ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Argonne National 

Laboratory–West – Site Description

Document Number:

ORAUT-TKBS-0026-2

Effective Date: 09/09/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 24

Subject Expert: Norman D. Rohrig

Document Owner

Approval: Signature on File Date: 09/09/2004

Norman D. Rohrig, TBD Team Leader

Approval: Signature on File Date: 09/09/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 09/09/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 09/09/2004

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

3

Acronyms and Abbreviations ...............................................................................................................4

2.1 Introduction ANL-West, February 1951 to present .................................................................... 7

2.2 Experimental Breeder Reactor No. I, April 1951 to December 30, 1963...................................... 8

2.3 Boiling Water Reactor Experiment No. 1, Late 1953 to July 22, 1954 ....................................... 10

2.4 Boiling Water Reactor Experiment No. 2, October 19, 1954, to March 1955 ............................. 11

2.5 Boiling Water Reactor Experiment No. 3, June 9, 1955, to December 1956.............................. 11

2.6 Boiling Water Reactor Experiment No. 4, December 3, 1956, to June 1958.............................. 11

2.7 Boiling Water Reactor Experiment No. 5, February 9, 1962 to September 1964 ....................... 12

2.8 Zero Power Reactor No. 3, October 1955 to November 1970 ................................................... 12

2.9 Argonne Fast Source Reactor, October 29, 1959, to Late 1970................................................ 13

2.10 Transient Reactor Test Facility, February 23, 1958, to April 1994 ............................................. 13

2.11 Experimental Breeder Reactor No. II, May 1961 to September 30, 1994 .................................. 14

2.12 Hot Fuel Examination Facility, 1964 to Present......................................................................... 14

2.13 Zero Power Physics (Plutonium) Reactor, April 18, 1969 to April 1992 (standby)...................... 15
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2.14 Neutron Radiography Facility, October 1, 1977 to present ........................................................ 15

2.15 Fuel Assembly and Storage Building, 1970 to present .............................................................. 16

2.16 Other ANL-West Support Facilities............................................................................................ 16

References ........................................................................................................................................18

Glossary ............................................................................................................................................19

LIST OF TABLES

Table

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

Page

10

LIST OF FIGURES

Figure

2-1 Map of the Idaho National Engineering and Environmental Laboratory ....................................

Page

9
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 06/21/2004 00-A New technical basis document for the ANL-W 

Site Facilities and Processes. Initiated by 

Norman D. Rohrig.

Draft 08/17/2004 00-B Incorporates NIOSH review comments. Initiated 

by Norman D. Rohrig.

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

Rohrig.
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ACRONYMS AND ABBREVIATIONS

α alpha particle

AEC Atomic Energy Commission

AFSR Argonne Fast Source Reactor

AL Analytical Laboratory

ANL Argonne National Laboratory

ANL-E Argonne National Laboratory East

ANL-W Argonne National Laboratory West

ARA Army Reactor Area (later Auxiliary Reactor Area)

β beta particle

BORAX Boiling Water Reactor Experiment 

BORAX I Boiling Water Reactor Experiment No. 1

BORAX II Boiling Water Reactor Experiment No. 2

BORAX III Boiling Water Reactor Experiment No. 3

BORAX IV Boiling Water Reactor Experiment No. 4

BORAX V Boiling Water Reactor Experiment No. 5

BWR boiling-water reactor

C Celsius/centigrade

CAM continuous air monitor

CFA Central Facilities Area

c/m counts per minute

CTF Core Test Facility

DOE U.S. Department of Energy

DOE-ID DOE-Idaho Operations Office

DOELAP DOE Laboratory Accreditation Program

DU depleted uranium

EBR-I Experimental Breeder Reactor No. I

EBR-II Experimental Breeder Reactor No.II

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EFS Experimental Field Station

F Fahrenheit

FASB Fuel Assembly and Storage Building

FMF Fuel Manufacturing Facility

γ gamma

HFEF Hot Fuel Examination Facility

ICPP Idaho Chemical Processing Plant

IWP Industrial Waste Pond

IET Initial Engine Test

IFR Integral Fast Reactor

IMBA Integrated Modules for Bioassay Analysis

INL Idaho National Laboratory

INEL Idaho National Engineering Laboratory
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INEEL Idaho National Engineering and Environmental Laboratory

