up a 150-ft stack (Thornton and Rothstein 1962, 

p154). 

External exposure

2.2.2.1 Heat Transfer Reactor Experiment No. 1, December 27, 1955, to January 3, 1959

occurred from the activation and MFP inventory after tests were completed and 

personnel worked on the associated test assemblies [19]. 

On November 4, 1955, the HTRE-1 reactor was placed on the test pad at the IET and brought to 

power. On December 30, 1955, the reactor was again brought to power with a J47 turbojet engine 

attached. The reactor was run on the test stand for 150.8 hr at full power, 20 MW thermal, exceeding 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 19 of 91

the design requirement of 100 hr. During the first 6 hr of full-power operation, fuel element damage 

occurred in three cartridges caused by a defect in the insulation liners. After the damaged elements 

were replaced, power operation resumed. The test was successful in proving the reactor could run 

the engines without chemical fuel. The water-moderated and water-cooled reactor used enriched 

uranium fuel clad in nickel-chromium (Thornton and Rothstein 1962).

2.2.2.2 Heat Transfer Reactor Experiment No. 2, July 1957 to March 28, 1961

The HTRE-2 "parent core" was similar to the HTRE-1 core except the central seven air tubes were 

removed and replaced by a hexagonal void 11 in. across the flat. A corresponding opening was made 

in the top shield plug so sections of advanced reactors could be inserted in the parent core without 

removal of the core from the shield. This converted it to a materials test reactor, which subjected test 

fuels to environments reaching 2,800°F for extended periods and higher temperatures for short 

periods. The ANP Program advanced the technology of high-heat ceramic reactor fuels. HTRE-2 

operated for 1,299 hr at powers up to 14 MW. Test sections consisted of metallic fuel elements

combined with air-cooled zirconium hydride moderators and beryllium oxide fuel elements for use in 

ceramic reactors (Thornton and Rothstein 1962; Stacy 2000).

2.2.2.3 Heat Transfer Experiment No. 3, 1958 to December 1960

A new HTRE engine was developed with the reactor, engine shielding, and heat transfer system 

arranged in a horizontal configuration anticipating final design in an airframe. HTRE-3 operated for 

126 hr, running two modified J47 jet engines at power as high as 32 MW thermal. This reactor was 

water-cooled and water-moderated, and used uranium fuel clad in nickel-chromium. In December 

1960, HTRE-3 ran two turbojet engines at 2,000°F without the help of any chemical fuel (Thornton

and Rothstein 1962).

On November 18, 1958, a nuclear excursion occurred during IET No. 13. The excursion was the 

result of the dynamic and shim rods being withdrawn by the control system under the influence of an 

erroneous reactor power indication. Activity was released from the exhaust stack and a narrow band 

of fallout was contained fully within the INL boundaries. The maximum dose rate observed in the 

Assembly and Maintenance area and approximately 3,000 ft from the cloud centerline was 

0.04 mrem/hr. The maximum fallout observed, at about 4 hr after the incident, measured 0.8 to 

2.0 mrem/hr at contact roughly 1-1/2 mi from IET (DOE 1991a).

After each HTRE test, the cooling water was drained and replaced with mercury to provide shielding 

to personnel working on the reactors (Stacy 2000). After the HTRE program ended, HTRE-2 and -3 

were parked in the RPSSA. The HTRE engines were moved to the EBR-I historic site in 1988.

Internal exposure potential existed from the airborne release inventory during reactor operation as 

argon and other constituents in the cooling air became radioactive. Fuel elements occasionally 

ruptured, which released MFPs and activation products to the outdoor environment. In some cases 

the cooling air flow to fuel elements was deliberately blocked to determine fuel failure parameters and 

characteristics. Fourteen of the tests were categorized as Group 1 releases under National Council 

on Radiation Protection and Measurements methodology (NCRP 1996); tests 10 and 4 were ranked 

the highest. Seven tests were ranked as Group 2. Fission product inventories were based on 

documented reactor operating histories (Till et al. 2002). Operations and support personnel were 

inside the pressurized control room (TAN 620) during reactor operation during the testing phase of the 

HTRE program, minimizing their exposure potential. 
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External exposure

2.2.2.4 Systems for Nuclear Auxiliary Power 10A Transient No. 1, Early 1960s

occurred from the MFP and activation product inventory during work associated 

with test assemblies after tests were completed [20]. 

The AEC initiated a safety program to evaluate hazards associated with using nuclear reactors for 

aerospace power systems. The portion of the program concerned with determining the kinetic 

behavior of the SNAP 10A/2 reactors and the consequences of certain nuclear accidents involving 

these reactors was called SNAPTRAN. The tests were conducted at the IET facility at TAN. The 

SNAPTRAN program extended the SPERT reactor safety testing program (see Section 2.9) to 

aerospace applications. Three test series involving three reactors investigated the behavior of SNAP

10A/2 fuel under large-transient, power-excursion conditions. SNAPTRAN-1 was subjected to 

nondestructive tests in conditions approaching but not resulting in damage to the zirconium-hydrideuranium fuel (Stacy 2000). The SNAP 10A/2 reactors were 9 in. in diameter by 12 in. long and were 

composed of a sodium/potassium (NaK)-cooled core containing 37 rods of fully enriched uranium in a 

zirconium-hydride matrix (Cordes et al. 1967). 

2.2.2.5 Systems for Nuclear Auxiliary Power 10A Transient No. 3, April 1, 1964

SNAPTRAN-3 was the first of two destructive tests on a version of the small reactor (SNAP 10A/2) 

designed to supply auxiliary power in space. The test, conducted at the IET facility on April 1, 1964, 

simulated the accidental fall of a reactor into water or wet earth, as could occur during assembly, 

transport, or launch abort. The test demonstrated that the reactor would destroy itself immediately 

rather than building up a high inventory of radioactive fission products. The test involved reactor 

operation at a power level of 30,000 MW for 1.5 ms. More than 99% of the fission product inventory 

was retained in the surrounding water and reactor fuel remains. No airborne iodine was detected, so 

it was presumed that halogens were retained in the water as well as particulate radionuclides (Cordes 

et al. 1965). 

2.2.2.6 Systems for Nuclear Auxiliary Power 10A Transient No. 2, 1965 to January 11, 1966

This test version of the small space reactor, SNAP 10A/2, was intentionally destroyed on January 11, 

1966. It provided information on the dynamic response, fuel behavior, and inherent shutdown 

mechanisms of these reactors in an open-air environment. In normal operation, the control drums of 

the SNAP 10A/2 would be rotated to obtain criticality after the reactor was in orbit. In a launch abort, 

however, impact on the earth might cause the reactor drums to rotate inward and the core to go 

critical and conceivably destroy itself, releasing fission products to the surrounding environment. The 

test data contributed to an understanding of reactor disassembly on impact and methods for 

assessing or predicting the radiological consequences. The reactor core was 93% enriched fuel, 

containing 4.75 kg of 235U. The reactor operated at a power level of 36,000 MW for 1.5 ms. The 

fission product release fractions were reported to be 0.75 for noble gases, 0.70 for iodines, 0.45 for 

tellurium, and 0.04 for solids. Reactor operation was assumed to generate 4,681 Ci of 41Ar. 

SNAPTRAN meteorological conditions for testing were strict. Weather requirements were to consist 

of lapse conditions with no rainfall, and were to persist a minimum of 3 hr after the tests. Wind had to 

be southwest (180 ̊ to 240 ̊) between 10 and 30 mph (Cordes et al. 1967).

Internal exposure potential during the SNAP 10A test series was well controlled by requiring 

personnel to stay inside the IET facility during tests. During the SNAPTRAN-2 and -3 tests, the 

reactors were totally destroyed, releasing fission products including 131I, all of which was tracked 

(Cordes et al. 1965, 1967).
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External exposure

2.2.3

to MFPs and MAPs was received during cleanup of the reactor debris, which was 

scattered around the test pad and along the four-track rail system. Dosimetry was required for all 

personnel [21]. 

The WRRTF is approximately 1.25 mi southeast of the TSF area. The first facility constructed at 

WRRTF was the Low Power Test Facility (LPTF) in 1958. This facility, also known as Semiscale, was 

in the east quadrant of WRRTF. The Shield Test Pool Facility (STPF), in the west quadrant of 

WRRTF, was built as part of the ANP Shielding Experimentation Program; in 1963, the pool facility 

was modified for the Experimental Beryllium Oxide Reactor (EBOR). The EBOR project was 

cancelled in 1966 before construction was complete (Stacy 2000, p. 262). 

Water Reactor Research Test Facility, 1958 to 1973

The Semiscale facility in LPTF was a forerunner of the LOFT. It was a scaled mockup of one loop of 

a four-loop PWR. The facility was electrically heated to provide steam to run blowdown tests (Shaw, 

Boucher, and Loomis undated). No radiological exposure resulted from the blowdown tests [22]. 

Due to the electrically heated experiment cores in the facilities discussed in the following sections, 

internal and external exposure from neutron reactions in the reactors did not exist [23]. Exposures will 

be noted as applicable at the end of the pertinent test series.

2.2.3.1 Shield Test Pool Facility

The STPF, initially used for a reactor experiment known as Susie, was used for bulk shielding 

experiments performed in support of the ANP Shielding Experimentation Program. Susie utilized a 

sample canister box that was pressurized with inert gas or air to keep it dry. It was at the reactor 

centerline and contained 16 sample tubes in which organic samples were irradiated. Susie was a 

swimming-pool-type reactor; water-moderated, water-cooled, water-reflected, and shielded by 

approximately 17 ft of water. The fuel loading was 4 kg of 235U, and the nominal power level was 

2 MW thermal. After the ANP Program ended in 1961, Susie was used by other programs at INL

(Walsh 1961). The facility would later become known as EBOR. After the EBOR project was 

cancelled, other experiments were conducted using tracer-level nuclides to investigate the reaction of 

water in piping, as well as instrument calibrations. The cell area was filled with piping for pressurizedwater tests using up to 25-Ci 137Cs sources in a radiography-type environment in which the source 

material was cranked out of a shielding cask to be in proximity to the piping. The cesium and tracer 

sources have been removed [24]. 

Internal exposure at the STPF was not known to occur due to the low power of Susie, the water 

environment, and use of sealed small sources [25]. 

External exposures

2.2.3.2 Low Power Test Facility, 1958 to 1973

occurred from the use of sealed sources, primarily 137Cs [26]. 

The LPTF was used to conduct several low-power (less than 100-W) reactor research programs. The 

LPTF contained two shielded cells with three independent control rooms and necessary support 

facilities. The north cell, Room 101, was called the Critical Experiment (CE) cell and the south cell, 

Room 102, was the Initial Criticality (IC) cell. The test cells are of poured concrete construction with a 

4-ft-thick wall between them. The walls between the cells and the control room are 5 ft thick and 30 ft

high. The outside wall of the IC cell is 2 ft thick, and the outside wall of the CE cell is 3 ft thick and

30 ft high (Kunze and Chase 1970). 
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The construction of the facility was such that more than one reactor program could be running at the 

same time [27]. Heavy experiment pieces could be moved in or out of each cell through large rollup 

doors in the back. 

2.2.3.2.1 Critical Experiment Tank, 1958 to 1960

The Critical Experiment Tank (CET) reactor was part of the ANP Program in the CE Cell of the LPTF. 

The CET was a low-power reactor originally designed to mock-up the HTRE-1 and HTRE-2 reactors. 

It was used primarily to perform critical experiments for insert tests in the HTRE-2 power plant. Fuel 

test bundles intended for testing in HTRE-2 were first evaluated for reactivity characteristics in the 

CET. The water-moderated CET utilized a beryllium reflector (Becar et al. 1961).

CET was one of three low-power reactors supporting the ANP Program, along with the STPF Reactor 

(Susie) and the Hot Critical Experiment (HOTCE) (Hoefer 1957). The ANP Program ended in 1961.

2.2.3.2.2 Hot Critical Experiment, 1958 to March 28, 1961

Located in the LPTF CE cell, HOTCE was an elevated-temperature critical experiment designed to 

obtain information on temperature coefficients of solid moderated reactors. The fuel elements 

consisted of fuel bearing stainless-steel wire 1/8 in. in diameter. The maximum loading was 50.4 kg 

of 93.2% enriched UO2. The reactor used a hydrided zirconium moderator and a beryllium reflector. 

The hexagonal prism-shaped core and reflector were mounted such that the fuel cells were horizontal. 

One half was mounted on a fixed table and the other in a movable table so the two halves were 

separated. The normal operating power was 1 W for a period of 1 to 3 hr. The reactor could be 

operated at 100 W for short periods (Hoefer 1957). 

2.2.3.2.3 Split Table Reactor System, 1971

The purpose of the Split Table Reactor System was to provide nuclear information on a variety of fast 

and thermal spectrum reactors. The reactors were to be assembled, operated, and revised to perform 

experiments with both thermal and fast systems. The reactor operated in the CE cell of the LPTF.

The reactor was a split-table type, 7 ft 2 in. wide and 11 ft long. A hydraulic system opened and 

closed the reactor. The table was an aluminum matrix structure composed of a stack of hexagonal 

tubes mounted horizontally on each table half. When the two halves were brought together, a single 

reactor assembly was formed. Normal operating power level was 0 to 500 W thermal not to exceed 

1,000 W or 10 kW-hr per month (Lofthouse 1971).

2.2.3.2.4 Fast Spectrum Refractory Metals Reactor (710), March 1962 to 1968

The 710 reactor was a split-table, low-power critical facility at LPTF. The objective was to collect data 

for a proposed fast-spectrum, refractory-metal reactor concept called the 710 Reactor. The concept 

involved using metals such as tungsten and tantalum in a compact, very-high-temperature reactor for 

generating power in space (Stacy 2000, p. 263). 

2.2.3.2.5 Cavity Reactor Critical Experiment, May 17, 1967, to Early 1970s

The Cavity Reactor Critical Experiment (CRCE) was an outgrowth of a program started by the 

National Aeronautics and Space Administration (NASA) at LPTF to investigate the propulsion of 

space rockets by nuclear power, offering the possibility of much greater thrust per pound of propellant 

than chemical rockets. The concept for the cavity reactor core was that the uranium would be in a 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 23 of 91

vapor, or gaseous state. Hydrogen propellant flowing around it would theoretically attain much higher 

temperatures (up to 10,000°F) than in conventional solid-core rockets. The experiments at TAN used 

simulated hydrogen propellant and produced data on the reactor physics feasibility of a gaseous core 

being able to go critical. The core was uranium hexafluoride (UF6); the experiments were performed 

at the relatively low temperature of about 200°F. In the proposed ultimate application, the ball of 

uranium gas would have been held in place by the hydrogen flowing around it, something like a pingpong ball suspended in a stream of air. Uranium core temperatures as high as 100,000°F were 

considered possible. During the CRCE program, the UF6 was always contained in the storage vessel 

or the reactor cavity (Stacy 2000, p. 261). 

2.2.3.2.6 Spherical Cavity Reactor Critical Facility, 1972 to November 1973

The Spherical Cavity Reactor Critical Facility (SCRCE) was the final experiment in reactor physics 

work for the NASA-sponsored program to determine the feasibility of a reactor going critical with a 

gaseous core of UF6. Previous work used a cylindrical configuration because of its ease of 

construction. The spherical shape was considered a more likely geometry for the ultimate application 

in a rocket to Mars. The SCRCE assembly consisted of two aluminum tanks, one inside the other, 

with D2O in the space between the two tanks. The D2O would act as a reflector and moderator and, 

during normal shutdown, would be transferred from the reactor to a storage tank (INC 1969).

2.2.3.2.7 High Temperature Marine Propulsion Reactor (630-A), 1962 to 1964

The 630-A reactor was a low-power critical experiment operated at the LPTF. The mission of the 

630-A was to explore the feasibility of an air-cooled, water-moderated system for nuclear-powered 

merchant ships. Development ended in December 1964 after decisions to lower the priority of the 

entire Nuclear Power Merchant Ship Program (Stacy 2000, p. 261). 

Internal exposure was not known to occur [28]. 

External exposures

2.2.4

occurred during cell entry after a reactor run [29]. 

The LOFT reactor at TAN 650 was a centerpiece in the safety testing program for commercial power 

reactors. The reactor was a scale model of a commercial pressurized-water power plant built to 

explore the effects of loss-of-coolant accidents (LOCAs). Thirty-eight nuclear power tests were 

conducted with various accident scenarios, including the accident at TMI. Among other goals, the 

program investigated the capability of emergency core cooling systems to prevent core damage 

during a loss-of-coolant accident (LOCA). Experiments at LOFT simulated small-, medium-, and 

large-break LOCAs, sometimes complicated with other events such as “loss of offsite power.” LOFT 

was deactivated in 1986, following completion of the LP-FP-2 experiment, the most significant severe 

fuel damage test ever conducted in a nuclear reactor. That test, which involved the heating and 

melting of a 100-rod experimental fuel bundle, provided information on the release and transport of 

fission products that could happen during an actual commercial reactor accident where core damage 

occurred (Stacy 2000, p. 264). 

Loss-of-Fluid Test Facility, 1973 to July 9, 1985

The LOFT facility was built in a steel-domed reactor containment vessel, 97 ft high, with a basement. 

The LOFT Control and Support Building is four stories high with a basement. The structures are 

attached at the basement level, and both have reinforced concrete exterior walls. 
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The control room, visitor center, experimental data recording and display area, sample counting area, 

and operation support room were in TAN 630, a two-story underground building. During the final 

preparations for a test, the containment vessel doors were closed and the only access to the facility 

was through a shielded underground tunnel. During reactor operation and testing, personnel were 

restricted to safe facilities in TAN 630.

Internal exposure was possible from airborne fission product activity in the containment soon after 

shutdown. Entries were monitored with a CAM, and respiratory protection was worn as required. On 

July 9, 1985, following completion of the LP-FP-2 test, leakage was discovered from the fission

product monitoring system and the primary coolant system, allowing fission products to enter the 

reactor building. Over the following 2-mo period, 8,780 Ci of noble gas (88Kr) and 0.09 Ci of iodine 

(

131I) were released to the environment (Hoff, Chew, and Rope 1986; Author unknown 1985).

