ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 31

Document Title:

Argonne National Laboratory – West –

Site Description

Document Number: ORAUT-TKBS-0026-2

Revision: 01

Effective Date: 06/22/2007

Type of Document TBD

Supersedes: Revision 00

Subject Expert(s): Clark B. Barton

Site Expert(s): Earl Graham, Norman D. Rohrig

Approval: Signature on File Approval Date: 06/19/2007

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 06/19/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 06/19/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 06/20/2007

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 06/22/2007

Brant A. Ulsh, PhD CHP

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 2 of 31

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

06/21/2004 00-A New technical basis document for the ANL-W Site Facilities and 

Processes. Initiated by Norman D. Rohrig.

08/17/2004 00-B Incorporates NIOSH review comments. Initiated by Norman D. 

Rohrig.

09/09/2004 00 First approved issue. Initiated by Norman D. Rohrig.

06/22/2007 01 Approved revision initiated for biennial review. Added Purpose, 

Scope, and Attributions and Annotations sections. Updated 

References and formatting. Incorporates formal internal review 

comments. No further changes occurred as a result of formal NIOSH 

review . Constitutes a total rewrite of document. This revision 

results in no change to the assigned dose and no PER is required. 

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

Ann M. Jenkins.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 3 of 31

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

2.1 Introduction ..............................................................................................................................7

2.1.1 Purpose........................................................................................................................8

2.1.2 Scope...........................................................................................................................8

2.2 Site Activities and Processes, February 1951 to February 2005...............................................8

2.2.1 Experimental Breeder Reactor No. 1, April 1951 to 

December 30, 1963.................................................................................................... 10

2.2.2 Boiling-Water Reactor Experiment No. 1, Late 1953 to July 

22, 1954 .....................................................................................................................11

2.2.3 Boiling-Water Reactor Experiment No. 2, October 19, 1954, 

to March 1955 ............................................................................................................ 11

2.2.4 Boiling-Water Reactor Experiment No. 3, June 9, 1955, to 

December 1956.......................................................................................................... 12

2.2.5 Boiling-Water Reactor Experiment No. 4, December 3, 1956, 

to June 1958...............................................................................................................12

2.2.6 Boiling-Water Reactor Experiment No. 5, February 9, 1962, 

to September 1964..................................................................................................... 13

2.2.7 Zero Power Reactor No. 3, October 1955 to November 1970..................................... 13

2.2.8 Argonne Fast Source Reactor, October 29, 1959, to Late 

1970s .........................................................................................................................13

2.2.9 Transient Reactor Test and Experiment Facility, February 

23, 1958, to April 1994 ............................................................................................... 14

2.2.10 Experimental Breeder Reactor No. II, May 1961 to 

September 30, 1994................................................................................................... 14

2.2.11 Hot Fuel Examination Facility, 1964 to 2006............................................................... 15

2.2.12 Zero Power Plutonium (Physics) Reactor, April 18, 1969, to 

April 1992 (standby) ................................................................................................... 16

2.2.13 Neutron Radiography Facility, October 1, 1977 to 2006.............................................. 16

2.2.14 Fuel Assembly and Storage Building, 1970 to present................................................ 17

2.2.15 Other ANL-W Support Facilities.................................................................................. 17

2.3 Attributions and Annotations ..................................................................................................17

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

Glossary ............................................................................................................................................26
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 4 of 31

LIST OF TABLES

TABLE TITLE

2-1 Radionuclides of concern for all reactors................................................................................

PAGE

10

LIST OF FIGURES

FIGURE TITLE

2-1 Idaho National Laboratory showing ANL-W facilities ................................................................

PAGE

9
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 5 of 31

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

BORAX Boiling-Water Reactor Experiment 

BWR boiling-water reactor

CAM continuous air monitor

CFA Central Facilities Area

Ci curie

cm centimeter

DOE U.S. Department of Energy

DOE-ID DOE Idaho Operations Office

DU depleted uranium

EBR Experimental Breeder Reactor

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

FASB Fuel Assembly and Storage Building

FMF Fuel Manufacturing Facility

ft foot

g gram

gal gallon

HFEF Hot Fuel Examination Facility

hr hour

IDO Idaho Operations Office

IMBA Integrated Modules for Bioassay Analysis

in. inch

INL Idaho National Laboratory

IREP Interactive RadioEpidemiological Program

IWP Industrial Waste Pond

kg kilogram

kW kilowatt

m meter

MAP mixed activation product

MFP mixed fission product

mi mile

min minute

mR milliroentgen

mrem millirem

MW megawatt
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 6 of 31

n neutron

NIOSH National Institute for Occupational Safety and Health

NRAD Neutron Radiography Facility

NRTS National Reactor Testing Station

ORAU Oak Ridge Associated Universities

POC probability of causation

psi pounds per square inch

psig pounds per square inch gauge

RAM radiation (or remote) area monitor

s second

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TLD thermoluminescent dosimeter

TREAT Transient Reactor Experiment and Test (facility)

U.S.C. United States Code

yd yard

ZPPR Zero Power Plutonium (later Physics) Reactor

ZPR Zero Power Reactor

μCi microcurie

°C degree Celsius

§ section or sections
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 7 of 31

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-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 8 of 31

2.1.1

The purpose of this TBD is to assist in the evaluation of worker dose from Argonne National 

Laboratory – West (ANL-W) processes using the methodology in External Dose Reconstruction 

Implementation Guideline (NIOSH 2006) and Internal Dose Reconstruction implementation Guideline 

(NIOSH 2002). 