INTEC Idaho Nuclear Technology and Engineering Complex

IREP Interactive RadioEpidemiological Program

IWP Industrial Waste Pond

kW kilowatt

MAP mixed activation products

MFP mixed fission products

mR milliroentgen

mrem millirem

MW megawatt

NaK sodium potassium

Nf neutron flux-fast

NIOSH National Institute for Occupational Safety and Health

NRAD Neutron Radiography Facility

NRF Naval Reactors Facility

NRTS National Reactor Testing Station

OCAS (NIOSH) Office of Compensation Analysis and Support

ORAU Oak Ridge Associated Universities

PIC Pocket Ionization Chamber (i.e., "pencil" dosimeter)

psig pounds per square inch gauge

R Roentgen

rad radiation absorbed dose

RAM radiation/remote area monitor

rem Roentgen equivalent man

RESL Radiological Environmental Sciences Laboratory

RLWTF Radioactive Liquid Waste Treatment Facility

RWMC Radioactive Waste Management Complex

RSWF Radioactive Scrap and Waste Facility

SPERT Special-Power Excursion Reactor Test

SPF Sodium Processing Facility

STP Sewage Treatment Pond

t thermal

Τ1⁄2 Half-life of a nuclide or isotope

TAN Test Area North

TBD Technical Basis Document

TLD Thermoluminescent dosimeter

TRA Test Reactor Area

TREAT Transient Reactor Test

TRIGA Training Research and Isotope General Atomic

U.S.C. United States Code

WCA Waste Characterization Area
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WIPP Waste Isolation Pilot Plant

ZPR-III Zero Power Reactor No.3

ZPPR Zero Power Plutonium Reactor (later Zero Power Physics Reactor)
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2.1 INTRODUCTION ANL-WEST, FEBRUARY 1951 TO PRESENT

The purpose of this technical basis document (TBD) is to provide an Argonne National Laboratory -

West (ANL-W) Site Profile that contains information for use by the Oak Ridge Associated Universities 

(ORAU) Team, for any dose reconstruction action, to evaluate the total individual occupational dose 

for claimants under the Energy Employees Occupational Illness Compensation Program Act 

(EEOICPA) of 2000. 

TBDs and Site Profile Documents are general working documents that provide guidance concerning 

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 the National Institute for Occupational Safety and Health (NIOSH) 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 mean an “atomic weapons employer 

facility” or a “Department of Energy facility” as defined in the EEOICPA (42 U.S.C. 7384l (5) and (12)). 

The purpose of this TBD is to assist in the evaluation of worker dose from ANL-W processes using the 

methodology in External Dose Reconstruction Implementation Guideline (NIOSH 2002a) and Internal 

Dose Reconstruction implementation Guideline (NIOSH 2002b). This part describes facilities,

processes, and historical information related to worker internal and external exposures for use when 

actual monitoring data might be unavailable. 

This TBD provides supporting technical data to evaluate, with claimant-favorable assumptions, the 

total ANL-W occupational dose that can be reasonably associated with worker radiation exposure as 

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

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

methods used to prepare ANL-W 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 ANL-W exposure and dose values.

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

1949-1999, by Susan M. Stacy (Stacy 2000) was a resource in the development of facility and 

process descriptions.

The Idaho National Engineering and Environmental Laboratory (INEEL) or “the Site” is an 890-squaremile reservation encompassing almost 572,000 acres with a maximum distance of about 39 miles 

from north to south and 36 miles from east to west. It is 30 to 60 miles west of Idaho Falls, Idaho. 

The Site is on the Snake River Plain of southeastern Idaho at an elevation of about 5,000 ft. INEEL 

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

The Site has operated 52 reactors plus fuel handling and reprocessing and radioactive waste storage 

and disposal facilities since it began operations in 1949, about 10 of them by ANL-W. 

ANL-W was originally known as “the Idaho Division” of ANL. It is an extension of Argonne National 

Laboratory – East (ANL-E) near Chicago, Illinois. ANL is a U.S. Department of Energy (DOE)

research laboratory which has been operated by the University of Chicago throughout its life. The 

original site is in the southwest portion of INEEL, approximately 18 miles via Highway 20/26 east of 

Arco, 40 miles via Highway 26 northwest of Blackfoot, or 50 miles west via Highway 20 from Idaho 

Falls. The original ANL-W location is now a National Historic Landmark and is the site of the now 
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decommissioned Experimental Breeder Reactor – I (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 INEEL, about 35 miles 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. Figure 2-1 shows the relative location 

of each facility or technical area discussed in this TBD. 