External exposure

2.2.5

occurred to personnel working inside the containment vessel or on the primary 

system or sample systems associated with the reactor. During initial entry after a test, the fields in 

containment were ≥ 100 mrad/hr beta-gamma. The short-lived fission products would die off rapidly, 

reducing the general fields to ≤ 10 mrad/hr beta gamma [30]. 

The SMC project is in the ANP Program aircraft hanger (TAN 629) and surrounding buildings. The 

project consists of Phase I, Phase II, and support facilities. A Materials Development Facility, located 

in TAN 607A and the TAN Hazardous Waste Storage Area, located in TAN-628, were part of the SMC 

project, but have been decommissioned and turned over for other use. The SMC project is classified. 

In 1991, the mission was declassified in that the SMC manufactures armor made of DU for the U.S. 

Army. The major radioactivity in the DU is 238U, 234Th, 234mPa, and 234U (INEEL 2001). 

Specific Manufacturing Capability, 1985 to 2007

The SMC facilities consist of Phase I in TAN 629 and Phase II in TAN-679 and TAN-681. Metal 

fabrication activities are performed in Phase I facilities. Phase II facilities perform metal rolling in 

TAN-679 and waste processing in TAN-681. SMC formerly used a nitric acid system; this waste was 

processed in TAN-681. This system was removed and replaced with an aqueous system. The 

aqueous waste is processed in TAN-681. All radioactive aqueous waste is collected in storage tanks 

for treatment through an evaporator system and the remaining aqueous waste is solidified and 

disposed of as low-level radioactive waste.

Internal exposure occurred from normal operations using the DU processes. Metal fabrication is the 

primary source of airborne radiological activity, followed by a paint coating process [31]. 

External exposure

2.3 IDAHO NUCLEAR TECHNOLOGY AND ENGINEERING CENTER, NOVEMBER 1951 

TO 2007

occurred through working with the billets of DU. The large pieces were primarily 

handled remotely to minimize exposure [32]. 

INTEC, formerly the Idaho Chemical Processing Plant (ICPP or CPP), commonly known as the 

“Chem Plant”, is 53 mi west of Idaho Falls and occupies 200 acres in the middle of the INL

reservation. The Plant stored and processed spent nuclear fuel from university and test reactors all 

over the world, from commercial power plants, from most of the reactors at INL, and from U.S. 

Department of Defense (DOD) projects. INTEC received the first fuel shipment in November 1951. 

The first hot run started in February 1953. Reprocessing continued until the fuel reprocessing project 

was cancelled in 1992. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 25 of 91

The primary INTEC mission involved reprocessing spent nuclear fuel with highly enriched uranium, 

which entailed extracting reusable uranium from spent fuels. Each cladding (e.g., aluminum, 

zirconium, stainless-steel, and graphite) or fuel type called for a different process. The numerous fuel 

recycling processes required support facilities for fuel dissolution and recovery of fissionable materials 

(e.g., 29 different process cells) and waste processing. An innovative high-level liquid waste (HLLW) 

treatment process known as calcining was developed at INTEC as an additional major mission. 

Calcination reduced the volume of liquid radioactive waste generated during reprocessing and placed 

it in a more stable solid granular form.

In addition to calcination and fuel reprocessing operations, INTEC is a major fuel storage facility. 

INTEC houses a HLLW evaporation facility, HLLW storage (Tank Farm), airborne radioactive waste 

processing, nonradioactive liquid waste disposal, and the Remote Analytical Laboratory (RAL). 

Although fuel reprocessing ended in 1992, INTEC continues to support other nuclear projects. For 

example, chemical research continues to improve fuel recovery processes, spent nuclear fuel is safely 

stored and prepared for shipment to an offsite repository, development of technology to treat safely 

high-level and liquid radioactive waste that resulted from reprocessing spent fuel continues, past 

environmental releases are being remediated, and some facilities have undergone decontamination

and decommissioning (D&D). Inactive INTEC facilities are being evaluated for D&D.

Internal exposure could have occurred when workers were near breached and/or leaking systems 

containing mixed fission products, TRU materials, activation products, etc. [33].

External exposure 

2.3.1

occurred during maintenance work, laboratory work, fuel cutting, and other support 

work that briefly exposed workers to radiation rates from background levels to measured levels of 

≥ 50 rad/hr beta-gamma [34]. 

The INTEC Fuel Processing Facility (FPF), CPP 601/602, was used for the chemical separation of 

highly enriched uranium (HEU) from dissolved spent fuel during reprocessing and to solidify the 

recovered HEU for shipment off the site. The process dissolved the fuel in acid, producing uranyl 

nitrate and nitrates of fission products and some TRU materials. Solvent extraction with hexone and 

tributyl phosphate (Boardman ca. 1956) separated uranium from the fission products. 

Fuel Processing Facility, CPP 601/602, February 1953 to 1992

The Process Building, CPP 601, contains 29 heavily shielded underground process cells. The 

building is 250 ft long by 102 ft wide and extends to 57 ft below grade and 38 ft above grade and 

features at least seven corridors for different functions (Cederberg et al. 1974). Table 2-4 lists the 

cells and processes associated with the fuel reprocessing. With the exception of the Health Physics 

(HP) field office in V cell, cells were accessible only during shutdown periods for maintenance and 

decontamination activities. The process building was designed to handle modest quantities (up to 

several kilograms) of enriched spent fuel due to criticality considerations. It was designed to be 

remotely decontaminated so hands-on maintenance could occur. The use of hands-on maintenance 

results in many activities being conducted with significant dose rates and/or contamination levels [35]. 

Table 2-4. ICPP 601/602 process cell information (Cederberg et al. 

1974)

Cell Process description Cell function 

A EBR feed preparation Fuel dissolution

C MTR feed preparation Fuel dissolution

D MTR feed preparation Fuel dissolution
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 26 of 91

E STR-(S1W)feed preparation Fuel dissolution

F First-cycle extraction Uranium separation

G MTR feed preparation Fuel dissolution

H MTR first-cycle extraction Uranium separation

J Hot salvage

K Solvent recovery

L RaLa process cell Recovery of RaLa 

P First-cycle extraction

Q Second-cycle extraction

R Product transfer cell

S Third-cycle extraction

T Solvent pumps Hexone storage

U First-cycle aqueous raffinate treatment

V HP field office

W First-cycle solvent raffinate

X Sample dilution and decontamination

Y Second- and third-cycle raffinate treatment

Z Product room

The Laboratory Building, CPP 602, shares a common wall with CPP 601 and is 147 ft long, 102 ft

wide, and about 80 ft high (much underground). It is used to support activities in the process building. 

The final product was uranyl nitrate solution essentially free of impurities and fission products. It was 

shipped to the Y-12 Plant in 10-L polyethylene bottles in concentrations exceeding 250 g U/L (Lewis

et al. 2000). Beginning in 1971, a denitrator in a glovebox in CPP 602 was used to convert the uranyl 

nitrate to solid UO3 using a fluidized bed thermal conversion process.

In 1956, a process for the recovery of radioactive 140La (RaLa) was assigned to ICPP. The RaLa 

process took place in the complicated L Cell, and lasted into 1963. As freshly irradiated Materials 

Test Reactor (MTR) fuel was dissolved in acid, the dissolving process liberated gases, one of which

was 131I, with an 8-d half-life. The RaLa process recovered 140Ba for its 140La daughter product and 

shipped it to Los Alamos National Laboratory for use in weapon research projects [36]. 

Beginning in 1965, neptunium was collected from the second-cycle partitioning step. In 1972, this 

material was cleaned up using two cycles of hexone. Approximately 6.6 kg of neptunium was shipped 

to the Savannah River Site (SRS) for use as targets in making 238Pu in this demonstration project. 

The processing was done in CPP-601, and bottling of the product was done in the multicurie cell of 

CPP-627. The process did not separate the neptunium, so it remained a minor constituent of the 

product and waste [37]. 

When reprocessing was discontinued in 1992, these facilities were flushed to remove uranium and 

hazardous materials and placed in a standby condition. The Fuel Processing Restoration (FPR) 

Project would have replaced these facilities. FPR was about 40% complete when construction 

stopped in 1992. FPR was discontinued in a manner that preserved the facility for possible use in 

future missions at INTEC.

Three criticality events occurred during FPF operation:

• Criticality Accident of October 16, 1959 (Ginkel et al. 1960). A bank of storage cylinders 

containing dissolved spent EBR-1 fuel elements with enriched uranium was air-sparged (air 

was bubbled violently into the solution to mix it). The cylinders were geometrically safe, but 

the sparging initiated a siphon that transferred 200 L of the solution to a 5,000-gal tank 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 27 of 91

containing about 600 to 900 L of water. The resulting criticality lasted about 20 min. No 

workers were exposed to gamma or neutron radiation from the criticality because the criticality 

occurred in an unoccupied belowground cell. Airborne activity with some entrained liquid 

spread through the plant through vent lines and drain connections, triggering alarms and an 

evacuation. Pressure in the vessel with the criticality removed about 900 L of solution (76 L 

remained in the vessel) and unexplainably moved about 600 L into a companion vessel. A 

high radiation level (>25 R/hr) was discovered near the RaLa area above the waste tank

(Ginkel et al. 1960). Fields beyond the guardhouse measured 2 R/hr. The high fields were 

probably due to the large activity in the dissolved spent fuel outside its intended location. Two 

people who evacuated received internal exposures (< 30 mrem) as they passed the 

evacuation route area where radioactive gas was being released into the room from floor 

drains. Twelve evacuees received film badge doses from 2 to 50 rem (beta + gamma, mostly 

beta) (Ginkel et al. 1960). 

• Criticality Accident of January 25, 1961

•

. About 40 L of uranyl nitrate solution (200 g U/L) were

forced upward from a 5-in.-diameter section of an evaporator into a 24-in.-diameter vapor 

disengagement cylinder, well above normal solution level. Analysts later assumed that air 

entered associated lines while operators were attempting to clear a plugged line and improve 

pump function. When the air bubble reached the evaporator, solution was expelled from the 

lower section and a momentary criticality occurred in the upper section. Radiation triggered 

alarms, but no personnel received ≥100 mrem exposure (Stacy 2000). Concrete shielding 

walls surrounded the location of the criticality; the vent system prevented airborne activity from 

entering work areas; and equipment design prevented a persistent excursion (Paulus 1961). 

ICPP Criticality Accident of October 17, 1978. A criticality event occurred in the first-cycle 

tributylphosphate extraction system in the CPP-601 process building. The incident resulted in 

no personnel injury, no onsite or offsite contamination, and no damage to equipment or 

property, and the plume traveled over uninhabited areas to the southwest of the site (Casto

1980). The criticality caused approximately 3 × 1018 fissions of 235U. The Atmospheric 

Protection System at INTEC, which became operational in 1975, significantly reduced 

particulate emissions, and filtered all releases associated with the criticality event (INEL 1977, 

p 99). 

Internal exposure potential existed at the FPF and its support facilities from work with radioactive 

airborne particulates during maintenance activities, piping/valve changes, the criticality accidents, and 

stack releases [38]. 

External exposure

2.3.2

 occurred from work performed in Radiation Areas and High Radiation Areas

containing the nuclides referenced in Table 2-3. Brief exposures to measured levels of 50 rad/hr 

existed in the cells during piping/valve changes, decontamination, maintenance activities, working of 

production samples in the laboratories, etc. [39].

New fuels presented special problems in reprocessing. The Process Improvement Facility (PIF; CPP 

620 and CPP 637) and the Hot Pilot Plant (HPP, later Headend Processing Plant; CPP 640) were 

used to provide information to improve these processes.

Ancillary Facilities

The PIF includes a laboratory building with office space for the technical group. New ideas in 

reprocessing are developed and investigated at the laboratory bench scale. The laboratories were 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 28 of 91

designed to handle as much as 1 Ci of radioactive materials from Table 2-3 per laboratory. The waste 

stream discharged to the HPP waste system.

The HPP was used to test unproven equipment and systems. The facility consisted of five cells and 

associated utilities. The partition between two of the cells could be removed, making one large cell. 

The shielding around the cells was equivalent to that of the main plant and provided sufficient 

radiation shielding to run plant-level radioactive material. The graphite-based ROVER (nuclear 

rocket) fuels were processed beginning in 1983 for 14 mo in CPP 640 using two stages of burning to 

reduce the carbon content (Knecht et al. 1997). The ash was leached with a nitric/hydrofluoric acid 

mixture, extracted through three cycles of extraction, and converted to UO3. More than 100 kg of 

uranium were still in the ash at CPP after the burners were cleaned out in 1998.

The HPP waste system consists of three-level storage. High-level waste in storage can be routed to 

permanent storage tanks. Intermediate-level waste can be routed either to the waste evaporator 

system or the low-level waste tanks. Low-level waste is monitored and discharged to the disposal 

well downstream from the main service waste monitoring system. A service waste monitoring system 

is provided for this secondary stream. 

The Remote Analytical Facility in CPP-627 houses the RAL, the multicurie cell, a radiochemistry 

laboratory, and a decontamination facility.

The decontamination facility in CPP-627 provided support for cleaning tools and equipment for INTEC 

and other INL facilities. Items such as water pit gates, cooling pumps, vehicles, etc., were 

decontaminated in this facility. It was also used as a morgue and autopsy facility following the SL-1 

accident. Radiation levels to 25 rem/hr open-window beta-gamma were experienced for brief periods 

in the decontamination facility. Shielding other than temporary was not provided between the work 

stations there [40]. 

The RAL provides two rows of 15 analytical boxes behind a 9-in. iron shield wall (Stevenson and Lyon 

1955). The boxes can be remotely replaced to provide changed analytical capability. A sample 

transfer system below the boxes provided remote handling of samples brought to the boxes. The 

equipment is operated by hand-operated manipulators extending through the shielding, pneumatically, 

or with electronic controls. This facility has been used for remote examination of hazardous and 

radioactive materials to support INTEC operations. It continues to support INL and INTEC activities. 

The RAL handles, processes, analyzes, and experiments on hazardous and radioactive materials of 

all types. Samples collected in stations in the CPP-601 sample corridor were transferred to the RAL 

for analysis. Frequent samples during fuel reprocessing, taken in 10-mL sample bottles, had radiation 

readings of greater than or equal to 25 rad/hr. During the analytical process the samples were diluted 

to reduce personnel exposure [41]. 

The multicurie cell, which has walls 5 ft thick of barytes concrete, is designed to reduce the field from 

75,000 Ci of 1.6-MeV gamma emitter to 1 mR/hr (Boardman 1956). There is also a radiochemistry 

laboratory nearby to support operations in the multicurie cell. A walk-in hood in this area contained 

the custom processing facility. On February 9,1991, a small explosion destroyed the 6-in. outer 

diameter borosilicate glass dissolver section, contaminating four employees and a portion of the lab. 

Internal exposures ranged from 0.24 to 9.1 mrem/yr for 50 years. The unirradiated material came 

from a cleanup campaign at Argonne National Laboratory–East (ANL-E) and is suspected to have 

contained zirconium, which would react explosively with the nitric acid being used to dissolve the 

uranium (Decker 1991).
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 29 of 91

2.3.3

The original fuel storage facility at INTEC CPP-603 included a special fuel-storage building with three 

20-ft-deep storage pools for spent nuclear fuel. The facility is about one-third of a mile south of the 

main processing building. 

INTEC Fuel Storage Facility (CPP-603), 1950 to 2007 

Levels of airborne radioactivity of MFP around the CPP-603 unlined storage pools were a chronic 

problem from sodium-contaminated EBR-I fuel, which also led to contamination in the building. 

Efforts to clean up the water were aggravated by deionized water attacking the concrete pool. 

Anticontamination apparel was provided, but generally respiratory protection was not required. Air 

activity was routinely measured at 10 to 25% (Rich et al. 1974) of the radioactivity concentration 

guideline for soluble 90Sr.

In addition, CPP-603 contains the Irradiated Fuel Storage Facility (IFSF), a graphite fuel storage area, 

and a Fuel Cutting Facility. The IFSF stores dry fuel that is incompatible with underwater storage. 

The IFSF has 636 storage positions and is more than half full. Most of the spent fuel stored there 

came from the Fort St. Vrain commercial reactor in Colorado. Shipments from Fort St. Vrain ended in 

1991.

In the 1950s, the ICPP received a request to process spent fuel elements from SRS reactors. The 

14-ft elements were clad in aluminum and had to be cut to 18-in. lengths to fit in the dissolver vessel. 

Irradiation of the fuel changed metallurgical characteristics, so rather than cutting like regular 

aluminum they crumbled, necessitating development of a new technique and procedure. Equipment 

change and maintenance was extremely difficult because of the crumbling, which contributed to 

increased levels of exposure and contamination. As a result, complete modification of the process 

and equipment was required.

The pools in CPP-603 were built in 1950 and served as the primary spent fuel storage facility until 

1984. Fuel, once in underwater storage at Building 603, has been transferred to the newer 

underwater storage pools at the Fluorinel Dissolution Process and Fuel Storage Facility (FAST) or to 

dry storage. The facility is being evaluated for D&D. As part of this, the CPP-603 storage pools were 

cleaned of sludge, dewatered, and filled with grout in 2006 (Rockhold 2007). 

Internal exposure potential was greatest from fuel cutting and shearing, which created airborne 

radioactivity from the cut SRS fuel elements. A chronic exposure condition existed to MFPs from the 

pool water (Rich et al. 1974). 

External exposure

2.3.4

was received during fuel loading and unloading, movements to various locations in 

the storage pools, fuel cutting, cleanup activities, modifications, and D&D evaluations [42]. 

The HLLW Tank Farm includes 11 underground, stainless-steel, 300,000-gal storage tanks nested in 

concrete vaults east of process building CPP-601. The tanks were used to store radioactive liquid 

waste generated during the reprocessing of spent fuel and plant decontamination work. One tank 

was always kept empty for use as a transfer backup if a problem developed with one of the other 

tanks. All of the Tank Farm liquid has been calcined, reducing the volume and converting it to a more 

stable solid form. The underground tanks are encased in concrete vaults that have sumps and leak 

detection. The tanks are extremely corrosion-resistant. No leakage has been detected from the 

tanks. However, some leaks have occurred from transfer lines outside the tanks.