Purpose

2.1.2

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

claimants, the total ANL-W occupational dose that can be reasonably associated with worker radiation 

exposure as covered in the EEOICPA legislation. This TBD describes facilities, processes, and 

historical information in relation to worker internal and external exposures for use when actual 

monitoring data might be unavailable. The information in this TBD is primarily from the INL Site 

Description TBD (ORAUT 2005).

Scope

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

justification, or clarification of the associated information, are presented in Section 2.3.

2.2 SITE ACTIVITIES AND PROCESSES, FEBRUARY 1951 TO FEBRUARY 2005

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

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

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. In 1974, the 

NRTS became the Idaho National Engineering Laboratory and, in 1997, the Idaho National 

Engineering and Environmental Laboratory. In 2005, the site became the Idaho National Laboratory 

(INL). 

ANL-W operated on the INL site, an 890-mi2 (572,000-acre) reservation with maximum dimensions 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. The site is on the Snake River Plain of southeastern Idaho at an elevation of about 5,000 ft. 

INL played a major role in early reactor research and development; ANL-W performed much of this 

work. The INL site has operated 52 reactors, about 12 of them by ANL-W, plus fuel handling and 

reprocessing and radioactive waste storage and disposal facilities since it began operations in 1949

(Stacy 2000). 

ANL-W was originally known as the Idaho Division of Argonne National Laboratory (ANL). It was an 

extension of what is now known as Argonne National Laboratory – East (ANL-E) near Chicago, 

Illinois. ANL is a research laboratory that has been operated by the University of Chicago throughout 

its history. The original ANL-W site is in the southwest portion of INL, 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 ANL-W location is now a National Historic Landmark and is the site 

of the decommissioned Experimental Breeder Reactor No. 1 (EBR-I), Boiling-Water Reactor 

Experiment (BORAX), Argonne Fast Source Reactor (AFSR), and Zero Power Reactor No. 3 (ZPR-III) 

facilities. 

The more recent 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, ANL-W was absorbed into the 
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 9 of 31

INL during a contract reconfiguration. Figure 2-1 shows the relative location of each facility or 

technical area discussed in this TBD, including the EBR-II, Transient Reactor and Experiment Test 

(TREAT) facility, and Zero Power Plutonium Reactor (ZPPR). ANL-W operated the EBR-I and 

BORAX areas before their shutdowns. Other INL facilities that are indicated on Figure 2-1 include the 

Auxiliary Reactor Area, Central Facilities Area (CFA), Idaho Chemical Processing Plant (now the 

Idaho Nuclear Technology and Engineering Complex), Radioactive Waste Management Complex,

Special Power Excursion Reactor Test area, Grid III, Test Reactor Area, Experimental Field Station,

and Naval Reactors Facility (Hoff, Chew, and Rope 1987, p. 46).

Figure 2-1. Idaho National Laboratory showing ANL-W facilities.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 10 of 31

Dosimetry services at INL including ANL-W were unique among DOE facilities in that DOE Idaho 

Operations Office (IDO) personnel operated and provided internal and external dosimetry services. 

The IDO Dosimetry Branch provided and analyzed external dosimetry badges, counted workers in the 

whole-body counter, and analyzed bioassay samples. At first, IDO personnel at the Radiological 

Environmental Sciences Laboratory in Building CF-690 in the CFA maintained exposure histories of 

personnel based on dosimetry records, including bioassay data. With the advent of the DOE 

Laboratory Accreditation Program, dosimetry responsibility transferred to the prime contractor on 

January 2, 1989, to eliminate a conflict of interest on the part of IDO. 

Personnel working at ANL-W in designated radiological areas were typically required to wear state-ofthe-art dosimetry [film badges, thermoluminescent dosimeters (TLDs), personal ion chambers, etc.] as 

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

were monitored by remote and portable radiation/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]. If airborne radioactivity might have been present or if internal exposure was 

possible, applicable respirators were provided to prevent or reduce internal exposure [3]. 