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 (INEL) and, in 1997, the Idaho National 

Engineering and Environmental Laboratory. In 2004, the Site will become the Idaho National 

Laboratory (INL). 

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

Operations Office (DOE-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. Originally, DOE-ID personnel at 

the Radiological 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. 

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 [PICs], 

etc.) as well as respiratory protection, anticontamination clothing, etc. 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. If airborne radioactivity might be present or if internal exposure was possible, 

applicable respirators were provided to prevent/reduce internal exposure. 

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. 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 within Site locations. 

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

lists common radionuclides for reactors.

2.2 EXPERIMENTAL BREEDER REACTOR NO. I, APRIL 1951 TO DECEMBER 30, 1963

EBR-I, the first reactor built at INEEL, 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 

238U around the core provided the fertile material in which nuclear material breeding took place. 
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Figure 2-1. Map of the Idaho National Engineering and Environmental Laboratory showing location of

ANL-W, including EBR-II, Transient Reactor Test (TREAT) Facility, and Zero Power Plutonium 

Reactor (ZPPR). ANL-W operated EBR-I and Boiling Water Reactor Experiment (BORAX) areas 

before their shutdowns. Other INEEL facilities include the Auxiliary Reactor Area (ARA); CFA; Idaho 

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

(INTEC)]; Radioactive Waste Management Complex (RWMC); Special Power Excursion Reactor Test 

(SPERT) area; Grid III, Test Reactor Area (TRA); the Experimental Field Station (EFS); Naval 

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

the Core Test Facility (CTF) at TAN.
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Table 2-1. 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

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 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. 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 hr. The saturation level at full-power operation was approximately 24 μCi/gm. The second 

most significant activity was about 2 μCi/gm 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

Personnel received 

potential existed from airborne radioactivity from mixed fission products and 

activation products.

external exposure

2.3 BOILING WATER REACTOR EXPERIMENT NO. 1, LATE 1953 TO JULY 22, 1954

from mixed fission products and activation products during 

activities associated with reactor operation and maintenance.

BORAX-I was an open-top, water-cooled, water-moderated, boiling-water reactor (BWR) used to 

conduct a series of nondestructive experiments in the latter part of 1953 and early summer of 1954. 

The reactor was built in an excavated area of earth approximately one-half mile northwest of EBR-I 

and was housed in a 10-foot-diameter tank open to the atmosphere. The control room was 

approximately one-half mile away near the EBR-I reactor. 

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 feet by 350 feet 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 seconds to 25 mrem/hr and within 5 minutes 

to less than 1 mrem/hr. A detailed discussion of this incident is available (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.
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External exposure

2.4 BOILING WATER REACTOR EXPERIMENT NO. 2, OCTOBER 19, 1954, TO MARCH 

1955

occurred from direct radiation associated with the reactor operation and

maintenance activities from mixed fission products and activation products. BWRs typically have a 

N-16 high energy gamma shine from the primary water loop.

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. 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 varying enrichments of 235U in metal fuel 

plates. The boiling-water system operated 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. In 1955, tests of a turbine generator added 

to BORAX-II demonstrated that turbine contamination would not be a significant problem in BWRs. 

Internal exposure was possible from mixed activation products (MAP) associated with work activities 

from the reactor coolant. 

External exposure

2.5 BOILING WATER REACTOR EXPERIMENT NO. 3, JUNE 9, 1955, TO DECEMBER 

1956

occurred during routine operations and with loading and unloading the reactor. 

As a result of the BORAX-I and -II tests, a program began in March of 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 hours 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. 

Internal exposure was possible from MAP associated with work activities from the reactor coolant.

External exposure

2.6 BOILING WATER REACTOR EXPERIMENT NO. 4, DECEMBER 3, 1956, TO JUNE 

1958

occurred from routine operation and loading and unloading of the reactor.

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. 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 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. BORAX-IV 
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released approximately 4,565 curies 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:

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 MAP/mixed fission products (MFP) associated 

with the reactor coolant.