High Level Liquid Waste Underground Storage Tanks (Tank Farm), 1951 to 2007
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High-level waste at INL consists of acidic liquid and calcined solids. The acidic liquids have been 

stored in the underground tanks and included actual high-level waste as well as sodium-bearing 

waste that is managed as high-level waste. The stainless-steel tanks allowed the storage of waste in 

acidic form and resisted corrosion. When full, each tank contained only a few gallons of pure 

radioactive fission products. The rest of the solution was dissolved cladding-metal ions, process 

additives, and water. The tanks that received waste from the first-cycle extraction, which accumulated 

most of the fission products, had cooling systems to carry away decay heat to minimize corrosion. 

In the 1990s, a major effort of cleanup and repair in the valve boxes of the Tank Farm resulted in a 

large collective dose. This project, like many others, was preplanned and reviewed by the 

Westinghouse Idaho Nuclear Company As-Low-As-Reasonably-Achievable Committee. The workers 

wore TLDs and electronic dosimeters and/or self-reading pencil dosimeters. Although this dose was 

separately tracked, it is included in the cumulative dose records for those employees [43]. 

The HLLW Processing Facility, CPP-604, is east of CPP-601. Liquid waste generated from 

reprocessing activities was transferred to the Liquid Waste Evaporator, where the liquid was heated, 

reduced in volume, and stored in an underground tank. Soil, airborne, and ground-water 

contamination have resulted from these operations.

Part of the processing included a Rare Gas Processing Facility (CPP-604). Its purpose was to 

recover 85Kr from spent fuels. In 1958, the process was enhanced by replacing the liquid nitrogencooled carbon beds with a cryogenic distribution system. This gas product was shipped to Oak Ridge 

National Laboratory for commercial sale for use primarily in leak detection. INTEC was the only 

source of 85Kr outside the former Soviet Union (SCA, SRA, and WCC 1994).

Internal exposure potential existed from airborne radioactivity created during flushing operations, 

valve repairs, or other maintenance activities in contamination areas from MFPs and activation 

products [44]. 

External exposure

2.3.5

occurred during work in the contamination areas and on valve changes and piping 

maintenance where exposure rates measured as high as 500 rad/hr [45]. 

To remove liquid from the waste, the AEC developed a fluidized-bed calcination process, and built it 

at INTEC. Scientists at ANL tested the method in small-scale models in 1955. The process not only 

solidified the waste, but the product was granular, free-flowing, and easily handled by pneumatic 

transport techniques. Phillips engineers started designing the Waste Calcining Facility (WCF) in 

1956, and construction of WCF-1 started in 1958. The facility was constructed just east of the main 

INTEC process building and south of the storage tanks. Thick concrete shielding walls surrounded 

the process cells, which were below grade. The first campaign lasted until October 1964. Liquid 

waste was injected into a fluidized-bed chamber heated to 420 ̊C by a NaK heat exchanger system. 

Liquids evaporated and solids collected on the bed material, which was then collected in storage bins. 

Two 300,000-gal tanks and part of a third were emptied before the campaign was forced to stop 

because it had filled all available calcine bins. Half a million gallons of liquid had been transformed 

into 7,500 bulk cubic ft of solid waste. This was a reduction in volume of more than 9 to 1. The gases 

leaving the stack included 90Sr and 106Ru, but the levels were below guideline limits (AEC 1969). In 

1970, an in-bed oxygen atomized kerosene combustion system was installed raising the bed 

temperature to 500 ̊C, reducing wall temperatures and ruthenium concentrations in the off-gas.

Waste Calcining Facility 1 and New Waste Calcining Facility, December 1963 to 2002
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In 1982, the New Waste Calcining Facility (NWCF) replaced WCF-1. It converted liquid high-level 

radioactive waste from the Tank Farm into a granular solid similar in consistency to sand. The liquid 

waste was drawn from underground storage tanks and sprayed into a calciner vessel superheated by 

a mixture of kerosene and oxygen. The liquid evaporated while radioactive solids adhered to the 

granular bed material in the vessel. The off-gases were treated and monitored before release to the 

environment, and the residual calcine solids were transferred to large stainless-steel structures 

encased in thick concrete vaults called bin sets. The calciner was shut down in May 2000 while DOE 

evaluated whether to upgrade it to meet new emissions standards or develop a new technology to 

treat the remaining liquid in the Tank Farm. The calciner operated one last run in 2002 to eliminate 

the remaining HLLW in the 300,000-gal storage tanks. HLLW typically contained 300 Ci/m3

 [46]. 

To date, all HLLW has been removed from the Tank Farm and solidified through calcination. 

Removing the sodium-bearing waste remains one of the highest DOE priorities.

The NWCF was the location of a decontamination facility used for cleanup of radiologically 

contaminated materials from INTEC and occasionally from other INL processes [47]. The 

decontamination facility continues to function to support INL cleanup activities.

Internal exposure potential existed from releases of contamination to the occupational environment 

due to leaks from piping breaks, equipment failures, and other actions that permitted unplanned 

releases and from decontamination activities that would create airborne radioactivity [48]. 

External exposure

2.3.6

resulted from routine maintenance on the transfer piping and associated valves 

and equipment. Calcine process cell entries have been made for cleanup and maintenance activities 

with radiation fields measured to 50 rad/hr beta-gamma for brief periods. External exposure 

continues to accumulate from the decontamination facility [49]. 

FAST has two parts: a spent fuel storage area and the Fluorinel Dissolution Facility (FDF). The 

storage area consists of six stainless-steel-lined pools for storing spent nuclear fuel. The FDF 

includes a hot cell with 6-ft-thick concrete walls where spent fuel was dissolved in an acid solution. 

The Fluorinel process was used on Zircaloy-clad naval fuels and used three dissolver-complexer 

trains operating batchwise. Soluble neutron poisons and limiting the mass provided criticality control. 

When reprocessing ended in 1992, uranium and hazardous materials were flushed from the FDF, and 

this part of the facility was placed in a standby condition. About 1,546 kg of uranium were

reprocessed using the fluorinel dissolution process (Knecht et al. 1997). 

Fluorinel Dissolution Process and Fuel Storage Facility, 1984 to 2007

Internal exposure potential existed from airborne radioactivity that might have occurred from the 

processes associated with the FAST [50]. 

External exposure

2.3.7

occurred from the unloading and loading of irradiated fuel, from underwater fuel 

element examination, and from work in the FDF hot cell environment, which created radiation 

exposure as high as 50 rad/hr for brief periods [51].

The INL Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) 

Disposal Facility (ICDF) Complex is a new engineered facility south of INTEC and adjacent to the 

existing percolation ponds. This facility was planned to begin operation in July 2003 (INEEL 2003). It 

INL Comprehensive Environmental Response Compensation and Liability Act 

Disposal Facility Complex, July 2003 to 2007
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 32 of 91

is designed and authorized to accept wastes from INL CERCLA actions. The ICDF Complex includes 

the necessary subsystems and support facilities to provide a complete waste management system. 

The major components include disposal cells (landfill), an evaporation pond (consisting of two cells), 

and the Staging, Storage, Sizing, and Treatment Facility. The Complex covers approximately 

40 acres, with a landfill disposal capacity of approximately 510,000 yd3

. The evaporation pond is 

designated as a Resource Conservation and Recovery Act (RCRA) Corrective Action Management 

Unit and is the disposal site for leachate and other aqueous wastes generated as a result of operating 

the ICDF Complex. In addition, other aqueous wastes such as existing purge water can be disposed 

of in the evaporation pond in accordance with the ICDF evaporation pond Waste Acceptance Criteria.

Internal exposure from airborne radioactivity would exist if the integrity of the packaged material was

compromised during the handing and storage process [52]. 

External exposure 

2.3.8

occurs when shipments of radioactive materials consisting of MFPs and activation 

products are placed in storage at the facility. Normal radiation levels are not permitted to be greater 

than 200 mrem/hr at any edge of the transporting vehicle. Higher radiation fields might be permitted, 

under special conditions, by proper management authority [53]. 

The Independent Spent Fuel Storage Installation (ISFSI) is a new U.S. Nuclear Regulatory 

Commission (NRC)-licensed dry storage area for spent fuel and debris from the TMI accident. Fuel 

and debris were transferred to the INL TAN for examination, study, and storage after the accident. 

The fuel and debris were transferred to the ISFSI, which provides safe, environmentally secure, 

aboveground storage in metal casks inside concrete vaults. The transfer was completed in mid-2002.

TMI-2 Independent Spent Fuel Storage Installation, CPP-1774, 1998 to 2007

Internal exposure potential could exist from leaks or off-gassing from the storage containers [54]. 

External exposure 

2.4 ARGONNE NATIONAL LABORATORIES–WEST, FEBRUARY 1951 TO PRESENT

occurred during the unloading and placement of fuel in the storage vaults [55]. 

ANL-W was originally known as “the Idaho Division” of ANL. ANL-W is an extension of ANL-E near 

Chicago. ANL is a DOE research laboratory operated by the University of Chicago. The original INL 

site is in the southwest portion, approximately 18 mi via Highway 20/26 east of Arco, 40 mi via 

Highway 26 northwest of Blackfoot, or 50 mi west via Highway 20 from Idaho Falls. The original ANLW location is now a National Historic Landmark and is the site of the now-decommissioned EBR-I, 

Boiling Water Reactor (BWR) Experiment, Argonne Fast Source Reactor (AFSR), and Zero Power 

Reactor 3 (ZPR-III) facilities. 

The present ANL-W site is in the southeast portion of INL, about 35 mi west of Idaho Falls. There are 

52 major buildings at this site, including reactor buildings, laboratories, warehouses, technical and 

administrative support buildings, and craft shops. In February 2005, DOE reorganized; ANL-W 

became part of the Idaho National Laboratory (INL), and that portion of the site became the Materials 

and Fuels Complex (MFC).

2.4.1

EBR-I, the first reactor built at INL, was a NaK-cooled, solid-fuel (enriched uranium), unmoderated 

heterogeneous fast reactor designed for full-power operation at a level of 1 MW. It was built to 

explore the possibilities of breeding nuclear fuel and for the use of liquid metal cooling. A blanket of 

Experimental Breeder Reactor No. I, April 1951 to December 30, 1963
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 33 of 91

238U around the core provided the fertile material in which nuclear material breeding took place. 

Because the primary coolant was intensely radioactive during and shortly after operation, all primary 

components were enclosed in concrete-shielded cells. The secondary coolant, which was 

nonradioactive, required no shielding (Kittel, Novick, and Buchanan 1957). The facility was entirely 

inside a single building of brick, concrete, and steel. Construction on EBR-I began in May 1949 and 

was complete in April 1951. Reactor startup occurred on August 24, 1951.

On November 29, 1955, the reactor suffered a 40% to 50% core meltdown. Radiation detection 

instrumentation measured radioactivity in the building above normal background levels, and all 

personnel were evacuated. After the partial meltdown, the core assembly was removed from the 

reactor using a temporary cave on the reactor top and shipped to ANL- E (Kittel, Novick, and 

Buchanan 1957). The core was replaced and the reactor remained operational until December 30, 

1963. On August 26, 1966, EBR-I was dedicated as a National Historic Landmark. The principal 

radiological activity associated with the coolant during operation and shortly after shutdown was 24Na, 

τ1/2 = 15 hours. The saturation level at full-power operation was approximately 24 μCi/g. The second 

most significant activity was about 2 μCi/g of 137Cs, which apparently entered the system during and 

after the meltdown incident. No other long-lived activity was identified in the primary coolant. Shortlived activity in the form of 133Xe and 135Xe was observed in the cover gas (Haroldson et al. 1963).

Internal exposure potential existed from airborne radioactivity from mixed fission products and 

activation products [56]. 

External exposure

2.4.2

was received by personnel from MFPs and activation products during activities 

associated with reactor operation and maintenance [57]. 

The date information in this discussion is from Stacy (2000, p. 261). BORAX-I was an open-top, 

water-cooled, water-moderated, BWR used to conduct a series of nondestructive experiments in the 

latter part of 1953 and early summer of 1954 (Dietrich 1956). The reactor was built in an excavated 

area of earth approximately one-half mile northwest of EBR-I and was housed in a 10-ft-diameter tank 

open to the atmosphere. The control room was approximately one-half mile away near the EBR-I 

reactor. 

Boiling Water Reactor Experiment No. 1, Late 1953 to July 22, 1954

BORAX-I was intentionally destroyed in its final experiment on July 22, 1954. The explosion scattered 

fuel plate fragments and other debris over an area of approximately 200 ft by 350 ft south of the 

reactor area fence. Instrumentation at the control center showed an instantaneous radiation level 

higher than 500 mrem/hr, which decreased in about 30 s to 25 mrem/hr and within 5 min to less than 

1 mrem/hr. A detailed discussion of this incident is available (Griffiths, Sill, and Wilhelmsen 1956; 

Brodsky and Beard 1960).

Internal exposure might have occurred from airborne radioactivity during operation and other activities 

associated with a BWR, the core destruction, coolant, and cleanup activities [58]. 

External exposure 

2.4.3

occurred from direct radiation associated with the reactor operation maintenance 

activities from MFPs and activation products [59]. 

The date information in this discussion is from Stacy (2000, p. 261). BORAX-II was designed and 

built to replace BORAX-I to investigate a new reactor that would more closely approximate the 

Boiling Water Reactor Experiment No. 2, October 19, 1954, to March 1955
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characteristics of a practical power reactor operating on the boiling-water principle (Dietrich 1956). 

BORAX-II, a short distance northeast of the BORAX-I site, was built in 1954 and became operational 

on October 19, 1954. BORAX-II was significantly larger than BORAX-I. The vessel was shielded by 

concrete and housed in a sheet-metal building. Tests of new core combinations were tried using 

varying enrichments of 235U in metal fuel plates (AEC 1962). It was a boiling-water system operating 

at 300 psi, making it essentially a power experiment. The power level was about 6.4 MW (t) but, 

because it had no turbine generator, it produced no electricity. The energy produced was released in 

the form of steam (Dietrich 1956). In 1955, a turbine generator was added to BORAX-II and testing 

was done to demonstrate that turbine contamination would not be a significant problem in BWRs. 

Internal exposure was possible from MAPs associated with work activities from the reactor coolant

[60]. 

External exposure

2.4.4

occurred during routine operations and with loading and unloading of the reactor

[61]. 

The date information in this discussion is from Stacy (2000, p. 261). As a result of the BORAX-I and -

II tests, a program began in March 1955 to modify the BORAX-II reactor to produce electricity. The 

modified reactor became BORAX- III. The previous BORAX reactors were not designed to produce 

electricity, so a turbine generator was added to the facility to convert thermal energy to electricity. 

The modified facility was capable of generating 12 MW of thermal energy and 2,300 kW of electrical 

energy. For 2 hr on July 17, 1955, BORAX-III generated approximately 2,000 kW of electricity; 500 

kW were used to power the BORAX-III facility, 1,000 kW were used to power the CFA, and 500 kW 

were used to light the entire town of Arco, Idaho. BORAX-III became the first nuclear power plant in 

the world to generate electricity for an entire city. It was operational from June 9, 1955, until 

sometime before December 3, 1956. 

Boiling Water Reactor Experiment No. 3, June 9, 1955, to December 1956

Internal exposure was possible from MAPs associated with work activities from the reactor coolant

[62]. Fuel cladding failure occurred, and the resulting radionuclides were identified in the reactor 

water (Zinn et al. 1956).

External exposure

2.4.5

occurred from routine operation and loading and unloading of the reactor [63]. 

High radiation levels (up to 430 mR/hr) were reported associated with the steam system. Decay 

curves were measured for the steam, condensate, and reactor water (about 1 × 104 higher activity) 

(Zinn et al. 1956). Nitrogen-16 was identified as the principle source in the coolant (Dietrich, 

Lichtenberger, and Zinn 1956). 

BORAX-IV, the successor to BORAX-III, began operation in December 1956. This reactor, with a 

design power of 20 MW (t), was used principally to test high-thermal-capacity fuel elements made 

from ceramics of uranium and thorium (Handwerk, Hoenig, and Lied 1957). Like the previous BORAX 

reactors, BORAX-IV was a BWR operating at 300 psig. It was capable of producing 2.5 MW of 

electricity. It was brought to criticality on December 3, 1956, at atmosphere pressure. It operated 

with a core of uranium-thorium fuel elements until April 17, 1957. Beginning in May 1957, it was 

operated with a 59-element core at 300 psig and 216oC and continued intermittent operations until 

December 5, 1957. After the core was revised to increase the maximum power, the reactor was 

restarted on February 19, 1958, to evaluate the effect of operating with a fuel element defect and to 

locate defective elements in the core (Robertson and Hall 1959). BORAX-IV released approximately 

Boiling Water Reactor Experiment No. 4, December 3, 1956, to June 1958
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4,565 Ci of short-lived radionuclides to the atmosphere in March 1958 (Novick 1958). It operated until 

June 1958. The following measurements were made during reactor operations (Robertson and Hall 

1959): 

1. Radiation levels of the steam plant equipment

2. Quantitative determination of fission gases 138Xe and 88Kr, which were released to the 

atmosphere through the air ejector

3. Analysis of reactor water, condensed steam before the turbine, and condensed steam after the 

turbine (hot-well condensate) for fission products

4. Area contamination downwind from the reactor

Internal exposure might have occurred during work with the defective fuel elements or during planned 

releases of short-lived fission activity and from airborne MAPs/MFPs associated with the reactor 

coolant [64]. 

External exposure

2.4.6

occurred during operation and work with loading and unloading of the reactor [65]. 

The date information in this discussion is from Stacy (2000, p. 261). BORAX-V was a flexible BWR

with the same configuration as BORAX-IV, used primarily for testing nuclear superheating concepts. 

The facility was operational from February 9, 1962, until September 1964. 

Boiling Water Reactor Experiment No. 5, February 9, 1962, to September 1964

Internal exposure might have occurred from coolant and airborne activity during routine BWR 

operation with fuel elements made of ceramics of uranium and thorium, and associated maintenance 

work [66]. 

External exposure

2.4.7

occurred from routine activities associated with reactor operation and maintenance

[67]. 

The date information in this discussion is from Stacy (2000, p. 268). ZPR-III was a low-power, splittable reactor that achieved criticality by bringing two halves of a fuel configuration together. It was 

used to determine the accuracy of predicted mass geometries and critical measurements for fast 

reactor core designs.

