Draft

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute for Occupational Safety and Health 

INL SEC-00219 REACTOR PRIORITIZATION FOR 

EVALUATION OF ORAUT-OTIB-0054 APPLICABILITY

Contract No. 211-2014-58081 

SCA-TR-2016-SEC002, Revision 0 

Prepared by 

Stephen L. Ostrow, PhD

SC&A, Inc.

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755 

March 2016 

DISCLAIMER

This is a working document provided by the Centers for Disease Control and Prevention (CDC) technical 

support contractor, SC&A for use in discussions with the National Institute for Occupational Safety and 

Health (NIOSH) and the Advisory Board on Radiation and Worker Health (ABRWH), including its 

Working Groups or Subcommittees. Documents produced by SC&A, such as memorandum, white paper, 

draft or working documents are not final NIOSH or ABRWH products or positions, unless specifically 

marked as such. This document prepared by SC&A represents its preliminary evaluation on technical 

issues. 

NOTICE: This report has been reviewed to identify and redact any information that is protected by the 

Privacy Act 5 U.S.C. § 552a and has been cleared for distribution.
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SC&A, INC.: Technical Support for the Advisory Board on Radiation & Worker 

Health Review of NIOSH Dose Reconstruction Program

DOCUMENT TITLE: 

INL SEC-00219 Reactor Prioritization for Evaluation of ORAUTOTIB-0054 Applicability 

DOCUMENT NUMBER/

DESCRIPTION: 

SCA-TR-2016-SEC002 

REVISION NO.: 0 (Draft)

SUPERSEDES: N/A

EFFECTIVE DATE: March 2, 2016 

TASK MANAGER: John Stiver, MS, CHP 

PROJECT MANAGER: John Stiver, MS, CHP 

DOCUMENT 

REVIEWER(S): John Stiver 

Record of Revisions

Revision Number Effective Date Description of Revision

0 (Draft) 03/02/2016 Initial issue
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TABLE OF CONTENTS

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

1.0 Introduction ............................................................................................................................... 7

2.0 ORAUT-OTIB-0054 ................................................................................................................. 9

3.0 Evaluation ............................................................................................................................... 11

4.0 References ............................................................................................................................... 13

Attachment 1. INL Reactor Prioritization with Respect to ORAUT-OTIB-0054 Applicability 

Investigation .................................................................................................................................. 14
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ABBREVIATIONS AND ACRONYMS

Advisory Board Advisory Board on Radiation and Worker Health

AFSR Argonne Fast Source Reactor 

ALPR Argonne Low Power Reactor

ANP Aircraft Nuclear Propulsion

ANL-W Argonne National Laboratory-West

ARA Auxiliary Reactor Area

ARMF Advanced Reactivity Measurement Facility Number

ATR Advanced Test Reactor

ATRC Advanced Test Reactor Critical Facility

BORAX Boiling Water Reactor Experiment

CET Critical Experiment Tank

CFA Central Facilities Area

CFRMF Coupled Fast Reactivity Measurement Facility

CRCE Cavity Reactor Critical Experiment

Cs cesium

D2O deuterium oxide

DOE (U.S.) Department of Energy

EBOR Experimental Beryllium Oxide Reactor

EBR Experimental Breeder Reactor

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EOCR Experimental Organic Cooled Reactor

ESF engineered safety feature

ETR Engineering Test Reactor

ETRC Engineering Test Reactor Critical Facility

F Fahrenheit

FFTF Fast Flux Test Facility

FRAN Nuclear Effects Reactor 

Ft feet, foot

GCRE Gas Cooled Reactor Experiment
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GW gigawatt

GWth gigawatt thermal

HOTCE Hot Critical Experiment

HTGR high-temperature gas-cooled reactor

HTRE Heat Transfer Reactor Experiment

IET Initial Engine Test

IETF Initial Engine Test Facility

in inch

INL Idaho National Laboratory

kWe kilowatts electric

kWth kilowatts thermal

LOCA loss-of-coolant accident

LOFT Loss of Fluid Test Facility

LPTF Low Power Test Facility

LWR light water reactor

ML-1 Mobile Low-Power Reactor-1

MTHM metric ton heavy metal

MTR Materials Test Reactor

MTU metric ton uranium

MW megawatt

MWd megawatt days

MWe megawatt electric

MWth megawatt thermal

MW-hr megawatt-hour

N/A not applicable

Na sodium

NaK sodium-potassium (liquid metal)

NASA National Aeronautics and Space Administration

NRAD Neutron Radiography Facility

NRC (U.S.) Nuclear Regulatory Commission

NRF Naval Research Facility
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NIOSH National Institute for Occupational Safety and Health