External exposure

2.7 BOILING WATER REACTOR EXPERIMENT NO. 5, FEBRUARY 9, 1962 TO 

SEPTEMBER 1964

occurred during operation and work with loading and unloading the reactor. 

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. 

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 associated 

maintenance work.

External exposure

2.8 ZERO POWER REACTOR NO. 3, OCTOBER 1955 TO NOVEMBER 1970

occurred from routine activities associated with reactor operation and maintenance. 

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.

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 x 42 ft x 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-curie 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.
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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 235U/ft3

.

Internal exposure potential existed from possible airborne radioactivity during maintenance operations 

from MFP/MAP.

External exposure

2.9 ARGONNE FAST SOURCE REACTOR, OCTOBER 29, 1959, TO LATE 1970

occurred maintenance activities and during loading and unloading of the reactor 

fuel. 

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 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 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.

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

radioactivity. 

External exposure 

2.10 TRANSIENT REACTOR TEST FACILITY, FEBRUARY 23, 1958, TO APRIL 1994

occurred from maintenance activities and loading and unloading fuel.

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 occurred on February 23, 1959. The 

TREAT complex comprises a reactor building and a control building approximately 1 mile northwest of 

the EBR-II containment building (Freund et al. 1960).

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 atop the reactor during steady-state operations at 

100 kW. Access to the subreactor room was controlled during steady-state operation. Prior to 

transient operations, the building was evacuated of all personnel. General neutron and gamma 

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

al. 1960):

• Fast neutron Negligible

• Thermal neutrons 50-1,500 n/cm2

/sec

• 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.

External exposure

2.11 EXPERIMENTAL BREEDER REACTOR NO. II, MAY 1961 TO SEPTEMBER 30, 1994

occurred from routine operations. 

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 electrical 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. 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 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.

The reactor was submerged in a primary tank containing approximately 90,000 gallons 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 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 (SPF) 

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.

2.12 HOT FUEL EXAMINATION FACILITY, 1964 TO PRESENT

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. 

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 post-irradiation 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 DOE's 

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 (WCA) of the HFEF high bay.

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

materials could cause airborne radioactivity from mixed fission products and activation products. 

External exposure

2.13 ZERO POWER PHYSICS (PLUTONIUM) REACTOR, APRIL 18, 1969 TO APRIL 1992 

(STANDBY)

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

during equipment maintenance and refurbishment.

ZPPR is a split-table critical facility approximately 300 meters from EBR-II in the ANL-W area and 

about 3 miles 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-foot-diameter 

circular room with floor and walls of reinforced concrete. The roof is composed of layers of washed 

and dried sand and gravel supported by a catenary cable network.

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. 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. 

External exposure

2.14 NEUTRON RADIOGRAPHY FACILITY, OCTOBER 1, 1977 TO PRESENT

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

The Neutron Radiography (NRAD) Facility is a 250-kW, steady-state training research and isotope 

(built by 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 water-filled 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 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 
Effective Date: 09/09/2004 Revision No. 00 Procedure No. ORAUT-TKBS-0026-2 Page 16 of 24

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 mixed fission products and activation products.

External exposure

2.15 FUEL ASSEMBLY AND STORAGE BUILDING, 1970 TO PRESENT

occurred during sample handling and maintenance associated with radioactive 

samples. Remote handling techniques are used to minimize dose.

The Fuel Assembly and Storage Building (FASB) is a multipurpose facility about 100 yards NE 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 for preparing metallurgical samples, and 

inert atmosphere glove boxes and hoods. The facility ceased fuel assembly about in 1990 , but other 

radiological work is ongoing.

Internal exposure might occur from airborne radioactivity associated with the described processes 

from the uranium fuel and spent fuel examinations. 

External exposure

2.16 OTHER ANL-WEST SUPPORT FACILITIES 

occurred from movement of radiological samples and reactor fuel. 

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

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

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

subjected to radiation levels in the multi R/hr range on occasion in the Junior Cave Area. 

The Radioactive Scrap and Waste Facility (RSWF) 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 (RLWTF) 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 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.

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

These ponds are not suspected to be radiologically contaminated.
Effective Date: 09/09/2004 Revision No. 00 Procedure No. ORAUT-TKBS-0026-2 Page 17 of 24

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

worked in these facilities.