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

airborne radioactivity and direct radiation exposure. Bioassay programs were instituted to monitor 

and assess potential internal exposures. Environmental monitoring systems were placed around the 

site at multiple locations to measure direct radiation, fallout, and effluent discharges or releases to 

onsite personnel and members of the public. There have been film or TLD badges in specific building 

areas for each facility and around perimeter fences to measure direct environmental radiation 

accumulation as a check and balance on source terms within site locations [4]. 

The following sections describe the facilities and the associated operations for ANL-W. Table 2-1 lists 

common radionuclides for reactors.

Table 2-1. Radionuclides of concern for all reactors [5].

Am-241 Co-60 I-131 Ru-103

Ba-140 Cr-51 I-133 Ru-106

Ce-141 Cs-134 La-140 Sr-90/Y-90

Ce-143 Cs-137 Mn-54 Sr-91

Ce-144 Eu-152 Nb-95 Sr-92

Cm-242 Eu-154 Np-239 U-234

Cm-244 Fe-59 Pu-238 Zn-65

Co-58 H-3 Pu-239 Zr-95

2.2.1

The date information is from Stacy (2000, p. 262). EBR-I, the first reactor built at INL, was a NaKcooled, solid-fuel (enriched uranium), unmoderated heterogeneous fast reactor designed for fullpower 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 238U around the core provided the fertile material 

in which nuclear material breeding took place. 

Experimental Breeder Reactor No. 1, April 1951 to December 30, 1963

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 of EBR-I began in May 1949 and 

was complete in April 1951. Reactor startup occurred on August 24, 1951.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 11 of 31

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 (Zinn 1956; Stacy 2000, p, 135). After the partial meltdown, the core 

assembly was removed from the reactor through a temporary cave constructed 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 (Stacy 2000). 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 with a half-life of 15 hr. The saturation 

level at full-power operation was approximately 24 μCi/g. The second most significant activity was 

about 2 μCi/g 137Cs, which apparently entered the system during and after the meltdown incident. No 

other long-lived activity was identified in the primary coolant. Short-lived 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 (MFPs) and 

mixed activation products (MAPs) [6].

External exposure

2.2.2

occurred from MFPs and MAPs during activities associated with reactor operation 

and maintenance [7].

The date information is from Stacy (2000, p. 261). BORAX-I was an open-top, water-cooled, watermoderated, boiling-water reactor (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 approximately 0.5 mi northwest of EBR-I and was housed in a 10-ft-diameter tank open to the 

atmosphere. The control room was approximately 0.5 mi away near EBR-I. 

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 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. 

Detailed discussions of this incident are available (Griffiths, Sill, and Wilhelmsen 1956; Brodsky and 

Beard 1960).

Internal exposure might have occurred from airborne radioactivity during operation and other 

associated activities, core destruction, coolant, and cleanup activities [8].

External exposure

2.2.3

occurred from direct radiation associated with the reactor operation and 

maintenance activities from MFPs and MAPs [9].

The date information is from Stacy (2000, p. 261). BORAX-II replaced BORAX-I as a new reactor that 

would more closely approximate the 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. It was significantly larger than BORAX-I. 

The vessel was shielded by concrete and housed in a sheet-metal building. Tests of new core 

combinations used various enrichments of 235U in metal fuel plates (AEC 1962). The boiling-water 

system operated at 300 psi, which essentially made BORAX-II a power experiment. The power level 

was about 6.4 MW-thermal but, because it had no turbine generator, it produced no electricity. The 

energy produced was released in the form of steam (Dietrich 1956).

Boiling-Water Reactor Experiment No. 2, October 19, 1954, to March 1955
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 12 of 31

Internal exposure was possible from MAPs associated with activities involving the reactor coolant [10]. 

External exposure

2.2.4

occurred during routine operations and loading and unloading of fuel [11].

The date information is from Stacy (2000, p. 261). A program began in March 1955 to modify the 

BORAX-II reactor to produce electrical energy. 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 at the NRTS, 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 activities involving the reactor coolant [12]. 

Fuel cladding failure occurred, and resulting radionuclides were identified in the reactor water (Zinn et 

al. 1956).

External exposure

2.2.5

occurred from routine operation and loading and unloading of fuel [13]. 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 (~104 higher activity) (Zinn et al. 1956). 

Nitrogen-16 was identified as the principal radiation source in the coolant (Dietrich, Lichtenberger, and 

Zinn 1956). 

The date information is from Stacy (2000, p. 261). BORAX-IV, the successor to BORAX-III, began 

operation in December 1956. This reactor, with a design power of 20 MW-thermal, 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 atmospheric pressure. BORAX-IV operated with a core of uranium-thorium 

fuel elements until April 17, 1957. Beginning in May 1957, it 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 identify defective elements in the core

(Robertson and Hall 1959). BORAX-IV released approximately 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):

Boiling-Water Reactor Experiment No. 4, December 3, 1956, to June 1958

• Radiation levels of the steam plant equipment

• Quantitative determination of fission gases 138Xe and 88Kr, which were released to the 

atmosphere through the air ejector

• Analysis of reactor water, condensed steam before the turbine, and condensed steam after the 

turbine (hot-well condensate) for fission products

• Area contamination downwind from the reactor
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 13 of 31

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 and MFPs associated with the reactor 

coolant [14].