OMRE Organic Moderated Reactor Experiment

ORAU(T) Oak Ridge Associated Universities (Team)

ORNL Oak Ridge National Laboratory

OTIB ORAUT Technical Information Bulletin

PBF Power Burst Facility

Pu plutonium

RMF Reactivity Measurement Facility

RML Radiation Measurement Laboratory

SCRCE Spherical Cavity Reactor Critical Experiment

SEC Special Exposure Cohort

SL-1 Stationary Low-Power Reactor

SNAP Systems for Nuclear Auxiliary Power

SNAPTRAN Systems for Nuclear Auxiliary Power Transient

SPERT Special Power Excursion Reactor Test

Sr strontium

STPF Shield Test Pool Facility

STR Submarine Thermal Reactor

SUSIE Shield Pool Test Facility Reactor

TAN Test Area North

THRITS Thermal Reactor Idaho Test Station

TRA Test Reactor Area

TREAT Transient Reactor Test Facility

TRIGA Training, Research, Isotope General Atomics (reactor)

U uranium

UF6 uranium hexafluoride

UO2 uranium dioxide

USC United States Code

W watt

WRRTF Water Reactor Research Test Facility

ZPPR Zero Power Physics Reactor
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1.0 INTRODUCTION

Following a series of meetings and discussions among the Advisory Board on Radiation and 

Worker Health (hereafter referred to as the “Advisory Board”), SC&A, and the National Institute 

for Occupational Safety and Health (NIOSH) and its technical contractor, Oak Ridge Associated 

Universities (ORAU), about NIOSH’s March 12, 2015, release of the Special Exposure Cohort 

(SEC) Petition SEC-00219 Evaluation Report for the Idaho National Laboratory (INL) (NIOSH 

2015a),1 the Advisory Board requested that SC&A initially review two issues as part of a graded 

approach to assess the complex issues at this complex site: (1) class definition and (2) dose 

reconstructability and gap analysis. With respect to the latter, inherent in the SEC framework is 

the assumption that doses can be reconstructed with sufficient accuracy for site areas and time 

periods that lie outside the SEC class definition and that are not being held in reserve for further 

evaluation by NIOSH. Operations at INL involving radioactive materials were very complex, as 

many unique nuclear reactors and experiments were built and tested, irradiated nuclear fuel 

handled and processed, and radioactive waste disposed of. An SC&A report (SC&A 2015a) 

examined one aspect of the dose reconstructability assumption for one of the several major site 

areas, the Test Reactor Area (TRA). Another SC&A report (SC&A 2015b) examined dose 

reconstructability for some of the reactors in another major site area, Test Area North (TAN).

1 Revision 1 of the ER was released on July 21, 2015 (NIOSH 2015b).

A primary tool that NIOSH uses for internal dose reconstruction is the guidance appearing in 

ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal Dose-Related 

Gross Beta and Gross Gamma Analyses (hereafter referred to as “OTIB-0054”) (ORAUT 2015). 

Except for certain situations, OTIB-0054 assigns fission and activation product intakes for 

different radioisotopes that are directly tied to an indicator radionuclide [strontium-90 (Sr-90) or 

cesium-137 (Cs-137)]. OTIB-0054 generated nine different representative reactor cases, which 

are intended to envelope reactor and nuclear fuel types and operating scenarios to which workers 

might have been exposed. SC&A (2015a) evaluated whether OTIB-0054 is applicable to the 

three large materials-testing reactors located in the TRA: the Materials Test Reactor (MTR), the 

Engineering Test Reactor (ETR), and the Advanced Test Reactor (ATR). SC&A (2015b) 

similarly evaluated whether OTIB-0054 is applicable to the three Heat Transfer Reactor 

Experiment (HTRE) reactors located in TAN that supported the Aircraft Nuclear Propulsion 

(ANP) program. Subsequently, at the November 10, 2015, INL Work Group meeting, the 

Advisory Board members directed SC&A to screen reactors other than the six already addressed 

and create a prioritized list of reactors for detailed examination at a later date with respect to 

OTIB-0054 applicability; this report is the result of SC&A’s screening. 

Susan Stacy, in her comprehensive review of the history of INL from inception through 1999, 

Proving the Principle (Stacy 2000), lists in Appendix B the 52 reactors that were built at INL 

(including two that never operated) and provides a brief summary of each. SC&A, following the 

practice of NIOSH in its INL reports (e.g., the site profile), uses Stacy’s list as a convenient 

 
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framework to examine the various reactors. Attachment 1 lists all the reactors and, for each, 

notes its operating period, provides a brief summary description, and presents SC&A’s screening 

assessment of its priority ranking with respect to performing a detailed review on whether OTIB0054 bounds its makeup and operating conditions. 