Zero Power Reactor No. 3, October 1955 to November 1970

The building consisted of a reinforced concrete high bay assembly room and a one-story section 

containing the control room, work room, vault, laboratory rooms, offices, etc. The assembly (reactor) 

room of reinforced concrete was approximately 45 ft by 42 ft by 29 ft high (Brittan et al. 1961).

The assembly machine was a platform on which two tables or carriages were mounted, one of which 

was moveable. Half of the reactor was built up on each carriage by inserting drawers containing the 

reactor material into a matrix structure. Each half of the assembly contained five safety control rods 

and a 15-Ci polonium-beryllium neutron source. A hinged platform could be swung into place 

between the halves on which workers could stand while loading or unloading the machine. 
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The storage vault room was approximately 29 ft long by 26 ft wide with walls and roof of reinforced 

concrete 9 in. thick. The fuel slugs were stored either on racks or in special “birdcage”-type 

containers that limited the storage density to 2 kg of 235U per cubic foot (Brittan et al. 1961). 

Internal exposure potential existed from possible airborne radioactivity during maintenance operations 

from MFPs/MAPs [68]. 

External exposure

2.4.8

occurred during maintenance activities and during loading and unloading of the 

reactor fuel [69]. 

AFSR was a small fast reactor facility designed to produce neutrons for the development of special 

equipment for the fast reactor programs of EBR-I, EBR-II, and ZPR-III. The reactor, with a design 

power of 1 kW, was in a prefabricated Butler-type building with its own heating and air compressor 

plant built in 1958 near the EBR-I facility. No water was plumbed into the building. Control and safety 

mechanisms were in a pit below the reactor. The reactor, designed to supply both fast and thermal 

neutron fluxes for laboratory experiment, was built around a cylindrical core of solid HEU with a 

blanket of solid DU (Brunson 1959). Reactor startup occurred on October 29, 1959; the reactor was 

operational until sometime in the late 1970s when it was moved to a new location adjacent to the 

ZPPR facility at the ANL-W site. The reactor is now shut down and defueled.

Argonne Fast Source Reactor, October 29, 1959, to Late 1970

Internal exposure might have occurred during routine operations that could create airborne 

radioactivity [70]. 

External exposure 

2.4.9

occurred from maintenance activities and loading and unloading of fuel [71]. 

TREAT was an air-cooled thermal heterogeneous system designed to evaluate reactor fuels and 

other material under conditions simulating various types of reactor excursions. Construction began in 

February 1958 and ended in November 1958, and criticality was achieved on February 23, 1959. The 

TREAT complex consists of a reactor building and a control building approximately 1 mi northwest of 

the EBR-II containment building (Freund et al. 1960).

Transient Reactor Test Facility, February 23, 1958, to April 1994

The reactor building features a high bay section and an adjacent service wing. The high bay section 

contains the reactor, fuel storage pit, instrument room, and the basement subreactor and equipment 

rooms. The control building is a single-story concrete block structure approximately one-half mile 

from the reactor that contains control panels and necessary instrumentation for remote control of the 

reactor.

Shielding permitted personnel access around and on top of the reactor during steady-state operations 

at 100 kW. Access to the subreactor room was controlled during steady-state operation. Before

transient operations, the building was evacuated. General neutron and gamma radiation levels 10 ft

from the reactor during operations at 100 kW were (Freund et al. 1960):

• Fast neutron Negligible

• Thermal neutrons 50–1,500 n/cm2

/s 

• Gamma 5–8 mrem/hr
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Internal exposure might have occurred during routine operations that could create airborne 

radioactivity; however, it was not expected to occur [72]. 

External exposure

2.4.10

occurred from routine operations [73]. 

The date information in this discussion is from Stacy (2000, p. 262). EBR-II, at the ANL-W site, is a 

liquid sodium-cooled, unmoderated, heterogeneous fast breeder reactor rated at 62.5 MW (t), with an 

intermediate closed loop of secondary sodium and a steam plant capable of producing electric power 

through a conventional turbine generator. A fuel processing facility is attached to the reactor. EBR-II 

was designed to prove the breeding of fuels, the feasibility of a central power station, and onsite fuel 

processing. These objectives were met in the late 1960s, and the role of EBR-II changed to test 

reactor. Construction of EBR-II ended in May 1961, and the reactor reached criticality on September 

30, 1961. It operated until September 30, 1994, when it was taken to a subcritical configuration and 

shut down to start a defueling operation. On January 19, 2001, ANL-W verified that the liquid-metal 

sodium coolant had been completely drained from the reactor vessel. At present, the reactor is 

defueled, the sodium systems have been drained, and the power plant is depressurized.

Experimental Breeder Reactor No. II, May 1961 to September 30, 1994

The reactor was submerged in a primary tank containing approximately 90,000 gal of liquid sodium. 

This tank was suspended in an airtight steel-shell containment building of 1-in.-thick steel plate, which 

would contain an accidental release of fission products, etc., from the primary system. The structure 

of the primary system is designed to contain the energy release associated with a reactor incident. 

The reactor building is designed to confine the effects of a maximum sodium-air interaction caused by 

a major sodium release. The reactor consists of an enriched core surrounded on all sides by a fertile 

blanket of depleted uranium (McVean et al. 1962; Koch et al. 1957). 

The Sodium Plant contains the pumping, purification, and storage facilities for the secondary sodium 

system. It also contains a receiving station for the sodium. The building was not normally occupied. 

The primary and secondary coolant from EBR-II is converted in the Sodium Processing Facility from 

its elemental, chemically unstable form, to a chemically stable composition suitable for landfill 

disposal.

The Fuel Manufacturing Facility (FMF) is a secure facility designed for the fabrication of EBR-II fuel. 

The FMF vault stores special nuclear material in support of the EBR-II shutdown.

An additional building, the Laboratory and Office Building near the EBR-II plant, provided supporting 

analytical and personnel facilities. 

2.4.11

The Hot Fuel Examination Facility (HFEF) complex comprises two hot cell facilities, HFEF/South and 

HFEF/North. HFEF/South, originally known as the Fuel Cycle Facility and/or the Fuel Conditioning 

Facility, was used to demonstrate remote processing and refabrication of uranium-fission metal-alloy 

driver fuel elements in a closed cycle with EBR-II. 

Hot Fuel Examination Facility, 1964 to 2007

Some 35,000 fuel elements were remotely reprocessed and refabricated into EBR-II subassemblies 

between 1964 and 1968. HFEF/South contains two large heavily shielded hot cells, one with an inert 

gas (argon) atmosphere, the other with an air atmosphere. The shielding walls of both cells are of 

high-density concrete. The HFEF/South air cell was decontaminated and refurbished in 1969 and 

again in 1976. 
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HFEF/North is a large alpha-gamma hot cell facility that was activated in March 1975. This facility 

provided the capability for postirradiation and nondestructive or destructive examination of fuel and 

material experiments irradiated in EBR-II. HFEF/North contains two hot cells, one with an argon gas 

atmosphere and the other with an air atmosphere. The air atmosphere cell was known as the 

decontamination cell. The shielding walls of both cells are of high-density concrete (Baca 1979).

HFEF began operation as a fully automated facility for examining highly radioactive experimental 

reactor fuel elements and other components in 1975. The examinations conducted in HFEF provide 

data that are essential for determining the performance and condition of fuels and materials irradiated 

in DOE reactor facilities. HFEF continues in operation as a vital component of the DOE energy

research program.

Remote characterization of material to be shipped to the Waste Isolation Pilot Plant (WIPP) in New 

Mexico for disposal takes place in the Waste Characterization Area of the HFEF high bay.

Internal exposures might have occurred during cell entries, when suspended radioactive 

contamination materials could cause airborne radioactivity from MFPs and activation products [74]. 

External exposure

2.4.12

occurred when entries to the hot cell were made after experiment processes or 

during equipment maintenance and refurbishment [75]. 

The date information in this discussion is from Stacy (2000, p. 268). ZPPR is a split-table critical 

facility approximately 300 m from EBR-II in the ANL-W area and about 3 mi north of U.S. Highway 20. 

The facility is divided into two areas, the mound area and the support wing. The mound area consists 

of the reactor cell, fuel storage vault, workroom, and equipment rooms as well as access and escape 

tunnels. The reactor cell is a 50-ft-diameter circular room with floor and walls of reinforced concrete. 

The roof consists of layers of washed and dried sand and gravel supported by a catenary cable 

network.

Zero Power Physics (Plutonium) Reactor, April 18, 1969 to April 1992 (Standby)

The basic element of the ZPPR is a bed-and-table system, which holds the matrix assembly. The two 

tables, one moveable and one stationary, are supported on a cast-steel bed. Neutron fields from the 

plutonium fuel are present between the two halves when the reactor was off and open (Simons, 

Young, and Thalgott 1972). The control and safety rod drives are mounted near the rear of each 

table. The main floor consists of the reactor control room, offices, an electronics shop, and a core 

coating room. The core coating room, adjacent to the control room and the entrance to the mound 

area, is used to clean core stimulants such as 235U and stainless steel. The room contains two hoods 

for handling suspect materials and a core coating machine that is used primarily to dry and coat 

depleted uranium with a protective film.

Internal exposure potential was minimal due to the use of hoods and other protective equipment [76]. 

External exposure

2.4.13

occurred from working with reactor processes, loading and unloading fuel, etc. [77]. 

The Neutron Radiography (NRAD) Facility is a 250-kW, steady-state Training, Research, Isotope 

General Atomics (TRIGA) reactor in the basement beneath the HFEF/North main cell. The reactor 

core consists of fuel elements surrounded by graphite assemblies. The core is submerged in a waterfilled tank. NRAD began operation on October 12, 1977, with two radiography stations. The East 

station services the hot cell complex where specimens can be radiographed without removing them 

Neutron Radiography Facility, October 1, 1977 to 2006
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from the hot cell environment. The North station is outside the cell in a separate, clean, shielded 

location for the radiography of irradiated or unirradiated items without introducing them into the 

contaminated cell. Cask handling and specimen shielding allow for full-size reactor assemblies. The 

radiography room is easily accessible for development work (Richards and McClellan 1979).

NRAD has limited irradiation capabilities in the core. It has a water-filled port at the center of the core 

and a dry port at the edge of the core. NRAD operates a MF Physics linear particle accelerator that is 

used for nondestructive assays of waste and expended nuclear fuel.

Internal exposure potential exists from possible airborne radioactivity, primarily from the hot cell 

environment from MFPs and activation products [78]. 

External exposure

2.4.14

occurred during sample handling and maintenance associated with radioactive 

samples. Remote handling techniques are used to minimize dose [79]. 

The Fuel Assembly and Storage Building (FASB) is a multipurpose facility that supports development 

of low-enrichment uranium fuel for research reactors, storage of spent fuel, and examination of the 

condition of other experimental projects. The East (clean) room houses offices, restrooms, etc. The 

West room contains a vault for the storage of nuclear material. It also contains equipment for 

performing materials testing and for preparing metallurgical samples, and inert atmosphere 

gloveboxes and hoods. The facility no longer does fuel assembly, but other radiological work is 

ongoing.

Fuel Assembly and Storage Building, 1970 to 2007

Internal exposure might occur from airborne radioactivity associated with the described processes 

from the uranium fuel and spent fuel examinations [80]. 

External exposure

2.4.15

occurred from the movement of radiological samples and reactor fuel [81]. 

At ANL-W, an Analytical Laboratory provides the capability for performing chemical and physical 

measurements of radioactive and nonradioactive samples. The facility includes six analytical hot cells 

and general and specialized chemistry laboratories. Personnel were subjected to radiation levels 

above building background on occasion in the Junior Cave Area. 

Other Argonne National Laboratory–West Support Facilities 

The Radioactive Scrap and Waste Facility provides in-ground retrievable dry storage for nuclear fuels 

and other highly radioactive scrap and waste, and interim storage for EBR-II spent fuel.

The Radioactive Liquid Waste Treatment Facility evaporates low-level radioactive liquid waste 

generated at ANL-W facilities into solidified residue packaged in shielded containers. 

The Industrial Waste Pond (IWP) is an unlined evaporative seepage pond that is fed by a system of 

drainage ditches. It has been used since 1964 to receive wastewater from a number of sources. The 

largest sources of liquid industrial waste going to the IWP are blowdown effluents from the main and 

auxiliary cooling towers, auxiliary boilers blowdown; water from once-through air conditioning, and 

cooling water from other sources. There might be inadvertent low-level radioactive contamination in 

this pond.
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The three sanitary Sewage Treatment Ponds (STPs) north of EBR-II cover an area of about 2 acres. 

These ponds are not suspected of containing radiological contamination. 

Internal exposure is possible from airborne radioactivity associated with the types of samples worked 

in these facilities [82]. 

External exposure

2.5 RADIOACTIVE WASTE MANAGEMENT COMPLEX, MAY 1952 TO 2007

occurred from working with the variety of radioactive materials associated with 

these facilities (MFPs, activation products, and TRU materials) [83]. 

The RWMC is 51 mi west of Idaho Falls. The first 13 acres were fenced in May 1952 for shallow-land 

disposal of solid low-level radioactive waste, and burial of TRU waste and hazardous substances 

such as organic and inorganic chemicals. The current RWMC mission includes interim storage of 

TRU waste and shipment of stored TRU waste to the WIPP near Carlsbad, New Mexico, for

permanent disposal.

For the first 2 years, only low-level radioactive waste was buried at the RWMC. In 1954, the Rocky 

Flats Plant (RFP) in Colorado began shipping defense waste with TRU elements. By 1957, the 

original 13 acres were nearly filled, and the RWMC was expanded to 97 acres. In 1970, it was 

expanded again to 168 acres and is currently 177 acres. After 1970, TRU waste was placed in 

retrievable storage on asphalt pads and covered with an earthen berm. This waste is stored in drums 

and boxes in engineered modules. From 1970 to the present, low-level waste has been disposed of 

in 20 pits, 58 trenches, and 21 soil vault rows. INL has been repackaging (as needed) and shipping 

TRU waste to the WIPP.

Early packaging configurations were thought to be not suitable for extended storage, and could 

present future hazards to the workers, the public, or the environment. Early wastes were pushed out 

of trucks into open pits or trenches and covered with soil by heavy equipment, which might have 

occasionally damaged containers in the covering process. Long-tongued dump trailers were used to 

minimize exposure to personnel from the waste. Other types of heavy equipment such as cranes and 

cherry pickers were used to pick shielded containers from trucks for unloading the contents into soil 

vaults, etc. There were occasional brief exposures greater than 100 mrem/hr (photon energy ≥ 

250 keV).

In addition to administrative buildings, the RWMC has the facilities and processes described in the 

following sections.

2.5.1

The 97-acre Subsurface Disposal Area is in the western section of the RWMC. It contains an active 

shallow-land burial area for the permanent disposal of solid low-level waste. It also contains pits and 

trenches where mixed TRU and low-level waste was buried between 1954 and 1970. Solid waste 

from RFP comprised a large fraction of the waste received at the RWMC. For example, in 1969 

approximately 250,000 ft3 of waste from RFP with a reported activity of more than 35,000 Ci was

buried at the RWMC. RFP waste was usually contaminated with plutonium isotopes and 241Am (Till et 

al. 2002, pp. 26–27).

Subsurface Disposal Area 
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2.5.2

The Intermediate Level Transuranic Storage Facility has had 53 drums of 233U stored in metal cargo 

containers in an open yard surrounded by concrete block shielding. The facility also has several inground shielded storage wells that are used to store highly radioactive materials.

Intermediate Level Transuranic Storage Facility

2.5.3

The 56-acre Transuranic Storage Area (TSA) is in the southern section of the RWMC, dedicated to 

storage of contact-and remote-handled packages of solid TRU waste. This waste was received at INL

after 1970 and was stored aboveground.

Transuranic Storage Area

2.5.4

The Stored Waste Examination Pilot Plant (SWEPP) certifies waste to ensure that it meets repository 

acceptance criteria. Examinations used a shielded 450-kVp X-ray facility and a neutron assay system 

that developed 14-MeV neutrons. Certified waste has been stored temporarily in permitted storage 

areas until it is packaged and sent to the WIPP. Uncertified waste is segregated until processing is 

available that will enable it to meet acceptance criteria. 

Stored Waste Examination Pilot Plant

2.5.5

This loading station is used to load TRU waste into Transuranic Package Transporter-II shipping 

containers for shipment to the WIPP. Responsibility for the station has been turned over to British 

Nuclear Fuels Limited (BNFL), Inc.

Transuranic Package Transporter Loading Station

2.5.6

BNFL operates the Advanced Mixed Waste Treatment Project under contract with DOE. The facility, 

which is under construction, will retrieve and process approximately 65,000 m3 of mixed TRU waste in 

temporary storage at the TSA, treat the waste to meet environmental laws and disposal criteria, and 

package it for shipment to the WIPP.

Advanced Mixed Waste Treatment Project 

The exposures below represent all subsections of Section 2.5. 

Internal exposure might have resulted when workers were close to breached and or leaking waste 

containers containing MFPs, TRU materials, activation products, etc., from the materials in storage at 

the RWMC [84]. 

External exposure

Radionuclides at the RWMC cross the spectrum based on the character of operations at this facility. 

Those listed in Table 2-3 would represent many of the long-lived fission products. Activation products 

from reactor facilities were also a concern for external exposure.

might have resulted from working adjacent to the waste containers in storage or 

disposal in the form of intermediate-level, low-level, TRU, and mixed waste, and from SWEPP X-ray 

(30 to 450 keV) and neutron (2 to 20 MeV) waste package examination and certification processes. 