External exposure

2.2.6

occurred during operation and loading and unloading of fuel [15]. 

The date information is from Stacy (2000, p. 261). BORAX-V was a flexible BWR with the same 

configuration as BORAX-IV and was 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 from ceramics of uranium and thorium and from associated 

maintenance work [16].

External exposure

2.2.7

occurred from routine activities associated with reactor operation and maintenance

[17].

The date information is from Stacy (2000, p. 268). ZPR-III was a low-power, split-table 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 that 

contained the control room, workroom, vault, laboratory rooms, offices, etc. The assembly (reactor) 

room of reinforced concrete was approximately 45 by 42 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. Workers could swing a hinged platform into place 

between the halves on which they could stand while loading or unloading the machine.

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 containers that 

limited the storage density to 2 kg 235U/ft3

.

Internal exposure potential existed from possible airborne radioactivity during maintenance operations 

from MFPs and MAPs [18].

External exposure

2.2.8

occurred during maintenance activities and during loading and unloading of fuel

[19].

The date information is from Stacy (2000, p. 261). AFSR was a small fast reactor facility designed to 

produce neutrons for the development of special equipment for the EBR-I, EBR-II, and ZPR-III fast 

reactor programs. AFSR, built in 1954 with a design power of 1 kW, was in a prefabricated Butlertype building with its own heating and air compressor plant near the EBR-I facility. No water was 

Argonne Fast Source Reactor, October 29, 1959, to Late 1970s
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 14 of 31

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 experiments, was built 

around a cylindrical core of solid highly enriched uranium with a blanket of solid depleted uranium 

(DU; Brunson 1959). Reactor startup occurred on October 29, 1959; the reactor was operational until 

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.

Internal exposure might have occurred during routine operations that could create airborne 

radioactivity [20]. 

External exposure

2.2.9

occurred from maintenance activities and loading and unloading of fuel [21].

The following information is from Stacy (2000, p. 268). TREAT was an air-cooled thermal 

heterogeneous system designed to evaluate reactor fuels and other material under conditions that 

simulated various types of reactor excursions. Construction began in February 1958 and ended in 

November 1958, and criticality occurred 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 and Experiment 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 0.5 mi from the 

reactor that contains control panels and necessary instrumentation for remote control of the reactor.

Shielding permitted personnel access around and atop 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 of all personnel. General neutron and gamma 

radiation levels at a distance of 10 ft from the reactor during operations at 100 kW were (Freund et al. 

1960):

• Fast neutrons Negligible

• Thermal neutrons 50 to 1,500 n/cm2

/s

• Gamma 5 to 8 mrem/hr

Internal exposure might have occurred during routine operations that could create airborne 

radioactivity; however, it was not expected to occur [22].

External exposure

2.2.10

occurred from routine operations [23]. 

The date information 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-thermal 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 and the feasibility of a central power station and onsite fuel 

processing. It met these objectives in the late 1960s, and its role 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 

Experimental Breeder Reactor No. II, May 1961 to September 30, 1994
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 15 of 31

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 are drained, and the power plant is depressurized.

The reactor was submerged in a primary tank that contained 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 was designed to contain the energy release associated with a 

reactor incident. The reactor building was designed to confine the effects of a maximum sodium-air 

interaction caused by a major sodium release. The reactor consisted of an enriched core surrounded 

on all sides by a fertile blanket of DU (McVean et al. 1962; Koch et al. 1957).

The Sodium Plant contains 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 was 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 stored 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.

2.2.11

The Hot Fuel Examination Facility (HFEF) complex consists of 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 2006

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 and 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 (Baca 1979). 

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 (Baca 1979). HFEF/North contains two hot cells, one with 

an argon gas atmosphere and the other with an air atmosphere. The air atmosphere cell is 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 the determination of 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 in New Mexico for 

disposal takes place in the Waste Characterization Area of the HFEF high bay.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 16 of 31

Internal exposure might have occurred during cell entries when suspended radioactive contamination 

materials could cause airborne radioactivity from MFPs and MAPs [24]. 

External exposure

2.2.12

occurred when entries to the hot cell were made after experiment processes or 

during equipment maintenance and refurbishment [25].

The date information 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 Plutonium (Physics) Reactor, April 18, 1969, to April 1992 (standby)

The basic element of the ZPPR is a bed-and-table system that holds the matrix assembly. The two 

tables, one moveable and one stationary, are supported on a cast steel bed. 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 DU with a protective film [26].

Internal exposure potential was minimal due to the use of hoods and other protective equipment [27]. 