External exposure occurred from working with the variety of radioactive material associated with these 

facilities (mixed fission products, activation products and transuranics).
Effective Date: 09/09/2004 Revision No. 00 Procedure No. ORAUT-TKBS-0026-2 Page 18 of 24

REFERENCES

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

Conference on Remote Systems Technology, San Francisco, California.

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, October Idaho Falls, Idaho.

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

Radiation Emergencies, Including Lecture Notes from a Training Session, February 12-14, 

1958, ID-8206, INEEL, September, Idaho Falls, Idaho.

Brunson, G. S., 1959, Design and Hazards Report for the Argonne Fast Source Reactor, AFSR, ANL6024, Idaho Falls, Idaho June.

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

Report on the Transient Test Reactor Facility TREAT, ANL 6034, Idaho Falls, Idaho, June.

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

Report EBR-1, Mark I, ANL 6411, Idaho Falls, Idaho. February.

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

Changes in the Core, ANL-5731, Idaho Falls, Idaho, November.

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

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

Reactor II (EBR-II), ANL 5719, Idaho Falls, Idaho, May.

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, Idaho Falls, Idaho, February.

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

Implementation Guidelines, Revision 1, OCAS-IG-001, Office of Compensation Analysis and 

Support, Cincinnati, Ohio.

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

Implementation Guidelines, Revision 0, OCAS-IG-002, Cincinnati, Ohio.

Novick, M., (Director, Idaho Division, Argonne Nation Laboratory), 1958, Letter to J.R. Horan (Idaho 

Operations Office), Idaho Falls, Idaho, 1 April.

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

Facility,” Proceedings of 27th Conference on Remote Systems, San Francisco, California, 

November.

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

Environmental Laboratory 1949 - 1999, U.S. Department of Energy, Idaho Operation Office, 

Idaho Falls, Idaho.
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GLOSSARY

activation

The process of inducing radioactivity by irradiation.

annual dose equivalent

The dose equivalent received in a year. The annual dose equivalent is expressed in units of 

rem (sievert).

Atomic Energy Commission

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

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

background radiation

The radiation in an ambient environment that includes cosmic rays, radiation from natural 

sources, and man-made sources.

beta (β) dose

A designation for external dose referring to the dose from beta and less-energetic gamma and

X ray radiation; typically a shallow dose or dose to the lens of the eye. 

beta radiation

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

radioactive elements. Most direct fission products emit beta radiation. 

boiling-water reactor

A nuclear reactor concept in which the coolant water is permitted to boil as it absorbs the heat 

of the nuclear reaction. 

breeder reactor

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

fuel. 

containment building

A safety feature of reactors typically engineered to be an airtight building, to prevent the 

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

the containment.

control rod

A device manipulated within a nuclear reactor constructed of material to absorb neutrons for 

the purpose of slowing down or increasing the nuclear reaction. 

contamination, radioactive

Particulate matter on surfaces or in the air that is radioactive. 

control room

The operating center of a nuclear reactor from which the reactor is operated and monitored.

coolant

In a reactor, a gas or fluid (water, liquid metal, etc.) contacting reactor fuel for the purpose of 

reducing or removing heat generated by the nuclear reaction.
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core

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

operation.

curie

A special unit of radioactivity equal to 3.7x1010 disintegrations per second (dps).

decontaminate

A process removing radioactive particles from a person, place, or object.

depleted uranium

Uranium material that has undergone a process to remove 235U resulting primarily in 238U.

dose equivalent (H)

The product of the absorbed dose (D), the quality factor (Q), and any other modifying factors. 

The dose unit is the rem.

dose

Amount of energy from ionizing radiation absorbed per unit of mass.

dosimeter

A device used to measure accumulated radiation exposure. 

dosimetry

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

from external or internal sources of radiation. 

enriched uranium

Uranium enhanced from its natural state to contain a higher abundance of the isotope 235U. 

excursion

Planned or accidental increase in the normal operating power level of nuclear reactions.

exposure

Technically, a measure of X-ray or gamma radiation capability to ionize air (units of Roentgen). 

film

Generally means a "film packet" that contains one or more pieces of film in a light-tight 

wrapping. The developed film has an image caused by radiation that can be measured using 

an optical densitometer.