The priority rankings are divided into three categories: High, Medium, and Low.2

2 There is also a category for those reactors that are not considered in the prioritization process. 

 Though based 

on a substantial amount of research, the rankings are still somewhat subjective because a full 

analysis would involve actually performing the detailed OTIB-0054 applicability analyses 

themselves. The assignment of reactors to priority ranking categories considers reactor design 

factors, such as the types of fuel, enrichment, cladding, moderator, and coolant; and operational 

modes, such as steady-state or periodic operations, as common for experimental reactors; and 

whether the reactor performed within design limits or was deliberately or inadvertently taken 

outside those limits, such as in tests supporting power reactor safety programs. 

It should be noted that the initial list of 52 reactors was quickly reduced by subtracting the three 

TRA and three TAN reactors already evaluated in SC&A 2015a and SC&A 2015b, respectively, 

the four reactors at the Naval Reactor Facility (NRF) because that area is outside the Energy 

Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA) program, the 

12 reactors at Argonne National Laboratory-West (ANL-W) because that area is not included in 

the INL SEC-00219, and the two reactors that were canceled before operations. Subtracting these 

24 reactors leaves 28 candidate reactors for further study. 

 
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2.0 ORAUT-OTIB-0054

After a series of initial runs using the ORIGEN2 isotope generation and depletion code (Croff 

1980), OTIB-0054 selects four actual reactors to represent different general categories of 

reactors that might envelope the wide variety of reactors at the different sites considered in the 

EEOICPA program.

3

3

 ORAUT 2015 should be consulted for the details of the selection process. 

The representative reactors are listed in Table 1. 

Table 1. ORAUT-OTIB-0054 Representative Reactors

Category Reactor

High-flux reactors Advanced Test Reactor (ATR)

Na-cooled fast reactors Fast Flux Test Facility (FFTF)

Pu production reactors Hanford N-Reactor

Research reactors TRIGA with stainless steel cladding

Source: ORAUT 2015

Multiple ORIGEN-S (ORNL 2015) runs produced a total of nine representative cases for the four 

reactors. ORIGEN-S is a more modern and capable version of ORIGEN and is part of the 

SCALE code system (ORNL 2015) for nuclear safety analysis and design, developed and 

maintained by ORNL for the U.S. Nuclear Regulatory Commission (NRC). Table 2 (OTIB Table 

5-2) lists the parameters and basis selected for each of the cases.

Table 2. ORIGEN-S Irradiation Parameters for the Nine Representative Reactor Cases

Case Parameters Basis

ATR 1 

Specific power = 2,379.1 MW/MTU

Irradiation time = 132.27 days

Burnup = 314,684 MWd/MTU

Maximum burnup at nominal power. 

ATR 2 

Specific power = 8,651.2 MW/MTU

Irradiation time = 36.4 days

Burnup = 314,904 MWd/MTU

Maximum burnup at maximum assembly

power. 

ATR 3 

Specific power = 2,379.1 MW/MTU

Irradiation time = 56 days 

Burnup = 133,230 MWd/MTU

Nominal burnup at nominal power. 

FFTF 1 

Specific power = 163.8 MW/MTHM

Irradiation time = 929.4 days 

Burnup = 152,230 MWd/MTHM

Maximum burnup at nominal power. 

 
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Case Parameters Basis

FFTF 2 

 

Specific power = 163.8 MW/MTHM

Irradiation time = 488.3 days 

Burnup = 79,984 MWd/MTHM

Nominal burnup at nominal power. 

N Reactor 1 

Specific power = 10.4 MW/MTU

Irradiation time = 114.2 days 

Burnup = 1,188 MWd/MTU

Production of weapons-grade plutonium

(nominal 6% Pu-240 content) at nominal 

power. 

N Reactor 2 

Specific power = 10.4 MW/MTU

Irradiation time = 285.6 days 

Burnup = 2,970 MWd/MTU

Production of fuel-grade plutonium (nominal 

12% Pu-240 content) at nominal power. 

TRIGA 1 

Specific power = 15.57 MWd/MTU

Irradiation time = 730.1 days 

Burnup = 11,368 MWd/MTU

Maximum burnup at nominal power. 

TRIGA 2 

Specific power = 15.57 MW/MTU

Irradiation time = 115.2 days 

Burnup = 1994 MWd/MTU

Nominal burnup at nominal power. 