External exposure

2.2.13

occurred from working with reactor processes, loading and unloading of fuel, etc.

[28].

The Neutron Radiography (NRAD) Facility is a 250-kW, steady-state Training, Research, Isotope,

General Atomics 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 water-filled 

tank. NRAD began operation in March 1978, with two radiography stations (Richards and McClellan 

1979). The East station services the hot cell complex where specimens can be radiographed without 

removal from the hot cell environment (Richards and McClellan 1979). 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 (Richards and McClellan 1979). Cask handling 

and specimen shielding allow for full-size reactor assemblies. The radiography room is easily 

accessible for development work (Richards and McClellan 1979).

Neutron Radiography Facility, October 1, 1977 to 2006

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 an 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 MAPs [29].

External exposure occurred during sample handling and maintenance associated with radioactive 

samples. Remote handling techniques are used to minimize dose [30].
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 17 of 31

2.2.14

The Fuel Assembly and Storage Building (FASB) is a multipurpose facility about 100 yd northeast of 

EBR-II 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 preparing metallurgical samples and 

inert atmosphere gloveboxes and hoods. The facility ceased fuel assembly about 1990, but other 

radiological work is ongoing. 

Fuel Assembly and Storage Building, 1970 to present

Internal exposure might have occurred from airborne radioactivity associated with the described 

processes from the uranium fuel and spent fuel examinations [31]. 

External exposure

2.2.15

occurred from movement of radiological samples and reactor fuel [32].

At ANL-W, an Analytical Laboratory provides the capability for performing chemical and physical 

measurements of radioactive and nonradioactive samples. This facility includes six analytical hot 

cells (the Junior Cave area) and general and specialized chemistry laboratories. Personnel were 

subjected to radiation levels in the multiroentgen-per-hour range on occasion in the Junior Cave area

[33]. 

Other ANL-W Support Facilities

The Radioactive Scrap and Waste Facility provides in-ground retrievable dry storage for nuclear fuels 

and other highly radioactive scrap and waste as well as interim storage for EBR-II spent fuel [34].

The Radioactive Liquid Waste Treatment Facility evaporates low-level radioactive liquid waste 

generated at ANL-W facilities into solidified residue packaged in shielded containers [35].

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 boiler blowdown; water from once-through air conditioning, and 

cooling water from other sources. There might be inadvertent low-level radioactive contamination in 

this pond [36].

The three Sanitary Sewage Treatment Ponds north of EBR-II cover an area of about 2 acres. These 

ponds are not suspected to be radiologically contaminated [37].

Internal exposure is possible from airborne radioactivity associated with the various types of samples 

worked in these facilities [38].

External exposure

2.3 ATTRIBUTIONS AND ANNOTATIONS

occurred from working with the variety of radioactive materials associated with 

these facilities (MFPs, MAPs, and transuranic elements) [39].

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations to assist in the process of worker dose 

reconstruction. These callouts are listed in this section of the document with information provided to 

identify the source and justification for each item. Conventional references are provided in the next 
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 18 of 31

section of this document that link data, quotations, and other information to documents available for 

review on the Oak Ridge Associated Universities (ORAU) Team servers.

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

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

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

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

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

document, including all findings and conclusions. In all cases where such information or previous 

studies or writings are included or relied upon by the Document Owner, those materials are fully 

attributed to the source.

Earl Graham served as one of the initial Subject Experts for this document. Mr. Graham was 

previously employed at ANL-W 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. Graham, those materials are 

fully attributed to the source.

[1] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This statement by Mr. Graham is based on information from a review of Cipperley (1968), 

Andersen, Perry, and Ruhter (1995), and INEEL (2001).

[2] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This statement by Mr. Graham is based on information from a review of Andersen, Perry, and 

Ruhter (1995) and INEEL (2001).

[3] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This statement by Mr. Graham is based on information from a review of Andersen, Perry, and 

Ruhter (1995) and INEEL (2001).

[4] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This statement is based on a review of Hoff et. al. (1987).

[5] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

The radionuclides listed in Table 2-1 are mixed fission products, mixed activation products and 

corrosion products that are common to reactors. This table is based on a review of Brodsky 

and Beard (1960).

[6] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Haroldson et. al. (1963).
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 19 of 31

[7] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Haroldson et. al. (1963).

[8] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Dietrich, Lichtenberger, and Zinn (1956) and Griffiths, 

Sill, and Wilhelmsen (1956). 

[9] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Dietrich, Lichtenberger, and Zinn (1956) and Griffiths, 

Sill, and Wilhelmsen (1956). 

[10] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Dietrich, Lichtenberger, and Zinn (1956) and Griffiths, 

Sill, and Wilhelmsen (1956). 

[11] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Dietrich, Lichtenberger, and Zinn (1956) and Griffiths, 

Sill, and Wilhelmsen (1956). 