Approximately 150 6M drums containing as much as 500 g of 232/233U per drum have been stored 

under earthen covers on pads in the TSA (232U is always present as a contaminant with 233U and 

accounts for a significant in-growth of high-energy gamma emitters). Depending on the waste 

involved, radiation exposure levels might have been near background levels, or could have exceeded 

a few hundred millirem per hour with photon energies greater than 250 keV for short periods [85]. 
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2.6 CENTRAL FACILITIES AREA, AUGUST 2, 1943, 2007

The CFA is the main service and support center for INL programs and the other technical areas on 

the 890-mi2 site. It is about 50 mi west of Idaho Falls, Idaho, just north of Highway 20 in the southcentral portion of the site. CFA buildings and activities support transportation, maintenance, capital 

construction, environmental and radiological monitoring, security, fire protection, warehouses, 

calibration laboratories, and a cafeteria. There is a small amount of research and development (R&D) 

work. What is now the INL started as an offshoot of the Naval Proving Ground command area 

(dedicated August 2, 1943), where the Navy tested ordnance from fighting ships. Buildings 

constructed by the Navy became the staging area for INL development that began in earnest in 1950. 

The area continued to expand as a central service area for the NRTS. Eventually it acquired the 

name “Central,” or officially CFA. 

In the late 1950s and into the 1960s, small amounts of radioactivity were processed through a 

Sewage Treatment Plant, CF-674, to a drying pond. Most of the radioactivity was from the “hot” 

laundry, although small amounts could enter from CF-656 and CF-690. CF-656 was a Reactor 

Engineering Laboratory where tracer-level radionuclide and chemistry work occurred. CF-690 

included the laboratory where analytical chemistry was done on bioassay samples, naturally occurring 

radionuclides, and other special projects. The dispensary included an X-ray unit for medical use (30

to 250 keV) [86]. The exposures discussed below are for all subsections of Section 2.6.

Internal exposure. CFA internal exposure potential was primarily from MFPs, activation products, and 

TRU materials associated with articles cleaned at the laundry facilities. Laundry facility workers were 

included in bioassay programs [87]. 

External exposure

2.6.1

. CFA external exposure potential is greatest from calibration sources and X-ray 

equipment at the Health Physics Instrument Laboratory (HPIL) and the DOELAP Irradiation Facility 

with photon energies greater than 250 keV from calibration sources and 30 to 250 keV for X-ray 

photons. Neutron energies range from 2 to 20 MeV from the 252Cf source and the AmBe source. 

Personnel working in radiological areas must wear dosimetry devices [88]. 

The laundry, in the east portion of the CFA, washed coveralls and other protective clothing items used 

for radiological work. The laundry drain went to a septic tank and drain field with other sanitary waste. 

The laundry facility and drain field(s) are sources for low-level radioactive contamination, which covers 

the spectrum inherent to work in radiological contamination areas. The old laundry facility (CF-699) 

was used from 1950 and was demolished in 1994. The new laundry facility (CF-617) was used from 

1981 to 2001 and demolished in 2002 (Rockhold 2007).

Hot Laundry 

2.6.2

The HPIL, CFA-633, was a calibration facility used for radiological instrumentation standardization. 

The HPIL used 252Cf neutron sources, and alpha, beta, and gamma sources for HP instrument 

calibrations. All the sources were sealed [89]. A new facility, CF-1618, completed in late 2002, 

includes six automated irradiator systems and provides expanded neutron, gamma, and X-ray 

irradiation capabilities. The higher level sources require external exposure control, and personnel in 

the radiological work area must wear applicable dosimetry [90]. 

Health Physics Instrument Laboratory
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2.6.3

The DOELAP Irradiation Facility, CF-636, is an aboveground shielded bunker that houses radioactive 

sources used by DOE-ID for radiation measurement equipment calibrations. The bunker is on the 

access road east of the main road into the CFA. It houses an X-ray facility, seven 1-Ci 241Am sources, 

beta sources, and two 137Cs sources (20 and 1,000 Ci). At one time, there was an AmBe neutron 

source at this facility [91]. 

DOE Laboratory Accreditation Procedure Irradiation Facility

2.6.4

The RESL facility, operated by DOE-ID in CF-690, evaluates low-level environmental and other 

laboratory samples that pose minimal radiological risk from internal or external pathways [92]. The 

dosimetry facility, also in CF-690, has been used to process external dosimetry devices such as film, 

TLDs, etc. 

Radiological and Environmental Sciences Laboratory

2.6.5

CF-674 was used from 1953 to 1969 as a Chemical Engineering Laboratory to conduct calcine 

experiments on simulated nuclear waste. The experiments created liquid waste streams discharged 

to a pond designated as CFA-04. This waste stream was contaminated with calcine that contained 

low-level radioactive waste (DOE 2003).

CF-674 Building

2.7 TEST REACTOR AREA, MARCH 31, 1952, TO 2007

The TRA is approximately 5 mi north of the CFA. Eight reactors have been built and operated in the 

TRA. Three of the reactors – Materials Test Reactor, Engineering Test Reactor (ETR), and Advanced 

Test Reactor (ATR) – were high-flux reactors designed for materials testing. The remaining five – 

Reactivity Measurement Facility (RMF), Advanced Reactivity Measurement Facility No. 1 (ARMF-1), 

ARMF-2, Engineering Test Reactor Critical (ETRC), and Advanced Test Reactor Critical (ATRC) – 

were low-power reactors designed to perform reactivity measurements. At present, only the ATR and 

ATRC are operational.

Other TRA facilities of radiological concern are the TRA Hot Cells, Gamma Facility, Radiation 

Measurements Laboratory, Radiological Chemistry Laboratory, Liquid Waste Disposal Ponds, and 

High-Level Liquid Waste Disposal Tanks and Transfer Facility. When DOE merged INL and ANL-W 

in 2005, TRA became the Reactor Technologies Complex.

All personnel entering the TRA must wear a dosimetry badge, and those who work in or near

radiological control areas must wear PICs. Personnel working in radiological control areas are on a 

routine bioassay program and receive routine whole-body counts.

2.7.1

The date information in this discussion is from Stacy (2000, p. 264). The MTR (TRA 603) was the 

original reactor at the TRA and the second reactor to be operated at INL. Fueled with enriched 

uranium fuel, water-cooled and water-moderated, this reactor was a key part of the AEC postwar

reactor development program. It operated at a power level of 30 MW until September 1955, when 

thermal output was increased to 40 MW. It supplied a high neutron flux in support of a reactor 

development program subjecting potential reactor fuels and structural materials to irradiation.

Materials Test Reactor, March 31, 1952, to April 23, 1970
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The MTR gave the researcher several options to achieve sample irradiation. Lead experiments 

entered from the top of the reactor with positions around the core. Pneumatic ports on the reactor top 

enabled the insertion of capsules for irradiation in the graphite region around the core. A hydraulic 

rabbit system underneath the reactor enabled the insertion of specimens and their discharge to the 

canal during reactor operation. In addition, horizontal and angular beam holes made it possible to 

perform cross-section measurements and other physics research, including several neutron 

experiments. The high-flux radiation fields available in this reactor made it possible to accelerate the 

screening of test materials. In its early years, the MTR contributed to the design of pressurized-water, 

organic-moderated, liquid-metal-cooled, and other reactors. Its successful operation resulted in a 

family of plate-type reactor fuels.

The MTR logged more than 125,000 operating hours and more than 19,000 neutron irradiations. 

During August 1958, it became the first reactor to operate using 239Pu as fuel at power levels as high 

as 30 MW. In early 1970, the MTR was again fueled with 239Pu. The last core was named "Phoenix" 

after the legendary bird that lived 500 years, burned itself to ashes, then rose to live again. The 

plutonium cores demonstrated that a plutonium-fueled, water-moderated reactor could be controlled 

satisfactorily. In August 1970, the MTR was again brought to power for a 24-hr run to irradiate 1,000 

biological samples for iodine analysis [93]. 

Internal exposure was most probable during the first few hours of shutdown. When the reactor top 

was removed, airborne fission products would be released. During shutdown, airborne radioactivity of 

MFPs and activation products from maintenance activities resulted in the potential for internal 

exposure. Some experiments in loops resulted in releases, particularly of activation products [94]. 

External exposure

2.7.2

occurred during the operations associated with sampling a test reactor and the 

associated maintenance activities. The major contributors to external exposure were MFPs and 

MAPs that emitted beta and gamma radiation with energies typically above 250 keV [95]. 

The date information in this discussion is from Stacy (2000, p. 262). When the 175-MW ETR started 

in 1957, it was the largest and most advanced nuclear materials test reactor in the world. It provided 

larger test spaces than the MTR and a more intense neutron flux. ETR fuel, coolant, and moderator 

materials were evaluated under environments similar to those of power reactors. Several 

experimental loop facilities were designed to test the fuels for the ANP Program and the Navy fuel 

development program.

Engineering Test Reactor, September 19, 1957, to December 1981 

In 1972, a Sodium Loop Safety Facility was added to the ETR reactor core. With this, the reactor 

played a new role supporting the DOE breeder reactor safety program. ETR test programs were 

related to the core design and operation of breeder reactors. As testing progressed, the reactor was 

modified with a new top closure to accommodate the irradiation loop. Other additions included a 

helium coolant system and sodium-handling system. The ETR was the first complete reactor facility 

to be deactivated and the D&D to be documented immediately after shutdown.

Internal exposure from airborne radionuclides was minimal during normal reactor operation. 

Exposures might have occurred during shutdown because airborne fission products were often 

released when the reactor top was removed and access was made to the reactor subpile room and 

experiment cubicles for maintenance activities. Some releases of activation products from 

experiments, particularly in loops, occurred [96]. 
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External exposure

2.7.3

was received by workers in the reactor area during shutdown and changes of loop 

and lead experiment samples as required. There were cases of significant gamma fields exceeding 

50 mrem/hr from fission and activation products in the reactor subpile room, loop cubicles, and nozzle 

trench [97]. 

The date information in this discussion is from Stacy (2000, p. 260). The ATR is the latest materials 

testing reactor to be built in the TRA. It simulates the environment in a power reactor to study the 

effect of radiation on steel, zirconium, and other materials. The ATR produces an extremely high 

neutron flux, making it ideal for materials testing. Target materials are exposed to the neutron flux to 

test their durability in an environment of high-temperature, high-pressure, and high-gamma-radiation 

fields. Data that normally would require years to gather from ordinary reactors can be obtained in 

weeks or months from the ATR.

Advanced Test Reactor, July 7, 1967, to 2007

The ATR can operate up to a power level of 250 MW. Its unique four-lobed design can deliver a wide 

range of power levels to nine main test spaces, or loops. Each loop has its own distinct environment 

apart from that of the main reactor core. Smaller test spaces surrounding the loops enable additional 

tests. In addition, the ATR produces radioisotopes for use in medicine, industry, and other research.

Internal exposure might have occurred during reactor shutdown from mixed fission and activation 

products released to the air in the occupied environment of the building. The most probable locations 

would be from the reactor top, experiment cubicles, primary coolant rooms, and subpile room. During 

reactor operation, areas with airborne radiation are exclusion areas [98]. 

External exposures

2.7.4

are received primarily during reactor shutdown from fission and activation 

products in the fuel, experiments, and associated hardware. Work on the primary system, in the 

reactor tank, or in loop cubicles would have associated external exposure. Handling of isotope 

production samples would produce some exposure depending on the nuclide and quantity. Typical 

nuclides generated would include 192Ir, 99Tc, 60Co, and 131I [99]. 

The RMF was a very-low-power reactor in the east end of the MTR canal that operated at a power 

level of 100 or 200 W. Water was its moderator, reflector, and shield. It was designed to measure 

reactivity changes in materials irradiated in the MTR or ETR. The RMF was used to assay new and 

spent fuel elements and to assist in experiment scheduling by evaluating reactivity losses and flux 

depression caused by in-pile apparatus (Stacy 2000). 

Reactivity Measurement Facility, February 11, 1954, to April 10, 1962

Internal exposure. There was a potential for internal exposure from MFPs or MAPs from airborne 

radioactivity during maintenance activities [100]. 

External exposures

2.7.5

were minimal due to the low-power operating level and the depth of the pool. Any 

external exposures would have come from fission and activation products, primarily during fuel 

handling [101]. 

The ARMF-I reactor was in a small pool in the TRA-660 building east of the MTR building. It was 

used to determine nuclear characteristics of reactor fuels and other materials for testing in the MTR. 

Advanced Reactivity Measurement Facility No. 1, October 10, 1960, to 1974 
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2.7.6

The date information in this discussion is from Stacy (2000, p. 260). The ARMF-II reactor was in the 

opposite end of the tank occupied by ARMF-I. It had a “readout” system that automatically recorded 

measurements on data cards. This refinement over the ARMF-I reactor meant operators could 

process data quickly in computers. 

Advanced Reactivity Measurement Facility No. 2, December 14, 1962, to 1968

Internal exposure. ARMF-I and ARMF-II airborne activity of fission products could exist if a fuel 

element or sample was damaged during handling [102]. 

External exposure

2.7.7

from ARMF-I and ARMF-II was minimal due to their low operating power levels and 

the depth of the canal in which they are located. Exposures would be from fission products and/or 

activation products during fuel or experiment handling [103]. 

When the ARMF-II reactor was modified in 1968, it received a new name, the Coupled Fast Reactivity 

Measurement Facility (CFRMF). The core was modified to produce a region of high-energy neutron 

flux to provide physics information about the behavior of fast (unmoderated) neutrons. Physicists 

studied differential cross-sections and tested calculation methods. The CFRMF contributed to the 

development of fast neutron reactors (Stacy 2000).

Coupled Fast Reactivity Measurement Facility, 1968 to 1991

Internal exposure potential existed from airborne fission products that might have been released from 

the fuel [104]. 

External exposure.

2.7.8

 Minimal exposure resulted from working on the reactor top to move fuel or 

experimental components. MFPs or MAPs would be the major contributors [105]. 

ETRC was a full-scale, low-power nuclear facsimile of the ETR in TRA-635, similar in function to the 

ARMF and ATRC. It was used to determine the nuclear characteristics of fuel and experiments 

planned for irradiation in ETR and/or the power distribution effects for a given ETR fuel and 

experiment loading. ETRC enabled operators to predict the nuclear environment when completed 

experiments were removed or new experiments were added to calculate the irradiation and determine 

core life, control rod withdrawal sequences, reactivity worth, and core safety requirements.

Engineering Test Reactor Critical Facility, May 20, 1957 to 1982

Mockups of fuel and experiment loadings in ETRC were manipulated until a desired power distribution 

throughout the core was attained, satisfying pertinent safety requirements. ETRC low-power tests 

enabled the ETR to operate without interruption, saving time and money (Stacy 2000).

Internal exposure from airborne radioactivity of MFPs from damaged fuel or leaky experimental 

samples containing activation products is possible [106]. 

External exposure. Irradiated fuel or samples were transferred into or out of the ETRC under water to 

minimize exposure. The operating console was not on the reactor top, which minimized external 

exposure. External exposure would have been from MFPs and activation products during loading or 

unloading fuel or test samples [107]. 
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2.7.9

The ATRC performs functions for the ATR similar to those of the ARMF reactors for the MTR. It 

verified for reactor designers the effectiveness of control mechanisms and for physicists predictions of 

power distribution in the large core of the ATR. Low-power testing in the ATRC conserved time so the 

large ATR could irradiate experiments at high power levels, and verified the safety of a proposed 

experiment before it was placed in the ATR (Stacy 2000).

Advanced Test Reactor Critical Facility, May 19, 1964, to 2007

Internal exposure was possible from fission products released from damaged fuel or experiment 

samples containing activation products. The damage could occur during transfer into or out of the 

ATRC. Some fuel or samples had been preirradiated in the ATR or other facilities [108]. 

External exposure

2.7.10

occurred during canal work. The operators leaned over the canal parapet to work 

with irradiated fuel or experiment samples being transferred into or out of the core under water. 

During reactor operation, access to the canal parapet is restricted. The ATRC canal is 21 ft deep

[DOE 1996, pg. 9]. 

The TRA Hot Cells are southwest of the ETR reactor building. They consist of three separate cells 

with a common operating corridor. The operators are protected by thick concrete walls and special 

viewing windows, behind which they can handle, photograph, mill, measure, and weigh radioactive 

samples. The work in the cells has involved all types of samples including irradiated fuel, TRU 

materials, and isotope production material. Irradiated samples, including failed reactor fuel, can be 

put in Cell 1 or 3 from a shielded cask outside the building. Each cell has to be entered periodically to 

repair equipment or set up for a new job. Entry is through a shielded door in the back of the cell. 

TRA Hot Cell Facility, 1954 to 2007

Internal exposure might occur during cell entries from the samples and the work performed with 

irradiated samples, dust, or particles that could become airborne [109]. 

External exposure

2.7.11

occurs from samples that go in the hot cells. Exposure to beta, gamma, and 

neutrons has occurred. Fuel samples of different types and different ages result in mixed fission and 

activation products and TRU materials, including 252Cf. Many samples have been prepared as 

isotopic sources for industrial or medical applications. Cell entry and sample handling result in most 

of the external exposure, and some has occurred from sample ports and manipulator removal 

operations [110]. 

The TRA Gamma Facility was south of the original TRA main security gatehouse. The facility 

consisted of a 16-ft-deep canal with cadmium buckets designed to hold spent MTR fuel elements. 

Experimental samples were inserted in sample tubes and lowered into extremely high gamma fields

(Stacy 2000). 

TRA Gamma Facility, 1955 to Unknown 

Sponsors provided a large variety of materials and samples for gamma irradiation, including food 

products and some natural substances such as gold, diamonds, and oil. Irradiated samples, which 

were not radioactive, were surveyed thoroughly for external contamination on removal.

Internal exposure was possible from a spent fuel element used for irradiations that was unlikely to be 

damaged [111]. 
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External exposure

2.7.12

was minimal due to handling procedures and shielding from the canal and 

transport devices. Fuel was transported in large casks into and out of the canal [112]. 

The Radiation Measurements Laboratory (RML), in the MTR west wing, was previously called the 

MTR Counting Room. The RML specializes in measuring quantity and quality of alpha, beta, gamma, 

and neutron radiation samples. A variety of counting equipment and spectrometers are available in 

the RML. Over the years, an endless variety of samples has been brought in for counting. Some of 

the detector shielding was made of pre-World War II battleship steel. The original equipment has 

been replaced with more modern equipment. 

Radiation Measurements Laboratory, 1952 to 2007

Internal exposure is possible from airborne radioactivity from mishandled samples brought in for 

counting. The isotopes encountered would include fission products, activation products, TRU 

materials, and more [113]. 

External exposures

2.7.13

are very low due to the small samples required for the counting equipment. 