Source: Reproduced from ORAUT 2015, Table 5-2. 
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3.0 EVALUATION

NIOSH uses OTIB-0054 to determine internal doses to claimants using indicator radionuclides in 

cases where only gross beta or gross gamma measurements are available. The nine cases of 

Table 2 are intended to envelope reactor and nuclear fuel types and operating scenarios to which 

workers might have been exposed. The evaluation in this report examines whether OTIB-0054 

adequately envelopes the INL reactors listed in Attachment 1. The first step in the process was to 

identify which of the 52 listed reactors should be excluded from the evaluation: i.e., reactors that 

have already been examined elsewhere, that were in NRF or ANL-W, or that never operated. 

Table 3 shows the results of that winnowing process. 

Table 3. INL Reactors Excluded From the Prioritization Process

Reactor Category Reactor Number from Attachment 1

ANL-W (12) 5, 6, 7, 8, 9, 10, 17, 18, 33, 50, 51, 52

NRF (4) 27, 28, 32, 45

Already Evaluated (6) 3, 14, 22, 23, 24, 30

Never Operated (2) 16, 19

Total Excluded Reactors 24

Attachment 1 categorizes the remaining 28 reactors according to three prioritization levels: High, 

Medium, or Low, depending on a number of factors, such as the fuel properties, power level, and 

operating characteristics and history. The 24 reactors that are excluded from the prioritization 

process are included in the last group in Attachment 1 for completeness. It is apparent from 

looking at the summary descriptions of Attachment 1 that most of the INL reactors were 

different from the four representative reactors of OTIB-0054, since most of the former were oneof-a-kind experiments that might have utilized different fuels, moderators, and coolants, and 

were often deliberately or inadvertently operated beyond design limits, sometimes to failure. In 

addition, rather than operate at more-or-less steady-state conditions for some length of time, 

some INL reactors operated in pulsed mode, in which they produced a huge amount of power, 

but in a very short time interval, before they shut themselves down (due to strongly negative 

reactivity properties, such as negative void or temperature coefficients). Fuel burnups were 

frequently considerably lower than for the representative reactors of OTIB-0054, so that the 

long-lived decay products did not have the chance to build up in the fuel, resulting in different 

isotopic ratios than in the OTIB-0054 reactors. 

Notwithstanding the above considerations, SC&A categorized the 28 reactors as a guide to which 

should be considered first in a more detailed study, such as was done in SC&A 2015a for some 

reactors in the TRA and in SC&A 2015b for some in TAN. The results of the categorizations of 

Attachment 1 are summarized in Table 4, which shows that 13 reactors were put in the High 

category, 8 in the Medium category, and 7 in the Low category.
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Table 4. Priority Class Categorization 

Priority Class Reactor Number from Attachment 1

High (13) 11, 29, 31, 35,36, 39, 40, 41, 42, 43, 46, 47, 48

Medium (8) 2, 12, 13, 20, 26, 34,3 8, 49

Low (7) 1, 4, 15, 21, 25, 37, 44

Total Included Reactors 28
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4.0 REFERENCES

Croff 1980. Croff, A. G., ORIGEN2 – A Revised and Updated Version of the Oak Ridge Isotope 

Generation and Depletion Code, ORNL-5621, Oak Ridge National Laboratory. July 1980. 

NIOSH 2015a. SEC Petition Evaluation Report: Petition SEC-00219, Revision 0, National 

Institute for Occupational Safety and Health. March 12, 2015. 

NIOSH 2015b. SEC Petition Evaluation Report: Petition SEC-00219, Revision 1, National 

Institute for Occupational Safety and Health. July 21, 2015. 

ORAUT 2015. Fission and Activation Product Assignment for Internal Dose-Related Gross Beta 

and Gross Gamma Analyses, ORAUT-OTIB-0054, Revision 3, Oak Ridge Associated 

Universities. February 6, 2015. 

ORNL 2015. SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety 

Analysis and Design, Oak Ridge National Laboratory. Accessed February 18, 2016, at 

http://scale.ornl.gov. 

SC&A 2015a. NIOSH SEC-00219: Test Reactor Area Modeling, SCA-SEC-2015-0074-C, 

Revision 0, SC&A, Inc. September 28, 2015. 

SC&A 2015b. Review of NIOSH Strategy for Reconstructing Internal Doses to Workers at Test 

Area North, SCA-TR-2015-SEC0074A, Revision 0, SC&A, Inc. September 28, 2015. 

Stacy 2000. Stacy, Susan M., Proving the Principle: A History of the Idaho National 

Engineering and Environmental Laboratory 1949-1999, DOE/ID-10799, Idaho Operations 

Office, U.S. Department of Energy, Idaho Falls, ID. 2000. 
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ATTACHMENT 1. INL REACTOR PRIORITIZATION WITH RESPECT TO ORAUT-OTIB-0054 

APPLICABILITY INVESTIGATION 

Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking 

HIGH PRIORITY RANKING

11. Cavity Reactor Critical 1967–early 1970s CRCE was located in a shielded test cell of the Low The CRCE, with a gaseous core of UF6, lies 

Experiment (CRCE) Power Test Facility (LPTF) in the Water Reactor totally outside the envelope of any of the 

TAN - WRRTF – LPTF Research Test Facility (WRRTF) area of Test Area 

North (TAN). It was a National Aeronautics and 

Space Administration (NASA)-sponsored 

OTIB-0054 representative reactors. 