[12] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Zinn et. al. (1956).

[13] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Zinn et. al. (1956).

[14] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Robertson and Hall 1959.

[15] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Robertson and Hall 1959.

[16] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

BORAX-V.

[17] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

BORAX-V.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 20 of 31

[18] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Brittan et al. (1961).

[19] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Brittan et al. (1961).

[20] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

AFSR and a review of Brunson 1959.

[21] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

AFSR and a review of Brunson 1959.

[22] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Freund et al. (1960).

[23] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of Freund et al. (1960).

[24] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This statement by Mr. Graham is based on his understanding of the sources of exposure 

based on experience at HFEF. Because the hot cells were used to examine irradiated fuel, 

they would be contaminated with MFPs and MAPs. This contamination is a potential internal 

exposure hazard during hot cell entries.

[25] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This statement by Mr. Graham is based on his understanding of the sources of exposure 

based on experience at HFEF. Because the hot cells were used to examine irradiated fuel, 

they would be contaminated with MFPs and MAPs. This contamination is a potential external 

exposure hazard during hot cell entries.

[26] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

ZPPR and a review of Kunze and Chase (1970).

[27] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

ZPPR and a review of Kunze and Chase (1970).
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 21 of 31

[28] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at

ZPPR and a review of Kunze and Chase (1970).

[29] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This statement by Mr. Graham is based on his understanding of the sources of exposure 

based on experience at NRAD. Because the hot cells were used to examine irradiated fuel, 

they would be contaminated with MFPs and MAPs. This contamination is a potential internal 

exposure hazard.

[30] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

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 Richards and McClellan (1979).

[31] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

FASB.

[32] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

FASB.

[33] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

the Junior Cave.

[34] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of ERDA (1977).

[35] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of ERDA (1977).

[36] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

ANL-W and on a review of ERDA (1977). 

[37] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on a review of ERDA (1977).
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 22 of 31

[38] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

ANL-W.

[39] Graham, Earl. Intrepid Technology and Resources. Principal Health Physicist. June-August 

2003.

This information is based on understanding of the sources of exposure based on experience at 

ANL-W.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 23 of 31

REFERENCES

AEC (U.S. Atomic Energy Commission), 1962, Thumbnail Sketch, National Reactor Testing Station, 

Idaho Operations Office, Idaho Falls, Idaho, July. [SRDB Ref ID: 4497]

Andersen, B. L., O. R. Perry, and P. E. Ruhter, 1995, Technical Basis for Internal Dosimetry at 

LITCO, EDF # RDRU-95-3.0, LITCO Internal Report, Engineering Design File, U.S. 

Department of Energy, Idaho Falls, Idaho, January 31. [SRDB Ref ID: 13453]

Baca, J. P., 1979, "History and Status of the Fuel Examination Facility Complex," Proceedings of 27th 

Conference on Remote Systems Technology, 1979. [SRDB Ref ID: 1063]

Brittan, R. O., B. Cerutti, H. V. Lichtenberger, J. K. Long, R. L. McVean, M. Novick, R. Rice, and F. W. 

Thalgott, 1961, Hazard Evaluation Report on the Fast Reactor Zero Experiment ZPR III, ANL6408, University of Chicago, Argonne National Laboratory, Argonne, Illinois, October. [SRDB

Ref ID: 1090]

Brodsky, A., and G. V. Beard, editors, 1960, A Compendium of Information for Use in Controlling 

Radiation Emergencies, Including Lecture Notes from a Training Session in Idaho Falls, ID, 

February 12-14, 1958, TID-8206(Rev.), U.S. Atomic Energy Commission, Office of Health and 

Safety, Washington, D.C., September. [SRDB Ref ID: 1113]

Brunson, G. S., 1959, Design and Hazards Report for the Argonne Fast Source Reactor (AFSR), 

ANL-6024, University of Chicago, Lemont, Illinois, June. [SRDB Ref ID: 1088]

Cipperley, F. V., 1968, Personnel Dosimetry Standard Operating Procedure Effective March 30, 1958, 

IDO-12070, U.S. Atomic Energy Commission, Idaho Operations Office, Idaho Falls, Idaho, 

December. [SRDB Ref ID: 8086]

Dietrich, J. R., 1956, "Experimental Determinations of the Self-Regulation and Safety of Operating 

Water-Moderated Reactors," volume 13, pp. 88–101, Proceedings of the International 

Conference on the Peaceful Uses of Atomic Energy, Held in Geneva, 8 August-20 August 

1955, United Nations, New York, New York. [SRDB Ref ID: 29239]

Dietrich, J. R., H. V. Lichtenberger, and W. H. Zinn, 1956, "Design and Operating Experience of a 

Prototype Boiling Water Power Reactor," , volume 3, pp. 56–78, Proceedings of the 

International Conference on the Peaceful Uses of Atomic Energy, Geneva, 8 August-20 

August 1955, United Nations, New York, New York. [SRDB Ref ID: 29235]

ERDA (U.S. Energy Research and Development Administration), 1977, Final Environmental Impact 

Statement, Waste Management Operations, Idaho National Engineering Laboratory, ERDA1536, Assistant Administrator for Environment and Safety, Washington, D.C., September. 