Counting room personnel can encounter neutrons and beta and gamma emitters [114]. 

The Radiochemistry Laboratories are in the MTR west wing. They are used to support the RML and 

to perform independent R&D work. Investigators study methods to produce and purify medical 

radioisotopes and the effects of radiation on hazardous waste.

Radiochemistry Laboratories

Laboratories 109 to 112 were used primarily for chemical analysis of reactor primary systems and 

loop experimental coolants. The predominant radioactivity was from MFPs and MAPs. The south 

extension to the MTR Wing is the Alpha Laboratories, designed for the safe handling of hazardous 

alpha emitters such as 233U, 239Pu, 241Am, and other TRU materials, including 252Cf (Stacy 2000). 

Internal exposure potential exists from airborne radioactivity from the large variety of samples [115]. 

External exposures

2.7.14

occur when irradiated samples are brought into the laboratories. One laboratory 

has a shielded box, similar to a small hot cell, for handling highly radioactive samples [116]. 

The TRA Liquid Waste Ponds are east of the ETR reactor building. The 7.5-acre ponds were built for 

the disposal of low-level liquid waste from test reactor operations. When all three test reactors were 

operational, approximately 50 million gal of wastewater per month were discharged to the seepage 

ponds and the ETR disposal well. Most of the activity pumped to the ponds was 51Cr and 3

H (Nebeker 

and Lakey 1970).

Liquid Waste Disposal Ponds 

An estimated 3,000 wild ducks per year land on the pond, usually stay less than a week, and have 

some potential to carry activity off the site. An extensive study analyzed the ducks for ingestion of 

134Cs, 137Cs, 75Se, 131I, 239Pu, and 240Pu (Till et al. 2002).

Internal exposure. The seepage ponds accumulated a significant quantity of MFPs. The activity was 

fairly stable as long as water levels remained high. When the water was allowed to recede and soil 

was allowed to dry, the activity could become airborne by the winds [117]. 
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External exposure

2.7.15

occurs to the operators from old accumulated activation products and fission 

products during routine inspection or sampling of the ponds. Several studies of the water and the soil 

in the pond resulted in additional external exposure [118]. 

On the south side of the MTR HP office are four 1,500-gal HLLW catch tanks placed underground in 

concrete vaults. Tanks 1 and 2 receive water from the Hot Drain System, which includes the MTR 

floor drains. Tanks 3 and 4 receive liquid waste from the Radiochemistry Laboratory and the TRA Hot 

Cells. The liquid waste is accumulated until the tank is nearly full and then sampled. If the waste 

meets low-level waste criteria, it is transferred to the Retention Basin and then to the TRA Liquid 

Waste Disposal Ponds. If it does not meet the criteria, the waste is transferred to the HLLW holding 

tanks. When the Hot Waste tanks are nearly full, the wastewater is loaded on a tank truck and 

shipped to ICPP for processing. The HLLW tanks consist of two 7,500-gal and two 9,000-gal tanks

(Stacy 2000). 

High-Level Liquid Waste Tanks and Transfer Facility

Internal exposure potential exists from airborne radioactivity during entry to the holding tank pits for 

repairs from contaminants of MFPs and activation products [119]. 

External exposure

2.8 AUXILIARY REACTOR AREA, APRIL 1958 TO LATE 1990S 

due to sampling and transferring liquid waste is low because the tanks are in 

concrete-shielded underground vaults. Radiation sources are MFPs or MAPs when entry to the pits is 

necessary [120]. 

The ARA was originally the Army Reactor Experimental Area, which changed to the Army Reactor 

Area. It was established to test stationary, portable, or mobile reactors of low-, medium-, or highpower units. The ARA reactors were built and maintained by contractors with a mixed cadre of 

military personnel training on the operation of the facilities. ARA is 10 mi east of CFA; it began with 

the ARA-I site one-half-mile north of Highway 20. ARA-II, -III, and -IV are at half-mile intervals along 

an access road, Fillmore Avenue, north from Highway 20. After the Army phased out its program 

around 1965, ARA facilities were used for experiments and tests involving multiple radionuclides, 

particularly at the ARA-I hot cell and laboratory facilities (Stacy 2000). D&D of the ARA ended in 

2002.

2.8.1

ARA-I housed a hot cell facility (ARA-25) and laboratory with hoods and metallurgical equipment to 

support reactor and other radiological experimental work. It operated from the early 1960s through 

the late 1990s with periods of inactivity. The Hot Cell was involved with recovery of debris from the 

SL-I excursion accident and associated reactor and fuel experiments. The hot cell and laboratory 

were involved in a wide spectrum of activities from low-level alpha experiments to work with irradiated 

reactor samples exposed to core fluxes, including melted and destroyed fuel assemblies from reactor 

tests (Stacy 2000). 

ARA-I 

2.8.2

ARA-II, a half-mile north of ARA-I, was the site of a low-power BWR designed and built by ANL based 

on its BORAX experience. The reactor, designed to generate only 1,000 kW, was originally named 

the Argonne Low Power Reactor. After ANL handed over the finished plant to the Army’s operating 

contractor, Combustion Engineering, the Army named the reactor the Stationary Low-Power Reactor 

ARA-II/SL-1, February 23, 1960, to April 6, 1961
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Number 1, or SL-1. SL-1 went critical for the first time on August 11, 1958. It operated for periods 

between 1 and 6 wk, and then would be shut down for scheduled maintenance and test changes. 

SL-1 was shut down for scheduled annual maintenance on December 23, 1960, and was scheduled 

for a January 4, 1961, startup. During preparation for the run, the reactor went “prompt critical” at 

9:01 p.m. on January 3, 1961, creating a steam explosion that killed three persons and destroyed the 

reactor. The event released fission products (500,000 Ci in the building and 1,100 Ci to the 

atmosphere; Horan and Braun 1993) and created high-level radioactive contamination to 50 rad/hr 

around the ARA-II area. Initial recovery from the accident resulted in short-term exposure exceeding 

500 rad/hr to personnel in radiation fields. Extensive cleanup efforts followed, including complete 

dismantlement of the facility. The reactor vessel went to the TAN, some of the contaminated items 

went to the RWMC, and some debris was buried in a specially designated location (two large pits and 

a trench) about 1,600 ft from the SL-1 compound. The walls of the silo, the power conversion and 

fan-floor equipment, the shielding gravel, and the contaminated soil gathered during the cleanup went 

into the pits at SL-1. Three feet of clean earth shielded the material. An exclusion fence with hazard 

warnings around the area remains in place east of the reactor site. Operating power history and 

release information is in Till et al. (2002, p. 117).

2.8.3

ARA-III, another half-mile farther north, was the site for the Army Gas-Cooled Reactor Experiment 

(GCRE). GCRE was a water-moderated, nitrogen (gas)-cooled, direct and closed-cycle reactor. It 

generated 2,200 kW of heat, but no electricity. The Army wanted to develop a mobile nuclear power 

plant, and the GCRE was the first phase of that program, proving the principle of this concept. The 

reactor provided engineering and nuclear data for improved components. The GCRE was used to 

train military and civilian personnel in the operation and maintenance of gas-cooled reactor systems

(Stacy 2000, p. 263). 

ARA-III

2.8.4

ARA-IV, another half-mile north on Fillmore Avenue, was the site for the Mobile Low-Power Reactor 

(ML-1). The entire ML-1 plant was designed to be transported either by standard cargo transport 

planes or standard Army low-bed trailers in separate packages weighing less than 40 tons each. The 

ML-1 reactor was operated remotely from a control cab about 500 ft away. It could be moved after a 

36-hr shutdown. The reactor was designed for ease of operation and maintenance by technicians at 

remote installations, for reliable and continuous operation under extreme climatic conditions, and for 

the rigors of shipment and handling under adverse conditions. The Army phased out its reactor 

development program around 1965.

ARA-IV, March 30, 1961 to May 29, 1964

The following exposures relate to all ARA facilities (Stacy 2000, p. 265). 

Internal exposure potential was typical of that associated with operation of a low-power reactor with 

the exception of the Hot Cell (ARA-I) effluent and the SL-1 excursion, which resulted in fission product 

release (1,100 Ci to the atmosphere). Airborne radioactivity consisting of MFPs and activation 

products would cause most internal exposures [121]. 

External exposure was significant from the SL-1 excursion and recovery operations. Nine technical 

overexposures resulted, ranging from 15 to 27.3 rem (whole-body dose) (Horan and Braun 1993). 

Hot cell work and D&D efforts in the reactor areas contributed to exposure pathways. Hot cell entries 

after working with irradiated experiments were a source of high exposure. Expected exposure 

potential was from photons greater than 250 keV from the MFPs and activation products associated 
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with SL-1, Hot Cell operations, and other work connected with reactor operation and maintenance 

activities. Dosimetry badges were worn by workers in this area. However, extremity dosimetry might 

not have been used by all personnel during recovery from the SL-1 accident [122]. 

2.9 WASTE REDUCTION OPERATIONS COMPLEX/POWER BURST FACILITY/SPECIAL 

POWER EXCURSION REACTOR TEST AREAS 

The Waste Reduction Operations Complex/Power Burst Facility (PBF)/SPERT area is south-centrally 

located east of the CFA on the INL site and 51 mi west of Idaho Falls. The site was originally 

established to conduct research on small power reactors and reactor safety. Its current mission is 

storage of spent nuclear fuel, treatment and storage of mixed and low-level waste, and research to 

reduce hazardous and mixed waste. As part of the DOE mandate to treat legacy wastes and 

remediate the environment, these facilities now provide safe treatment, storage, and recycling of INL

radioactive, mixed, and low-level wastes. Many of the treated wastes originated at INL. Internal and 

external exposure will be noted at the end of Section 2.9 for the following subsections.

2.9.1

SPERT-I was an open-tank, light-water-moderated, reflected reactor, originally using 92% enriched 

uranium fuel. The reactor tank, about 4 ft in diameter and 14 ft high, was the first in a series of four 

safety-testing reactors designed to study the behavior of reactors when their power levels changed 

rapidly. Power runaways were deliberately produced by moving the control rods. The variables in the 

thousands of SPERT studies included fuel plate design, core configuration, coolant flow, temperature, 

pressure, reflectors, moderators, and void and temperature coefficients (Stacy 2000, p. 266). 

SPERT-I, June 11, 1955, to 1964

At 12:25 p.m. on November 5, 1962, destructive Test No. 1 was initiated with a plate-type core. A 

violent explosion occurred immediately after the final power excursion, during which complete fuel 

plate melting occurred in approximately 8% of the core, with partial melting in approximately 35% of 

the core. It was reported that “those isotopes which were collected were released as gases. No solid 

products were collected.” Test No. 2 began at 8:15 a.m. on November 10, 1963, and Test No. 3 

began at 1:14 p.m. on April 14, 1964 (Miller, Sola, and McCardell 1964). A number of nondestructive 

runs were conducted to gain operational information. All operations were conducted from a control 

building a half-mile from the reactor. SPERT-I tests demonstrated the damage-resistant capabilities 

of low-enrichment (4% 235U) uranium-oxide fuel pins similar to those used in water-cooled reactors 

powering large central stations. 

2.9.2

SPERT-II, south and east of SPERT-I, was an extension of the SPERT-I excursion tests. It was a 

closed PWR with coolant flow systems designed for operation with either light or heavy water. The 

pressure vessel was 24.5 ft high with a 10-ft inside diameter. Tests with heavy water (deuterium, an 

isotope of hydrogen) were desired because heavy-water reactors were of growing importance in 

Canada, Europe, and the United States. In addition, heavy-water tests enabled verification of physics 

calculations on the effects of neutron lifetime on power excursions (Stacy 2000, p. 265). 

SPERT-II, March 11, 1960, to October 1964

At present, the area is used as a lead storage facility; it has been primarily used for storage since 

1964. PBF-contaminated reactor coolant was stored in a tank at the facility and other components 

were stored in a radioactive material storage area. The radioactive liquid waste and radioactive 

material storage area have been removed.
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2.9.3

SPERT-III was the most versatile facility yet developed for studying the inherent safety characteristics 

of nuclear reactors. This reactor (which was planned as the third in the series of SPERT reactors but 

was the second built) provided the widest practical range of control over three variables: temperature, 

pressure, and coolant flow. The reactor was in a pressurized vessel similar to those used in 

commercial power production. Water could flow through the vessel at a rate as high as 

20,000 gal/min, handle temperatures as high as 650°F, and pressures as high as 2,500 psi (Stacy 

2000, p. 266). 

SPERT-III, December 19, 1958, to June 1968

2.9.4

SPERT-IV was an open-tank, twin-pool facility that permitted detailed studies of reactor stability as 

affected by varying conditions including forced coolant flow, variable height of water above the core, 

hydrostatic head, and other hydrodynamic effects. The water-moderated and reflected reactor used 

highly enriched, aluminum-alloyed, plate-type fuel elements. The SPERT-IV facility was modified by 

the installation of a capsule drive core (CDC), which permitted the insertion of fuel samples in a test 

hole in the center of the reactor core, where they could be subjected to short-period excursions 

without damaging the “driver” fuel in the rest of the core. Work on fuel destructive mechanisms 

continued until the PBF replaced the CDC. 

SPERT-IV, July 24, 1962, to August 1970

In commercial plants, the reactor cores contain tons of fuel. Analysts imagined the consequences if 

the coolant somehow failed to carry away the fission heat. Suppose a pipe leaked or broke? The 

SPERT tests had proven that such a situation would easily put a stop to the chain reaction: the loss 

of pressure would allow the water to turn to steam; the lower density of steam would fail to moderate

the neutrons; and the nuclear reaction would stop. But the radioactive decay of the fission products in 

the fuel elements would continue to produce heat and continue to need cooling. This concern 

spawned the LOCA Program and the PBF. The SPERT reactors were decommissioned and replaced 

with other operations, as noted below (Stacy 2000, p. 266). 

2.9.5

PBF is a much larger and more sophisticated reactor than the SPERT reactors. It was built on the 

site of the SPERT-I facility. PBF was initially developed to perform tests of nuclear reactor fuels 

during off-normal reactor operations. It was designed to simulate various kinds of imagined accidents 

caused by sudden increases in the reactor operating level. PBF was the only reactor in the world that 

could perform rapid power changes (bursts) within milliseconds. It performed simulated LOCAs and 

severe-fuel-rod-burst tests in a special assembly (loop) in the main reactor core. Fuel damage on 

experiments in the loop would transport fission products throughout the loop piping and through steam 

lines outside the shielded loop cubicle. Monitors detected and timed the precise movement of fission 

products as they escaped from a fuel rod with failed cladding. Data from these tests were used to 

develop and validate fuel behavior computer codes for the NRC. Retrieval of data and modification of 

the test configurations resulted in exposure to high radiation fields and potential for release of fission 

products in the reactor containment (Stacy 2000, p. 266). 

Power Burst Facility, September 22, 1972, to 1985

The PBF was a high-performance, water-cooled, uranium-oxide-fueled reactor designed to provide 

information on light-water reactors. Airborne effluents were filtered and passed through charcoal beds 

to remove iodine. Liquid wastes were pumped to a disposal well or held in tanks for transport to the 

ICPP. The reactor operated from September 22, 1972, until 1985, when it was placed on standby 

status. In 1998, the PBF was placed in shutdown status and is being prepared for fuel removal.
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2.9.6

The Lead Storage Facility is in the old SPERT-II facility. It is used to collect and store clean lead for 

the INL emergency lead inventory. The building was used in the past for storage of radioactive 

material. The area is not a radiological concern [123]. 

Lead Storage Facility

2.9.7

The Waste Experimental Reduction Facility (WERF) is a versatile waste treatment facility that began 

treating low-level radioactive wastes in 1982 at the location of SPERT-III after that facility’s D&D. Its 

original mission was to reduce the volume of low-level radioactive waste through incineration, 

stabilization, compaction, and metal sizing processes to prepare the wastes for safe permanent 

disposition before burial at the RWMC. In the beginning, metal was sized and melted into ingots in 

two furnaces until it was determined sizing alone was more cost-effective. An incinerator was added 

and used to reduce the volume and increase the stability of a wide variety of low-level wastes before

disposal at the RWMC. In 1984, the incinerator began treating RCRA-defined mixed (radioactive and 

hazardous) waste. Wastes from INL and other DOE facilities were treated under provisions in the 

INEEL Site Treatment Plan (Stacy 2000). INL began the closure process for the WERF, which should 

take 3 years, in 2001.

Waste Experimental Reduction Facility, 1982 to 2001

2.9.8

The Mixed Waste Storage Facility, in the SPERT-IV reactor building, is a RCRA storage facility for 

interim storage of mixed low-level wastes. It has regulatory approval to store polychlorinated 

biphenyls, corrosives, and flammables (Stacy 2000). Treatments are being developed for the types of 

waste stored in the facility.

Mixed Waste Storage Facility

Internal exposures were possible based on releases from operations at the SPERT reactors and PBF. 

Maintenance activities and other work with radioactive material (especially from PBF loop experiment) 

resulted in airborne MFPs and MAPs, making internal exposure possible, with 137Cs being the primary 

radionuclide [124]. 

External exposure

2.10 ORGANIC MODERATED REACTOR EXPERIMENT, SEPTEMBER 17, 1957, TO APRIL 

1963

resulted from experiment changes related to reactor experiment changes and 

maintenance activities. Cesium-137 was a primary nuclide for direct radiation exposure from fission 

products in the transport lines and in the loops at the PBF during Severe Fuel Damage tests when 

radiation levels were measured up to 50 rad/hr. Other radiological work activities resulted in much 

lower exposure rates from the MFPs and activation products [125]. 

The Organic-Moderated Reactor Experiment (OMRE), a few miles east of the CFA, was built to test 

the feasibility of the organic-cooled reactor concept. OMRE demonstrated the technical and 

economic feasibility of using a liquid hydrocarbon as both coolant and moderator. The reactor 

operated with a succession of cores. The waxy coolant was considered promising because it 

liquefied at high temperatures but did not corrode metal as water did. In addition, it operated at low 

pressures, significantly reducing the risk of leaks. However, it lacked test loops needed to investigate 

various organic coolants and experimental fuel elements (Stacy 2000, p. 266). 