Priority Ranking: High

experiment to investigate using nuclear power for 

space rocket propulsion, where a gaseous core would 

be suspended in a spherical tank by a fast-moving 

hydrogen propellant. The hydrogen could be heated 

to a very high temperature, producing a very high 

specific impulse that might be used, for example, on 

a voyage to Mars. The core in the CRCE was 

uranium hexafluoride (UF6), operating at the 

relatively low temperature of about 200° F. 

29. Loss of Fluid Test 

Facility (LOFT) 

TAN (Bldg. TAN-650)

1973–1985 The LOFT series of 38 nuclear power experiments, 

sponsored by the U.S. Nuclear Regulatory 

Commission (NRC), made major contributions to the 

light water reactor (LWR) safety program for 

commercial nuclear power plants by simulating 

system behaviors during a loss-of-coolant accident

(LOCA). The reactor, which had a maximum power 

of 50 MWth, and the associated components and 

systems, were built as a scale model of a 

commercial, four-coolant-loop pressurized water 

reactor and its engineered safety features (ESFs).

The LOFT facilities modeled a commercial 

LWR but often operated beyond usual limits to 

explore coupled nuclear and thermal hydraulic 

behavior in design-basis or beyond-design-basis 

accident scenarios. 

Priority Ranking: High
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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking 

The LOFT facility included five major systems: the 

reactor system, primary coolant system, blowdown 

suppression system, emergency core cooling system, 

and a secondary coolant system. The reactor fuel had 

the same characteristics as commercial fuel. The 

tests investigated whether the ESFs activated in a 

LOCA, sometimes coupled with loss of offsite 

power, could prevent or mitigate core damage and 

release of radioactive material. The LOFT reactor 

was located within a containment building to 

minimize radioactive releases to the environment. 

The experiments simulated different LOCAs due to 

small, medium, or large pipe breaks, including the 

actual Three Mile Island meltdown scenario of 1979. 

The resulting nuclear and thermal hydraulics data 

gave insight into system behavior during design and 

severe accidents, which could be used both to 

improve reactor system designs and to improve 

accident systems modeling codes. 

The last test, in 1985, which involved the deliberate 

melting of 100 fuel rods in an experimental fuel 

bundle, provided data on system performance in the 

event of a severe accident and measured radioactive 

releases from the damaged fuel. The test also 

resulted in the release of radioactivity into the 

containment building. 
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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

31. Mobile Low-Power 

Reactor No. 1 (ML-1)

Auxiliary Reactor Area IV 

(ARA-IV)

1961–1964 The Gas Cooled Reactor Experiment (GCRE) served 

as the prototype for the U.S. Army’s hightemperature, gas-cooled reactor (HTGR), water 

moderated ML-1, which was located in the ARA-IV 

area. The ML-1 was truly mobile in that it could be 

packed up after a 36-hour shutdown, its three skids 

put on trailers, and moved to a new location. The 

plant produced 3.4 MWth and, coupled to a compact 

power generation section, 330 kWe. 

The ML-1, built by Aerojet General, went critical on 

March 31, 1961, and operated until 1964; the Army 

subsequently ended the development program in 

1965. The fuel was 93% enriched contained in 61 

fuel elements with 19 pins per element, in an 

approximately 22" D x 22" H cylindrical core, 

moderated by light water and cooled by nitrogen. 

None of the representative reactor cases of 

OTIB-0054 are HTGRs. 

Priority Ranking: High

35. Organic Moderated 

Reactor Experiment 

(OMRE)

Separate area a few miles 

east of Central Facilities 

Area (CFA), between Waste 

Area Groups 4 and 5. 

1957 – 1963 The OMRE reactor, built by Atomics International, 

was part of an Atomic Energy Commission program 

to assess the feasibility and determine the nuclear 

and engineering technical basis of different reactor 

concepts in support of an emerging civilian nuclear 

power industry. OMRE used a waxy liquid 

hydrocarbon (known as “Sanowax”) rather than 

water or a liquid metal as both coolant and 

moderator. The relatively low-power (5–10 MWth), 

critical reactor tested various types and 

configurations of highly enriched uranium dioxide

(UO2) fuel elements. 