[SRDB Ref ID: 1143]

Freund, G. A., D. R. MacFarlane, P. Elias, J. D. Geier, and J. F. Boland, 1960, Design Summary 

Report on the Transient Reactor Test Facility (TREAT), ANL-6034, University of Chicago, 

Argonne National Laboratory, Argonne, Illinois, June. [SRDB Ref ID: 1035]

Griffiths, P., C. W. Sill, and M. Wilhelmsen, 1956, "Radiological Monitoring of a Nuclear Release," 

volume 13, pp. 119–121, Proceedings of the International Conference on the Peaceful Uses of 
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 24 of 31

Atomic Energy, Geneva, 8 August-20 August 1955, United Nations, New York, New York. 

[SRDB Ref ID: 29243]

Handwerk, J. H., C. L. Hoenig, and R. C. Lied, 1957, Manufacture of the ThO2-U02 Ceramic Fuel 

Pellets for BORAX-IV, ANL-5678, University of Chicago, Argonne National Laboratory, 

Lemont, Illinois, August. [SRDB Ref ID: 27051]

Haroldson, R. O., F. D. McGinnis, M. Novick, R. R. Smith, and F. W. Thalgott, 1963, Safety Analysis 

Report EBR-1, Mark IV, ANL-6411, University of Chicago, Idaho Division, Argonne, Illinois, 

February. [SRDB Ref ID: 1047]

Hoff, D. L., E. W. Chew, and S. K. Rope, 1987, 1986 Environmental Monitoring Program Report for 

the Idaho National Engineering Laboratory Site, DOE/ID-12082(86), U.S. Department of 

Energy, Idaho Operations Office, Idaho National Engineering Laboratory, Idaho Falls, Idaho, 

May. [SRDB Ref ID: 1673]

INEEL (Idaho National Engineering and Environmental Laboratory), 2001, "General Technical Basis 

and Facility Specific Documents," Attachment A, INEEL M&O Contractor Technical Basis for 

Internal Dosimetry, Idaho Falls, Idaho, May 30. [SRDB Ref ID: 8479]

Kittel, J. H., M. Novick, and R. F. Buchanan, 1957, The EBR-I Meltdown - Physical and Metallurgical 

Changes in the Core, ANL-5731, University of Chicago, Argonne National Laboratory, Lamont, 

Illinois, November. [SRDB Ref ID: 1041]

Koch, L. J., H. O. Monson, D. Okrent, M. Levenson, W. R. Simmons, J. R. Humphreys, J. Haugsnes, 

V. C. Jankus, and W. B. Loewenstein, 1957, Hazard Summary Report, Experimental Breeder 

Reactor II (EBR-II), ANL-5719, University of Chicago, Argonne National Laboratory, Lemont, 

Illinois, May. [SRDB Ref ID: 1013]

Kunze, J. F., and P. L. Chase, 1970, Facility Operating Manual, Low Power Test Facility, ANCR-1137, 

Idaho Nuclear Corporation, National Reactor Testing Station, Idaho Falls, Idaho, December. 

[SRDB Ref ID: 1037]

McVean, R. L., G. S. Brunson, B. C. Cerutti, W. B. Loewenstein, F. W. Thalgott, and G. K. Whitham, 

1962, EBR-II Dry Critical Experiments, ANL-6462, University of Chicago, Argonne National 

Laboratory, Argonne, Illinois, February. [SRDB Ref ID: 1044]

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August.

NIOSH (National Institute for Occupational Safety and Health), 2006, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 2, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August 25.

Novick, M., 1958, untitled letter to J. R. Horan (U.S. Atomic Energy Commission, Idaho Operations 

Office), Argonne National Laboratory, Idaho Division, Idaho Falls, Idaho, April 1. [SRDB Ref 

ID: 1065]

ORAUT (Oak Ridge Associated Universities Team), 2005, Idaho National Engineering and 

Environmental Laboratory – Site Description, Rev. 02, Oak Ridge, Tennessee, July 29.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 25 of 31

Richards, W. J., and G. C. McClellan, 1979, “Neutron Radiography at the Hot Fuel Examination 

Facility,” Proceedings of 27th Conference on Remote Systems Technology. [SRDB Ref ID: 

1062]

Robertson, R. F. S., and V. C. Hall Jr., 1959, Fuel Defect Test – BORAX-IV, ANL-5862, University of 

Chicago, Argonne National Laboratory, Lemont, Illinois, October. [SRDB Ref ID: 27049]

Stacy, S. M., 2000, Proving the Principle, A History of the Idaho National Engineering and 

Environmental Laboratory 1949 - 1999, DOE/ID-10799, U.S. Department of Energy, Idaho 

Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1152, 1155, 1158, 1163, 1165, 1168, 

1171, 1173, 1177, 1179, 1182, 1184, 1186, 1188, 1191, 1192, 1193, 1195, 1196, 1199, 1200, 

1201, 1206, 1209, 1211, 1212, 1215]

Zinn, W. H., 1956, "A Letter on EBR-I Fuel Meltdown," Nucleonics, volume 14, number 6, pp. 35, 103, 

104, 119.