A scaled-up reactor, the Experimental Organic-Cooled Reactor (EOCR), was built next to OMRE in 

anticipation of further development of the concept. The purpose of EOCR, which had special testing 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 54 of 91

loops and other advanced features, was to extend and advance the OMRE studies. During the final 

stages of its construction, EOCR was placed in standby (December 1962) when the AEC decided that 

the organic-cooled concept would not significantly improve performance over the achievements of

other reactor concepts for nuclear power. EOCR never operated. The building was recycled for other 

(non-nuclear) uses.

Internal Exposure. On November 16, 1960, an experiment was conducted to determine the feasibility 

of open-air burning of contaminated solvents accumulated at the OMRE facility. Approximately 

400 gal of liquid consisting of diesel oil, xylene, methyl-chloroform, and a small amount of water were 

placed in an open vessel and ignited. Because the reactor did not operate, no other potential for 

internal dose occurred [126]. 

External exposures

2.11 TEST GRID III, 1957 TO APRIL 24, 1970

were possible from MFPs and activation products associated with core changes 

and associated reactor maintenance in radiological areas [127]. 

Test Grid III, near Lincoln Boulevard south and east of NRF and east of the TRA and INTEC, was the 

site of several tests with atmospheric releases. It originated in the 1957 timeframe, and was based on 

concerns about what would happen from nuclear aircraft crashes, etc. This highly instrumented grid 

would measure atmospheric conditions and release information of the tests (Till et al. 2002). These 

experiments were planned and conducted by the DOE Health Services Laboratory rather than by a 

contractor. Exposure information is summarized at the end of Section 2.11 for the following 

subsections.

2.11.1

Fuel Element Burn Tests A and B were conducted on Grid III to support the GE ANP Program to 

evaluate the consequences of a nuclear aircraft crash involving a fire. Aged fuel elements were 

heated to assist with understanding the behavior of a fuel element in a large fire and to provide initial 

data on the percentage release of fission products to the environment. Average ground radiation 

levels near the burn site immediately following the burn test were 200 mrem/hr (Brodsky and Beard 

1960). Meteorological conditions had been carefully studied in advance and were closely monitored 

during the tests.

Fuel Element Burn Tests

FEBT-A was conducted at 2:19 p.m. on March 20, 1957, using an irradiated fuel element (well-aged) 

containing 5,000 Ci of fission products. A pool of jet fuel was ignited under the fuel element, which 

reached a temperature of about 2,250 ̊F. After the fire, the fuel element was intact with a small 

puncture in the cladding [128]. 

FEBT-B used an induction furnace that heated a fuel element containing 10,000 Ci of fission products 

to 5,000 ̊F. Most of the fuel element melted and dispersed within 90 s during inversion conditions 

(Brodsky and Beard 1960). The test was conducted at 6:47 p.m. on March 20, 1957. FEBT-B was 

ranked above other release events at onsite and offsite locations as an episodic event in Till et al.

(2002). 

2.11.2

Fission Products Field Release Tests were conducted on Grid III to represent accidents involving 

nuclear-powered aircraft. Nine tests evaluated release percentages, airborne radioactivity, and 

diffusion and deposition characteristics of fission products released from melted aircraft reactor fuel 

Fission Products Field Release Tests, July 25, 1958, to September 26, 1958
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elements (Convair 1959). Five tests were with fuel decayed for 922 to 985 d and four were with fuel 

decayed for 42 to 65 d. Operating temperatures were between 1,000 and 2,300 ̊C. To simulate a 

potential accident, the tests used an induction-type furnace to heat the elements rapidly to the melting 

point in approximately 2 min, and maintained this temperature for approximately 10 min after melting 

began (Convair 1959). Instruments situated about a fan-shaped grid with seven concentric arcs and a 

maximum radius of about 5 mi obtained cloud diffusion, meteorological, radiological, radiobiological, 

and deposition data. Till et al. (2002) contains information on meteorological conditions, furnace 

temperatures, release fractions, etc.

2.11.3

The four Relative Diffusion Tests involved the intentional release of 1 to 6 Ci of both methyl and 

elemental radioiodine. Details on these releases are limited, but some information is in DOE (1991b) 

and Till et al. (2002). 

Relative Diffusion Tests, November 30, 1967, to October 1, 1969

2.11.4 Experimental Cloud Exposure Study, May 3, 1968, to April 24, 1970

The Experimental Cloud Exposure Study tests in 1968 and 1969 consisted of 133Xe releases ranging 

from 32 to 600 Ci; tests in 1970 consisted of 24Na releases ranging from 6.6 to 120 Ci. The primary 

objectives for the tests included measuring total exposure at several downwind distances; determining 

dimensions of the plumes; documenting the release rate and height, wind speed, and temperature; 

and measuring the gamma energy spectrum at one or more points during the release. Releases were 

planned to occur during meteorological conditions characterized by winds from the southwest to 

minimize potential on- and offsite exposure and to ensure that the cloud passed over preset 

instrumentation. Voillequé (1969) discusses an outline of plans for the 133Xe release tests, including 

the general objectives and procedures associated with the tests (Till et al. 2002). 

Internal exposures were possible from the airborne radioactivity released from the tests and potential 

airborne radioactivity from the materials during handling. Grid III was highly instrumented to detect 

release fractions. FEBT-B had a high ranking for INL releases, and the least amount of monitoring, 

according to Till et al. (2002). Personnel participated in the site’s bioassay program [129]. 

External exposure would have resulted from working in the proximity of, handling, or transporting 

irradiated test fuel elements with a potential for exposure from photons greater than 250 keV 

consistent with that of other irradiated fuel loading and unloading tasks [130]. 

2.12 EXPERIMENTAL FIELD STATION/EXPERIMENTAL DAIRY FARM

The EFS was a 27-acre plot about 7 mi northeast of the ICPP near Grid III. This facility, also known 

as the Dairy Farm, was established to further studies on the pathway of 131I from a release to the 

human thyroid. It included pastures, a barn, six cows, and a grid of detection instruments in the 

pasture in regular lines and rows (Stacy 2000). 

2.12.1 Controlled Environmental Radioiodine (Release) Tests, May 27, 1963, to December 

1977

The primary objectives of the Controlled Environmental Radioiodine Test (CERT) releases were to 

establish relationships between the amounts of radioiodine in different environmental media. These

tests specifically studied relationships between air and soil and vegetation, vegetation and milk, and 

milk and human thyroids. They involved releases of elemental and methyl radioiodine ranging in 

amount from 0.05 to 8 Ci. Most of the releases occurred at the Experimental Dairy Farm. Others 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 56 of 91

occurred at ICPP, ARA, NRF, and CFA. In 1968, the name was changed to Controlled Environmental 

Release Test to reflect the release and study of additional radionuclides such as cesium, cerium, 

potassium, and krypton. Hawley (1964) reports that the CERT No. 1 release occurred near ground 

level over a 30-min period. Additional information regarding this test series is provided by Hawley 

(1964); Bunch (1966, 1968), and Zimbrick and Voillequé (1969). Early in the test program, the AEC 

granted permission to six DOE-ID volunteers to be part of a human experiment program. During initial 

tests these volunteers sat in the field during the release. In later tests, after the cows had eaten 

contaminated grass, the volunteers drank small quantities of milk. The series included 29 

experiments, although only a few of the early ones involved human consumption of milk [131]. 

Internal exposures were possible from all the releases and were intentional during the early phases of 

the tests. Exposures are well-documented in the dosimetry records [132]. 

External exposure was well below the level acceptable for radiological work from the tests. Personnel 

were required to wear dosimetry devices [133]. 

2.13 INL RESEARCH CENTER, 1984 TO 2007

The INL Research Center (IRC), a 35-acre site on North Boulevard in Idaho Falls, was built between 

1982 and 1984. The facility was dedicated in 1984 to further the INL research and engineering 

mission. The IRC has 58 laboratories in IF-603 for geophysics, chemistry, microbiology, and other 

sciences; 18 of these are general-purpose modules for electronics design, optics, lasers or materials 

testing, and nondestructive examination research and development. The Center conducts laboratory 

work with tracer-level radionuclides. One of the early missions was work for the Bureau of Mines with 

ores, some containing natural uranium. The INL Engineering Demonstration Facility (IF-657) houses 

several prototypical-scale R&D projects that support programs in military munitions assay, advanced 

sensor systems, environmental restoration, subsurface investigation, and materials science. There is 

a shielded 252Cf source (initially 2.5 mCi) in the high bay of IF-638. The onsite radiological control 

technician has additional low-energy plutonium, strontium, and americium sources for portable survey 

instrument response checks [134]. 

Internal exposure potentials are minimal because of the radiotracer-level, low-energy alpha and beta 

source material such as natural uranium, 14C, etc. [135].

External exposure. Low-level personnel exposures have been measured from the neutron source (2-

to-20-MeV range) and X-ray (30-to-250-keV range) equipment. Personnel dosimetry is required for all 

work in radiation areas [136]. 

2.14 ARMY REENTRY VEHICLE FACILITY SITE (OR STATION), 1965 TO MAY 1996

The Army built the Army Reentry Vehicle Facility Site (ARVFS) 12 mi northeast of CFA in 1965 for 

classified DOD experiments with an advanced reentry vehicle fuzing system. The facility consisted of 

an open-top cylindrical test pit, an underground bunker, and a system of cables and pulleys between 

the bunker and tank (Thiel 1997; Mobley 1987). 

ARVFS was used in 1965 to conduct an irradiation study using four spent MTR fuel elements to 

evaluate the accuracy of the Radiological Safety Analysis Computer-generated code cloud-gamma 

exposure information. Movement of the fuel to conduct the experiment and the transportation of the 

fuel to and from the facility was an external exposure source. Dose rates are not available (McCaslin 

1968).
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 57 of 91

The bunker was used to store NaK coolant radiologically contaminated from the EBR-I Mark II nuclear 

reactor core meltdown in November 1955. The NaK was stored from 1974 through 1995, when it was 

shipped to ANL-W for reprocessing. Engineering evaluations from 1986 through 1992 provided 

radiological data with maximum radiation exposure rates from the NaK containers to be ~ 40 R/hr. 

The radiological fission product inventory in the NaK reported in LaRue and Dolenc (1986) was 23.3 g 

and about 133 Ci. The Final Safety Analysis Report indicates that the conservative total dose to 

process the four NaK containers would be 0.394 rem, assuming the same person was involved with 

each step (Mobley and Keller 1991). 

Internal exposure potential was minimal at ARVFS because airborne radioactivity was not present 

during activities at the bunker [137]. 

External exposure occurred during radiological surveys and loading and unloading of the NaK 

containers for storage and or transport [138]. 

2.15 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. 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. 

Donald Marshall served as one of the initial Subject Experts for this document. Mr. Marshall 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. Marshall, those materials are 

fully attributed to the source. 

Frank Hinckley served as one of the initial Subject Experts for this document. Mr. Hinckley 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. Hinckley, those materials are 

fully attributed to the source. 

Boyd Leavitt served as one of the initial Subject Experts for this document. Mr. Leavitt 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 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 58 of 91

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. Leavitt, those materials are 

fully attributed to the source.

[1] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This statement is based on information gleaned from a review of the references ACC 1952 

and Cipperly 1958.

[2] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This statement is based on information gleaned from a review of ACC 1952. 

[3] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This statement is based on information gleaned from a review of ACC 1952.

[4] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This statement is based on a review of the reference Hoff et. al. 1986.

[5] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This statement is based on a review of the reference Stacy 2000. 

[6] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

The Warm Shop, Hot Shop and Hot Cells were used for the analysis of various reactor 

experiments, refueling and repairs. Reactor fuel contains mixed fission products due to the 

fissioning of the uranium and mixed activation products due to activation from neutron 

bombardment. It is reasonable to expect that the potential for internal exposure to mixed 

fission products and mixed activation products exists in a facility that handles reactor fuel and 

components. 

[7] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[8] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the safety program and sources of exposure at 

TAN.

[9] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the safety program at TAN and work experience at 

the facility.
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 59 of 91

[10] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Should one of the casks leak, internal exposures could occur as the casks contain spent 

nuclear fuel and are pressurized. 

[11] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

External radiation exposures could occur from work in the proximity of the casks due to the 

elevated radiation levels. Neutron dosimetry is assigned based on the employee’s job function 

and exposure potential; therefore not all employees are issued neutron dosimetry.

[12] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

The liquid waste stream contained mixed fission and mixed activation products. External 

exposures could occur during activities in the vicinity of the liquid waste system due to 

elevated dose levels. 

[13] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the reference Bonney et. al. 1995.

[14] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on work experience at the RPSSA. This internal exposure scenario 

is routinely included in safety analyses of this type of facility, as described in section 5.7.1 of 

the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose TBD. 

[15] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[16] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on work experience at TAN. 

[17] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on work experience at TAN. 

[18] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 60 of 91

[19] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[20] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[21] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at TAN and a review of 

the reference Cordes et al. 1965. 

[22] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

The tests were performed with nonradioactive materials.

[23] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at WRRTF and work 

experience at the facility.

[24] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on work experience at TAN.

[25] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at TAN.

[26] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at TAN. This external 

exposure scenario is routinely included in safety analyses of this type of facility, as described 

in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External Dosimetry TBD. 

[27] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the reference Kunze and Chase 1970.

[28] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at TAN.
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 61 of 91

[29] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at TAN. This external 

exposure scenario is routinely included in safety analyses of this type of facility, as described 

in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External Dosimetry TBD. 

[30] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[31] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.1.1 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[32] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at TAN. This external 

exposure scenario is routinely included in safety analyses of this type of facility, as described 

in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External Dosimetry TBD. 

[33] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.2 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[34] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[35] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at ICPP and work 

experience at the facility and a review of the reference Cederberg et. al. 1974.

[36] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the reference Stacy 2000.

[37] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at ICPP and work 

experience at the facility.
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 62 of 91

[38] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.2 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[39] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[40] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at ICPP and work 

experience at the facility.

[41] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the reference Stroschein and Maeser 1967.

[42] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at ICPP and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 

[43] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on work experience at the Tank Farm.

[44] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the Tank Farm and 

work experience at the facility. This internal exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 5.7.2.2 of the ORAUT-TKBS-0007-5 

INL – Occupational Internal Dose TBD. 

[45] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the Tank Farm and 

work experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 

[46] Norm Rohrig. Consulting Health Physicist. Intrepid Technology and Resources, Inc. April 25, 

2007.

ERDA 1977 p 106 shows gross beta activity of 4840 μCi/ml (Ci/m3

) for Aluminum-Zirconium 

blend waste and 340 μCi/ml (Ci/m3

) for Second and Third Cycle waste in the tank farm.
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 63 of 91

[47] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on work experience at the WCF. 

[48] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the WCF and work 

experience at the facility. This internal exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 5.7.2.3 of the ORAUT-TKBS-0007-5 

INL – Occupational Internal Dose TBD. 

[49] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the WCF and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 

[50] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the FDF and work 

experience at the facility.

[51] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the FDF and work 

experience at the facility.

[52] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the ICDF and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 

[53] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the ICDF and work 

experience at the facility.

[54] Boyd Leavitt. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the ISFSI and work 

experience at the facility.

[55] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on understanding the sources of exposure at the ISFSI and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 64 of 91

[56] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Haroldson et. al. 1963.

[57] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Haroldson et. al. 1963.

[58] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the references Dietrich et. al. 1956 and Griffiths 

et. al. 1956. 

[59] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the references Dietrich et. al. 1956 and Griffiths 

et. al. 1956. 

[60] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the references Dietrich et. al. 1956 and Griffiths 

et. al. 1956. 

[61] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the references Dietrich et. al. 1956 and Griffiths 

et. al. 1956. 

[62] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Zinn et. al. 1956.

[63] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Zinn et. al. 1956.

[64] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Robertson and Hall 1959.

[65] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Robertson and Hall 1959.

[66] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at BORAX-V.
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 65 of 91

[67] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at BORAX-V.

[68] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Brittan et. al. 1961.

[69] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Brittan et. al. 1961.

[70] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at AFSR and a review of the reference Brunson 1959.

[71] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at AFSR and a review of the reference Brunson 1959.

[72] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Freund et. al. 1960.

[73] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on a review of the reference Freund et. al. 1960. 

[74] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This statement by Mr. Graham is based on his understanding of the sources of exposure 

based on experience at HFEF. Since the hot cells were used to examine irradiated fuel, they 

would be contaminated with mixed fission and activation products. This contamination is a 

potential internal exposure hazard during hot cell entries.

[75] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This statement by Mr. Graham is based on his understanding of the sources of exposure 

based on experience at HFEF. Since the hot cells were used to examine irradiated fuel, they 

would be contaminated with mixed fission and activation products. This contamination is a 

potential external exposure hazard during hot cell entries.

[76] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at ZPPR and a review of the reference Kunze and Chase 1970. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 66 of 91

[77] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

Information provided is based on understanding of the sources of exposure based on 

experience at ZPPR and a review of the reference Kunze and Chase 1970.

[78] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This statement by Mr. Graham is based on his understanding of the sources of exposure 

based on experience at NRAD. Since the hot cells were used to examine irradiated fuel, they 

would be contaminated with mixed fission and activation products. This contamination is a 

potential internal exposure hazard.

[79] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This statement by Mr. Graham is based on his understanding of the sources of exposure 

based on experience at NRAD and a review of the reference Richards and McClellan 1979.

[80] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at FASB.

[81] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at FASB.

[82] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at ANL-W.

[83] Graham, Earl. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

Information provided is based on understanding of the sources of exposure based on 

experience at ANL-W.

[84] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.4 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[85] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the RWMC and work

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 67 of 91

[86] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the CFA and work 

experience at the facility.

[87] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the CFA and work 

experience at the facility.

[88] Norman Rohrig. Consulting Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the HPIL and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 

[89] Norman Rohrig. Consulting Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the HPIL and work 

experience at the facility.

[90] Norman Rohrig. Consulting Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the HPIL and work 

experience at the facility.

[91] Norman Rohrig. Consulting Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

The AmBe source was transferred to Idaho State University. This information is based on the 

experience of Mr. Rohrig at INL. 

[92] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the CFA and work 

experience at the facility.

[93] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on work experience at the MTR area.