It is doubtful whether OMRE would be 

adequately enveloped by any of the OTIB-0054 

cases because it used an organic coolant and 

moderator. 

Priority Ranking: High
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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

36. Power Burst Facility 

(PBF) 

Near the SPERT-I site. 

1972–1985 The PBF continued the reactor safety program begun 

with the Special Power Excursion Reactor Test 

(SPERT) series of facilities but was much larger than 

the SPERT reactors. Fuel and cladding combinations 

were varied and tested to failure. Transient testing, 

including LOCA scenarios that modeled designbasis and severe accident conditions at a nuclear 

power plant, led to fuel and cladding damage and the 

subsequent evolution of hydrogen and the release of 

fission products to the reactor containment. As its 

name implies, the PBF could produce very high, 

short-duration (millisecond) power excursions that 

were self-limiting. It could operate at a steady-state 

power of 20 MWe for a short period of time before 

initiating a very short super-critical power burst. 

The reactor was water-cooled and uranium-oxidefueled, with stainless steel cladding. The core 

consisted of a square array of 121 square cells, 

5.85 in × 5.85 in × 60 in, with an active fuel region 

of about 36 in located in a pressure vessel. Different 

fuel configurations, compositions, and enrichments 

were tested. The reactor was water cooled, 

moderated, and reflected. Some of the experiments 

used previously irradiated fuel (e.g., 38,000 

MWd/MTU) to better simulate accident conditions. 

The PBF used a variety of new and previously 

irradiated fuel and often operated to fuel failure, 

thus putting it outside the range of the nine 

representative cases of OTIB-0054. 

Priority Ranking: High
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39. Special Power 

Excursion Reactor Test No. 

I (SPERT-I) 

Separate complex east of 

CFA

1955–1964 The SPERT series of reactors, of which there were 

four, were deliberately subjected to large, rapid 

reactivity excursions in order to gather data on 

coupled neutronic and thermal-hydraulic responses 

as part of an Atomic Energy Commission safety 

assessment program in support of commercial 

pressurized and boiling water nuclear power plants. 

The many SPERT experiments, which varied fuel 

design, core configurations, reflectors, moderators, 

coolant flow, temperatures, and pressures, supplied 

data for development and validation of computer 

codes to simulate reactor dynamics and for 

establishing safe operating limits. The SPERT series 

started out with thin, aluminum or stainless steel 

clad, uranium fuel plates but later transitioned to fuel 

rods, which were more typical of power reactors. 

SPERT-I, the first reactor of the series, was an opentank, light water moderated and reflected reactor, 

with the uranium fuel enriched to 93.5%. Some 

experiments were also conducted with fuel enriched 

only a few percent to better simulate power reactor 

fuel. The fuel considered of plate type uranium and 

aluminum fuel assemblies (about 25" long) in a 

4-foot diameter and 14-foot deep carbon steel tank, 

lined with aluminum cladding. Fuel burnup was 

quite low because the reactor operated in the 

transient rather than the steady-state mode. 

While SPERT-I experiments operated outside 

established design limits, conditions were usually 

kept below those producing core damage. However, 

a planned 2,300 MWth excursion on November 5, 

1962, destroyed the core and distorted the reactor 

vessel. Subsequently, SPERT-I was rebuilt, and lowenriched fuel rods replaced the previously used highenriched fuel plates. A planned 17,400 MWth

The SPERT series modeled a commercial LWR 

but often operated beyond usual limits to 

explore coupled nuclear and thermal-hydraulic 

behavior in design-basis or beyond-design-basis 

accident scenarios. It also varied fuel and reactor 

component materials and designs. 

Priority Ranking: High
Effective Date:

March 2, 2016

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

excursion on November 12, 1963 (with 4% uranium 

oxide fuel rods), and a planned 35,000 MWth

excursion on April 14, 1964 (with 4% uranium oxide 

fuel rods), tested the resilience of the fuel rods; the 

latter test damaged some of the fuel rods. 

40. Special Power 

Excursion Reactor Test No. 

II (SPERT-II) 

Separate complex east of 

CFA

1960–1964 SPERT-II construction followed SPERT-III and 

continued to investigate transient behavior. Several 

different types of fuel assemblies were used, both 

light and heavy were tested as moderators and 

coolants, and different reflectors were also used. 

Unlike SPERT-I, SPERT-II was placed in a closed 

pressure vessel. 

The active length of the flat-plate fuel assemblies 

were about 24 inches. Each fuel plate consisted of an 

enriched uranium-aluminum alloy. The fuel was 

enriched to 93.5% in uranium-235 (U-235). Since 

the reactor operated in the transient, burst mode, 

total burnup was small. 