Zinn, W. H., H. Lichtenberger, M. Novick, G. K. Whitman, C. J. B. Zitek, J. G. Fields, V. C. Hall Jr., 

and R. O. Haroldsen, 1956, "Operational Experience with the Borax Power Plant," Nuclear 

Science and Engineering, volume 1, pp. 420–437. [SRDB Ref ID: 29242]
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 26 of 31

GLOSSARY

activation

Creation of a radioisotope by interaction of a stable (nonradioactive) element with neutrons, 

protons, or other types of radiation. 

activation product

Radioisotope produced through activation, usually in a nuclear reactor but also as a byproduct 

of nuclear weapons discharge and other nucleic reactions. 

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 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.

boiling-water reactor (BWR)

Nuclear reactor in which boiling water in the core serves as both coolant and moderator. 

breeder reactor

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

containment building

Building typically engineered to be air tight and made of thick heavy material to prevent the 

release of radioactive gases or radiological contamination to the atmosphere or area outside 

the containment. 

contamination, radioactive

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. 
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 27 of 31

core

Central region of a nuclear reactor where fission of the fuel takes place. 

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.

decontamination

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

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

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

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. As 

examples, spent (used) fuel elements, byproduct tails, residues from uranium isotope 

separation, and some weapons materials contain DU. DU can be blended with highly 

enriched uranium to make reactor fuel or used as a raw material to produce plutonium.

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.

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 thermoluminescent dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

enriched uranium

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7%. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade uranium 

contains greater than 90% 235U.

excursion

Planned or accidental increase in the normal operating power level of a reactor. 

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.

fast neutron

Neutron with energy equal to or greater than 10 kiloelectron-volts. This type of neutron causes 

fission in some isotopes (e.g., 238U, 239Pu).

film

Radiation-sensitive photographic film in a light-tight wrapping.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 28 of 31

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.

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. 

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 trillionths of a 

second to billions of years.

hot cell

Shielded laboratory for handling of radioactive materials with the aid of remotely operated 

manipulators. The walls and windows are made of materials that protect workers from 

radiation. 

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.

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). 
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 29 of 31

Isotopes have very nearly the same chemical properties but often have different physical 

properties.

moderator

Material such as water, heavy water, or graphite that slow 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 

microcuries per gram).

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.

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).

nuclear power plant

Electrical generating facility that uses nuclear fuel.

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.

photon

Basic unit of electromagnetic radiation. Photons are massless “packages” of light energy that 

range from low-energy microwave photons to high-energy gamma rays. Photons have 

energies between 10 and 100 kiloelectron-volts.

proton

Basic nucleic particle with a positive electrical charge and mass slightly less than that of a 

neutron. There are protons in the nuclei of every atom, and the number of protons is the 

atomic number, which determines the chemical element.

quality factor

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.
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 30 of 31

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 joules 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.

radioactive

Giving off ionizing radiation such as alpha particles or X-rays.

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. 

reactor vessel

Structure enclosing the fuel elements, control elements, coolant piping, and other structures 

that support 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.

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

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

and 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° Celsius and standard atmospheric pressure. 

An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 1 rad in soft 

tissue for higher energy photons (generally greater than 100 kiloelectron-volts). 
Document No. ORAUT-TKBS-0026-2 Revision No. 01 Effective Date: 06/22/2007 Page 31 of 31

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

sievert

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 sievert

equals 100 rem.

spent fuel

Fuel that has been in a reactor long enough to become ineffective because the proportion of 

fissile material has dropped below a certain level.

thermal neutron

Neutron in thermal equilibrium with its surroundings. The term usually refers to neutrons with 

energy less than 0.5 electron-volts. The average energy of thermal neutrons is 

0.025 electron-volts.

thermoluminescent dosimeter (TLD)

A device containing solid chips of material that when heated release the stored energy as light. 

The measurement of this light provides a measurement of absorbed dose. 

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 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 Energy Research and Development 

Administration in 1979.

X-ray radiation

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei. Once formed, there is no difference between X-rays and 

gamma rays, but gamma photons originate inside the nucleus of an atom. 

zero power

Reactor power level that maintains a chain reaction at an extremely low power level producing 

very little heat. Also called low power.