[94] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[95] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 68 of 91

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[96] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[97] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at ETR area and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 

[98] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[99] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at Advanced Test 

Reactor Area and work experience at the facility. This external exposure scenario is routinely 

included in safety analyses of this type of facility, as described in section 6.3.4 of the ORAUTTKBS-0007-6 INL – Occupational External Dosimetry TBD. 

[100] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[101] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at MTR area and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 69 of 91

[102] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[103] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the ARMF and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 

[104] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal 

Dosimetry TBD. 

[105] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the CFRMF and work 

experience at the facility. This external exposure scenario is routinely included in safety 

analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL 

– Occupational External Dosimetry TBD. 

[106] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the ETR Critical Facility 

and work experience at the facility. This internal exposure scenario is routinely included in 

safety analyses of this type of facility, as described in section 5.7.6 of the ORAUT-TKBS-0007-

5 INL – Occupational Internal Dose TBD. 

[107] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the ETR Critical Facility 

and work experience at the facility. This external exposure scenario is routinely included in 

safety analyses of this type of facility, as described in section 6.3.4 of the ORAUT-TKBS-0007-

6 INL – Occupational External Dosimetry TBD. 

[108] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on understanding the sources of exposure at the ATR Critical Facility 

and work experience at the facility. This internal exposure scenario is routinely included in 

safety analyses of this type of facility, as described in section 5.7.6 of the ORAUT-TKBS-0007-

5 INL – Occupational Internal Dosimetry TBD. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 70 of 91

[109] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal 

Dosimetry TBD and is supported by the reference Stroschien and Maeser 1967. 

 [110] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD and is supported by the reference Stroschien and Maeser 1967. 

[111] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal 

Dosimetry TBD. 

[112] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[113] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[114] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This facility is a low background facility and highly radioactive sources/samples are not 

allowed in.

[115] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This internal exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 5.7.6 of the ORAUT-TKBS-0007-5 INL – Occupational Internal Dose 

TBD. 

[116] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[117] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the references Nebeker and Lakey 1970, Till et. al. 

2002 and ERDA 1977a. 
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 71 of 91

[118] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the references Nebeker and Lakey 1970, Till et. al. 

2002 and ERDA 1977a.

[119] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the references Nebeker and Lakey 1970 and ERDA 

1977a.

[120] Don Marshall. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the references Nebeker and Lakey 1970 and ERDA 

1977a.

[121] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the reference Horan and Gammil 1961.

[122] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the reference Horan and Gammil 1961.

[123] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This storage area is not a radiological concern since the lead stored here is of the “low 

background” variety.

[124] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the references INEEL 2001 and ERDA 1977. 

[125] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the reference ERDA 1977.

[126] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the references ERDA 1977 and Rielly 1998. 

[127] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This external exposure scenario is routinely included in safety analyses of this type of facility, 

as described in section 6.3.4 of the ORAUT-TKBS-0007-6 INL – Occupational External 

Dosimetry TBD. 

[128] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003.

This information is based on a review of the references SCA, SRA, and WCC 1994 and DOE 

1991a.
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 72 of 91

[129] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the references SCA, SRA, and WCC 1994 and DOE 

1991a.

[130] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the references SCA, SRA, and WCC 1994 and DOE 

1991a.

[131] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the references Bunch 1968, Hawley et. al. 1964 and 

Zimbrick and Voillequé 1969.

[132] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the references Bunch 1968, Hawley et. al. 1964 and 

Zimbrick and Voillequé 1969.

[133] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the references Bunch 1968, Hawley et. al. 1964 and 

Zimbrick and Voillequé 1969.

[134] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the reference Rielly 1998.

[135] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the reference Rielly 1998.

[136] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the reference Rielly 1998.

[137] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the references Mobley and Keller 1991, Thiel 1997, 

and Mobley 1987.

[138] Frank Hinckley. Principal Health Physicist. Intrepid Technology and Resources, Inc. JuneAugust 2003. 

This information is based on a review of the references Mobley and Keller 1991, Thiel 1997, 

and Mobley 1987.
Document No. ORAUT-TKBS-0007-2 Revision No. 03 Effective Date: 08/17/2007 Page 73 of 91

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Andersen, B. L., O. R. Perry, and P. E. Ruhter, 1995, Technical Basis for Internal Dosimetry at 

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ANL-6024, University of Chicago, Lemont, Illinois, June. [SRDB Ref ID: 1088]

Bunch, D. F., editor, 1966, Controlled Environmental Radioiodine Tests, Progress Report Number 

Two, IDO-12053, Idaho Falls, Idaho, August.
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Bunch, D. F., editor, 1968, Controlled Environmental Radioiodine Tests, Progress Report Number 

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Casto, W. R., 1980, “ICPP Criticality Event of October 17, 1978,” Nuclear Safety, volume 21, number 

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Cederberg, C. K., H. V. Chamberlain, W. F. Holcomb, R. L. Nebeker, W. P. Palica, R. R. Smith, and 

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Cipperley, F. V., 1958, Personnel Metering Operating Techniques, Self-Service System, Effective 

August 20, 1951 through March 29, 1958, IDO-12007, U.S. Atomic Energy Commission, Idaho 

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Cipperley, F. V., 1968, Personnel Dosimetry Standard Operating Procedure Effective March 30, 1958, 

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Cordes, O. L., D. F. Bunch, J. R. Fielding, and J. K. Warkentin, 1967, Radiological Aspects of the 

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Cordes, O. L., R. P. Bird, G. A. Dineen, and J. R. Fielding, 1965, Radiological Aspects of SNAPTRAN 

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Station, Idaho Falls, Idaho, January. [SRDB Ref ID: 1122]

Decker, L. A., 1991, “Submittal of Report on the Investigation of the ICPP Custom Processing 

Dissolver Explosion, February 9,1991,” letter to A. A. Pitrolo (U.S. Department of Energy, 

Idaho Operations Office), LAD-04-91, Westinghouse Idaho Nuclear Company, Idaho National 

Engineering Laboratory, Idaho Falls, Idaho, June 3. [SRDB Ref ID: 26109]

Dietrich, J. R., 1956, "Experimental Determinations of the Self-Regulation and Safety of Operating 

Water-Moderated Reactors," P/481, Proceedings of the International Conference on the 

Peaceful Uses of Atomic Energy, Geneva, 8 August-20 August 1955, volume 13, pp. 88–101.

Dietrich, J. R., H. V. Lichtenberger, and W. H. Zinn, 1956, "Design and Operating Experience of a 

Prototype Boiling Water Power Reactor," P/851, Proceedings of the International Conference 

on the Peaceful Uses of Atomic Energy, Geneva, 8 August-20 August 1955, volume 3, pp. 56–

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DOE (U.S. Department of Energy), 1991a, 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]
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DOE (U.S. Department of Energy), 1991b, Idaho National Engineering Laboratory Historical Dose 

Evaluation Quality Assurance Files, DOE/ID-12119-QAF-001-128, Idaho Falls Office, Idaho 

Falls, Idaho. [SRDB Ref ID: 1130]

DOE (U.S. Department of Energy), 1996, Highly Enriched Uranium Working Group Report - Volume 

II: Number 5, Idaho National Engineering Laboratory Working Group and Site Assessment 

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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.

activation

Creation of a radioisotope by interaction of a stable (nonradioactive) element with neutrons, 

protons, or other types of radiation.

albedo dosimeter

Thermoluminescent dosimeter that measures the thermal, intermediate, and fast neutrons 

scattered and moderated by the body or a phantom from an incident fast neutron flux.

americium–beryllium (AmBe)

Common neutron source created by an alpha particle from 241Am interacting with beryllium to 

produce a large neutron yield with low gamma-ray yield.

annual dose equivalent

Dose equivalent received in a year in units of rem or sievert. See dose. 

anticontamination clothing (anti-Cs) 

Special clothing worn by personnel for protection from radiological contamination.

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

beta (β) dose

Designation (i.e., beta) on some records for external dose from beta and less-energetic X-ray 

and gamma radiation, often for shallow dose or dose to the lens of the eye. 

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.

blowdown

Sudden depressurization from a break in a pipe containing pressurized water in a reactor 

system. 

boiling-water reactor (BWR) 

Nuclear reactor in which boiling water in the core serves as both coolant and moderator. 
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breeder reactor

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

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

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.

collective dose equivalent

Sum of the dose equivalents of all individuals in an exposed population in units of person-rem 

or person-sievert. See dose. 

collective effective dose equivalent

Sum of the effective dose equivalents of all individuals in an exposed population in units of 

person-rem or person-sievert. See dose. 

containment

(1) State of being contained. (2) Heavy structure completely surrounding a source of 

radioactivity to prevent it from escaping or affecting the environment.

contamination

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

control rod

Neutron-absorbing device in a reactor used to slow or speed the reaction. Also called safety 

rod.

control room

Room in a nuclear reactor facility that houses the controls for operation and monitoring. 

coolant

Gas or liquid in a nuclear reactor that removes the heat from the fission process. 

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.

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.
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decommissioning

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

specified levels and often involving demolition of the facility.

decontamination

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

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

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

depleted uranium

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.

dose 

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays. Various terms narrow the type of dose, 

and some are additive: 

Absorbed dose is the amount of energy deposited in a substance by ionizing radiation.

Collective dose is the sum of the doses to a specific population.

Committed dose is the dose over time (usually 50 years for workers) to a specific organ or 

tissue from an intake of radioactive material.

Cumulative dose is the sum of all doses to the same portion of the body or to the whole 

body over time.

Deep dose is the dose at a 1-centimeter depth in tissue (1,000 milligrams per square 

centimeter).

Effective dose is the sum of the equivalent doses in the principal tissues and organs of the 

body, each weighted by a tissue weighting factor that accounts for the probabilities of 

fatal and nonfatal cancers according to severity and the average length of life lost due 

to an induced cancer. It indicates the biological effect of the radiation exposure in that 

tissue. 

Equivalent dose or dose equivalent is the absorbed dose in a tissue or organ multiplied by 

a weighting factor for the particular type of radiation. 

Organ dose is the dose to a specific organ.

Penetrating dose is that from higher energy photon (gamma and X-ray) radiation and 

neutron radiation that penetrates the outer layers of the skin. Nonpenetrating dose is 

that from beta and lower energy photon radiation. 

Personal dose equivalent is the dose equivalent in soft tissue below a specified point on 

the body at a specified depth. 
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Shallow dose is the dose at a 0.07-millimeter depth in tissue (7 milligrams per square 

centimeter).

Skin dose is the dose to the skin.

Whole-body dose is the dose to the entire body excluding the contents of the 

gastrointestinal tract, urinary bladder, and gall bladder.

dose equivalent (H)

Product of absorbed dose in units of rem or sievert in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose. 

dose of record

(1) Dose records that the U.S. Department of Energy provided to the National Institute for 

Occupational Safety and Health as part of each worker’s file. (2) Individual recorded dose 

such as that on a dosimetry card or in a dosimetry database.

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. See albedo dosimeter, pocket ionization chamber, and 

thermoluminescent dosimeter. 

dosimetry

Measurement and calculation of internal and external radiation doses.

emergency core cooling system 

Backup system that injects coolant into the core of a reactor to prevent overheating of the fuel 

and subsequent fuel damage. See meltdown. 

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.

excursion

Planned or accidental increase in the normal operating power level of nuclear reactions.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

extremities 

The portion of the arm from and including the elbow through the fingertips and the portion of 

the leg from and including the knee and patella through the toes. 

film

Radiation-sensitive photographic film in a light-tight wrapping.
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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.

flux

Rate of flow of mass, volume, or energy. Flux is often used to mean flux density. 

flux density

Flux per unit of cross-sectional area perpendicular to the direction of flow. In relation to health 

physics, usually describes the flow of neutrons or photons in a radiation field or X-ray beam.

fuel assembly

Arrangement of nuclear fuel for use in a nuclear reactor. Fuel is most commonly in the form of 

pellets arranged in rods of various diameters but can also be formed into plates or other 

shapes.

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. 

gray (Gy) 

International System unit of absorbed radiation dose, which is the amount of energy from any 

type of ionizing radiation deposited in any medium; 1 Gy equals 1 joule per kilogram or 

100 rads. 

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. 

highly enriched uranium (HEU) 

Uranium enriched to at least 20% 235U for use as fissile material in nuclear weapons 

components and some reactor fuels. Also called high-enriched uranium.

 

hot cell

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

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

materials designed to protect workers from radiation. 
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hot run

Operational or test run of a chemical process and equipment using radioactive materials.

interim storage

Storage between use and final disposition.

ionization chamber

Device that measures exposure or radiation dose. 

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. See 

beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray radiation. 

irradiate

To expose to ionizing radiation.

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.

megaPrefix that multiplies a unit by 1 million (1 × 106

).

meltdown

Melting of nuclear reactor fuel caused by a failure of the coolant to adequately carry away 

heat.

microPrefix that divides a unit by 1 million (multiplies by 1 × 10-6).

milliPrefix that divides a unit by 1,000 (multiplies by 1 × 10-3).

mixed waste

Unwanted material containing both radioactive and hazardous components. 

moderator

Material such as water, heavy water, or graphite that slows the neutrons in a reactor to cause 

more fission.

natural uranium

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). See uranium. 
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neutron

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.

neutron, fast

Neutron with energy equal to or greater than 10 keV.

neutron, thermal

Strictly, neutron in thermal equilibrium with surroundings. In general, neutrons with energy 

less than about 0.5 eV.

neutron radiation 

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron. 

nuclear energy

Energy released by nuclear reaction, some of which can be ionizing radiation. Of particular 

importance is the energy released when a neutron initiates fission or when two nuclei join 

together under millions of degrees of heat (fusion). Also called atomic energy.

nuclear power plant

Electrical generating facility using nuclear fuel.

nuclear waste 

See radioactive waste. 

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.

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.

personal dose equivalent [Hp(d)] 

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection. 
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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. See photon radiation. 

photon radiation 

Electromagnetic radiation of light energy (photons) from microwaves to gamma rays. Gamma 

rays and X-rays are examples of ionizing photon radiation, which have enough energy to 

penetrate matter, including the body, and deposit energy in that matter. 

pocket dosimeter 

See pocket ionization chamber. 

pocket ionization chamber (PIC) 

Cylindrical monitoring device commonly clipped to the shirt or laboratory coat pocket to 

measure ionizing radiation. Also called pencil, pocket pencil, pencil dosimeter, and pocket 

dosimeter.

pressurized-water reactor (PWR) 

Nuclear reactor in which water is kept under pressure in a vessel to prevent boiling. 

primary loop

Closed experimental system through which coolant flows as part of the control for a nuclear 

reaction using the main reactor as the primary source for neutron flux. See secondary loop. 

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. 

Quality factor, Q

Historical value assigned to reflect the average effectiveness of a particular kind of radiation in 

producing biological effects in humans, now called radiation weighting factor. The quality 

factor multiplied by the absorbed dose yields the dose equivalent. See dose. 

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joule per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose. 

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. 

radioactive 

Giving off ionizing radiation such as alpha particles or X-rays. 
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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. 

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. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

radioactive lanthanum (RaLa) 

Isotope 140La, which was tested for application in nuclear weapons.

reactor 

Device in which a fission chain reaction occurs under controlled conditions to produce heat or 

useful radiation for experimental purposes or to generate electrical power or nuclear fuel. 

reactor vessel 

Structure enclosing the fuel elements, control elements, coolant piping, and other structures 

that support the reactor core. 

reflector

Part of the structure of some nuclear reactors that reflects neutrons back into the reactor core. 

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. 

retention basin

Outdoor basin (any of several designs) in which liquid solutions are deposited and held 

pending evaporation or the precipitation of solids. 

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.

roentgen (R )

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts). 
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secondary loop

System in a reactor that receives transfer heat from a primary system. The secondary system 

is not exposed to the reactor core and is free of radioactivity under normal circumstances. 

See primary loop. 

semiscale

Informal name of a scale model of a nuclear reactor operated at INL. The core simulated the 

heat of a nuclear reaction by electrical means and was used to study the behavior of water 

and steam in accidents involving the loss of coolant caused by a break in a coolant system.

scram

Sudden shutdown of the nuclear reaction in a reactor, typically caused by rapid insertion of 

control/safety rods, when a dangerous or undesirable condition occurs.

shallow dose equivalent [SDE, Hs, Hp(0.07)] 

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

sievert (Sv)

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

skin dose

See shallow dose equivalent. 

spent nuclear fuel

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

a chain reaction.

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. 

thermoluminescence

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose. 

transuranic (TRU) elements 

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium. All isotopes of the transuranic elements are radioactive, they are naturally either 
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rare or nonexistent on Earth, and most are known only as a result of research using nuclear 

reactors and particle accelerators because of extremely short half-lives. 

transuranic waste 

Radioactive waste that contains transuranic elements and has radioactivity of 100 or more 

nanocuries per gram.

tritium

Radioactive isotope of hydrogen that contains one proton and two neutrons in its nucleus. 

Because tritium is chemically identical to the natural hydrogen atoms in water, it can be easily 

ingested. It decays by beta emission and has a radioactive half-life of about 12.5 years. Its 

uses include increasing the yield of a nuclear weapon.

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.

uranium (U) 

Heavy, metallic, and radioactive element with atomic number 92. Most natural uranium as 

found in ores is 238U with trace levels of other isotopes. Uranium-235 (0.7% of natural 

uranium) is fissile by itself and used in nuclear weapons as well as reactors. Uranium-238 

(99.3% of natural uranium) is fissionable by fast neutrons and used in nuclear reactors. 

Natural uranium contains a minute amount of 234U. See depleted uranium, enriched uranium, 

highly enriched uranium, and natural uranium. 

uranium oxide

Metallic compound of uranium and oxygen useful as nuclear fuel because it has a higher 

melting point than metallic uranium. However, its heat transfer properties are not as efficient 

as those of metallic uranium.

waste storage tank

Holding tank for liquid or gaseous wastes that might or might not be radioactive.

water-moderated reactor

Reactor in which water slows the speed of neutrons from fissioning atoms to increase the 

number of neutrons that cause fission. 

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose. 

X-ray 

See X-ray radiation. 
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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. 

zero power

Reactor power level that maintains a chain reaction at an extremely low power level producing 

very little heat. Also called low power.

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.