The SPERT series modeled a commercial LWR 

but often operated beyond usual limits to 

explore coupled nuclear and thermal-hydraulic 

behavior in design-basis or beyond-design-basis 

accident scenarios. It also varied fuel and reactor 

component materials and designs. 

Priority Ranking: High
Effective Date:

March 2, 2016

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0 (Draft)

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

41. Special Power 

Excursion Reactor Test No. 

III (SPERT-III)

Separate complex east of 

CFA

1958–1968 SPERT-III accommodated the widest variation in

several important parameters: temperature, pressure, 

and coolant flow. The core sat in a pressure vessel 

similar to that used in a commercial nuclear power 

plant, and the maximum operating temperature of 

668 °

F and pressure of 2,500 psig also simulated 

power plant conditions. The system could produce a 

maximum of 60 MWth for about 30 min of operating 

time, limited by the capacity of the heat removal 

system. The fuel plates contained 4.8% enriched 

UO2 clad in stainless steel; the overall core 

dimensions were about 2 ft diameter × 3 ft height. 

The reactor used ordinary water as coolant and 

moderator. 

The SPERT series modeled a commercial LWR 

but often operated beyond usual limits to 

explore coupled nuclear and thermal-hydraulic 

behavior in design-basis or beyond-design-basis 

accident scenarios. It also varied fuel and reactor 

component materials and designs. 

Priority Ranking: High

42. Special Power 

Excursion Reactor Test No. 

IV (SPERT-IV)

Separate complex east of 

CFA

1962–1970 SPERT-IV also investigated transient reactor 

behavior to provide neutronic and thermal-hydraulic 

data applicable especially to large, open pool 

reactors. The reactor was located in an open tank,

and the fuel consisted of a 93.5% enriched uraniumaluminum matrix. The facility utilized a number of 

different cores and other components and was 

operated over a wide range of several different 

parameters. Test scenarios included fuel destruction 

experiments. 

The SPERT series modeled a commercial LWR 

but often operated beyond usual limits to 

explore coupled nuclear and thermal-hydraulic 

behavior in design-basis or beyond-design-basis 

accident scenarios. It also varied fuel and reactor 

component materials and designs. 

Priority Ranking: High
Effective Date:

March 2, 2016

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0 (Draft)

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

43. Spherical Cavity 

Reactor Critical 

Experiment (SCRCE) 

TAN – WRRTF – LPTF

1972–1973 Located in a shielded test cell of the LPTF in the 

WRRTF of TAN, SCRCE was a NASA-sponsored 

program to determine the feasibility and explore the 

neutronics characteristics of a reactor going critical 

with a core of highly-enriched gaseous uranium 

(UF6) and a deuterium oxide (D2O) reflector and 

moderator. Heat from the reactor would be used to 

heat a gas that would propel a space vehicle. The 

low-power (<100 W) experiment was located in a 

shielded cell of the LPTF. Several different 

configurations were measured in the test series.

SCRCE, with a gaseous core of UF6 and a D2O 

reflector, lies totally outside the envelope of any 

of the OTIB-0054 representative reactors. 

Priority Ranking: High

46. Systems for Nuclear 

Auxiliary Power (SNAP) 

10A Transient No. 1 

(SNAPTRAN–1)

TAN – IETF

Early 1960s The Atomic Energy Commission’s Systems for 

Nuclear Auxiliary Power Transient (SNAPTRAN)

program, located at the Initial Engine Test Facility 

(IETF) of TAN, extended the SPERT reactor safety 

testing program to aerospace applications by testing 

the SNAP 10A/2 reactor under extreme conditions. 

The SNAP 10A/2 reactor was a compact cylinder, 

with a core about 9 inches in diameter and 13 inches 

high, with 37 1.212-inch diameter stainless steel-clad 

fuel rods. The reactors were fully enriched, 

zirconium hydride-uranium fueled, sodiumpotassium (NaK) cooled, and beryllium reflected. 

SNAPTRAN-1 was subjected to nondestructive, 

large-transient tests in conditions approaching but 

not resulting in damage to the fuel. The series of 

tests investigated the effect on the fuel of large 

power transients. 

The SNAPTRAN series often operated beyond 

usual limits to explore coupled nuclear and 

thermal-hydraulic behavior in design-basis or 

beyond-design-basis accident scenarios. It also 

varied fuel and reactor component materials and 

designs. 

Priority Ranking: High
Effective Date:

March 2, 2016

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0 (Draft)

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

47. Systems for Nuclear 

Auxiliary Power (SNAP) 

10A Transient No. 3 

(SNAPTRAN–3)

TAN – IETF

1964 SNAPTRAN-3 followed SNAPTRAN-2 and 

included a destructive test on April 1, 1964, to 

simulate the accidental fall of a reactor into water or 

wet earth following an accident on a nuclearpowered aircraft. The test demonstrated that the 

reactor would destroy itself immediately instead of 

building up a high inventory of radioactive fission 

products. The SNAP 10A/2 reactor operated at 30 

GW for 1.5 milliseconds. 