fission

A nuclear transformation characterized by the splitting of a nucleus into at least two other 

nuclei and the release of a relatively large amount of energy.

fission product

Elements or compounds resulting from fission.

flux

The intensity of particles (e.g., the number of neutrons passing through a unit area in unit 

time.)
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fuel assembly

An arrangement of nuclear fuel and its cladding material into a particular form and shape for 

use in a nuclear reactor. Fuel may be assembled in plates, rods of various diameters, or other 

shapes.

fuel reprocessing

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

products from spent fuel.

gamma rays

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

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

in an energy range of 10 keV to 9 MeV.

half-life

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

hot cell

A 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. 

ionization chamber

A device used to measure exposure or radiation dose. 

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

irradiate

To expose a substance to radiation. 

isotope

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

a differing number of neutrons (different mass number).

megawatt

A measure of electrical power equal to 1 million watts. 

millirem

A unit of radiation equal to one-thousandth of a rem (see rem).

microcurie

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

mixed waste

Waste that is both chemically hazardous and radioactive. 

moderator

A material used in a nuclear reactor to reduce the natural speed of neutrons ejected from 

fissioning atoms (water, graphite etc.).
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natural uranium

Uranium occurring in nature that has not been through an enrichment process.

neutron

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

hydrogen atom.

neutron, fast

Neutrons with energy equal or greater than 10 keV.

neutron, thermal

Strictly, neutrons in thermal equilibrium with surroundings. Generally, neutrons with energy 

less than about 0.5 eV.

nuclear energy

Energy released from a nuclear fission or fusion reaction. 

nuclear power plant

An electrical generating facility using nuclear fuel 

nuclear waste

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

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

nucleus

That part of an atom consisting of the total positive electrical charge and most of the mass. 

pocket dosimeters

A type of ionization chamber used by personnel to measure radiation dose. Other names are 

pencil dosimeter, pocket pencil, pocket ionization chamber (PIC).

photon

A quantum of electromagnetic energy often referred to as X-rays or gamma rays, but also 

including light and radiant heat. 

pressurized-water reactor

A concept in which water used to cool the reactor core is pressurized to prevent boiling. Heat 

is typically transferred from a primary system to a secondary system.

primary loop

A 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. 

proton

An elementary atomic particle with a positive electrical charge equal numerically to the charge 

of the electron and a mass slightly greater than 1 mass unit.

Quality factor, Q

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

The unit of absorbed dose.

radiation

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

ionizing radiation).

radioactivity

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

neutrons from unstable nuclei.

radioactive waste

Byproducts of nuclear processes that are radioactive and have no useful recyclable purpose 

(see nuclear waste).

radionuclide

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

the number of protons, neutrons, atomic number, and mass number. 

reactor vessel

A container enclosing the fuel elements, control elements, coolant piping, and other structures 

that support the core of a nuclear reactor.

reflector

Part of the structure of some nuclear reactors designed to reflect neutrons back into the core 

of the reactor.

rem

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

retention basin

An outdoor basin (of any of several designs) in which liquid solutions are deposited and held 

pending evaporation or the precipitation of solids. 

Roentgen (R)

A unit of exposure to gamma (or X-ray) radiation, defined precisely as the quantity of gamma 

(or X) rays that will produce a total charge of 2.58 x 10-4 coulomb in 1 kg of dry air STP. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher (~>100 keV) energy photons.

secondary loop

The system in a reactor receiving transfer heat from a primary system. The secondary system 

is not exposed to the reactor core and is free of radioactivity under normal circumstances. 

shielding

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

materials from radiation.

Sievert (Sv)

The SI unit for dose equivalent. (1 Sv = 100 rem.)
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spent fuel

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

a chain reaction.

spent fuel storage basin

A pool or pit made of reinforced concrete containing water and used to store spent nuclear 

fuel. The water acts as shielding and as a coolant 

thermoluminescence

Property of a material resulting in light emission of stored energy released by heat.

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. 

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1000 mg/cm2

); used to 

refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin with energies less than 250 keV or 

a radiograph.

zero power

Also called low power; operating a reactor to maintain a chain reaction at an extremely low 

power level producing very little heat. Zero power reactors are used as sensitive laboratory 

tools to pretest experimental loadings of test reactors and for other analytical purposes.