The SNAPTRAN series often operated beyond 

usual limits to explore coupled nuclear and 

thermal-hydraulic behavior in design-basis or 

beyond-design-basis accident scenarios. It also 

varied fuel and reactor component materials and 

designs. 

Priority Ranking: High

48. Systems for Nuclear 

Auxiliary Power (SNAP)

10A Transient No. 2 

(SNAPTRAN–2)

TAN – IETF

1965–1966 SNTAPTRAN-2 followed SNAPTRAN-3. The 

SNAP 10A/2 reactor was intentionally destroyed on 

January 11, 1966. It provided information on the 

dynamic response, fuel behavior, and inherent 

shutdown mechanisms of these reactors in an open 

air environment. The reactor produced a peak power 

of 74 GW for 1.5 milliseconds before it destroyed 

itself. 

The SNAPTRAN series often operated beyond 

usual limits to explore coupled nuclear and 

thermal-hydraulic behavior in design-basis or 

beyond-design-basis accident scenarios. It also 

varied fuel and reactor component materials and 

designs. 

Priority Ranking: High
Effective Date:

March 2, 2016

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0 (Draft)

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

MEDIUM PRIORITY RANKING

2. Advanced Reactivity 

Measurement Facility No. 

2 (ARMF-II). Renamed 

Coupled Fast Reactivity 

Measurement Facility 

(CFRMF) in 1968

TRA (Bldg. TRA-660)

1962–1991 The ARMF-II occupied the opposite end of the tank 

containing the ARMF-I and was similar to that 

reactor but had some additional features to improve 

its accuracy. In addition to measuring the nuclear 

properties of reactor fuels and materials, the reactor 

was also used for neutron radiography. The ARMFII was modified in 1968 to enable it to measure the 

nuclear characteristics of fast reactor fuels and 

materials; it was then renamed the Coupled Fast 

Reactivity Measurement Facility (CFRMF).

The reactor used MTR/ETR/ATR-type fuel but 

operated only periodically and at a low power 

level, resulting in low burnup of the fuel. Hence, 

it might not be enveloped by any of the three 

ATR cases of OTIB-0054, but modeling it with 

one of the ATR cases might be claimant 

favorable, since the ATR cases would have a 

greater buildup of longer-lived radionuclides. 

However, when operating with fast reactor fuels 

(as the CFRMF) it might not be adequately be 

enveloped by either an ATR case or a Fast 

Reactor Test Facility (FFTF) (fast reactor) case. 

Priority Ranking: Medium

12. Coupled Fast Reactivity 

Measurement Facility 

(CFRMF). Formerly 

named Advanced 

Reactivity Measurement 

Facility No. 2 (ARMF-II)

TRA (Bldg. TRA-660)

1968–1991 When the ARMF-II reactor was modified in 1968, it 

was given a new name, the CFRMF. A section of the 

core was modified to produce a region of highenergy neutron flux that was useful in comparing 

calculated and observed results. This tool provided 

physics information about the behavior of fast (i.e., 

unmoderated) neutrons. Physicists studied 

differential cross-sections and tested calculational 

methods. The CFRMF contributed to the 

development of fast neutron reactors.

The CFRMF operated only periodically and at a 

low power level, resulting in low burnup of the 

fuel. When running as a fast reactor, it might not 

be adequately enveloped by either an ATR case 

or a FFTF (fast reactor) case. 

Priority Ranking: Medium
Effective Date:

March 2, 2016

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0 (Draft)

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

13. Critical Experiment 

Tank (CET)

TAN – WRRTF – LPTF

1958–1960 Three low-power reactors supported the Aircraft 

Nuclear Propulsion (ANP) program by testing 

various components and collecting nuclear physics 

data: the Critical Experiment Tank (CET), the Hot 

Critical Experiment (HOTCE), and the Shield Pool 

Test Facility (STPF) Reactor (SUSIE). They were 

located in the LPTF, which was part of the WRRTF

of TAN. 

The water-moderated and reflected CET was a lowpower reactor that was originally intended to 

simulate the Heat Transfer Reactor Experiment 

(HTRE)-1 and HTRE-2 reactors in the ANP 

program, and to perform critical experiments of 

HTRE fuel bundles. It was also used as a source of 

neutrons to calibrate neutron sensors. 

It is not clear without a more in-depth 

investigation whether the CET would be 

adequately enveloped by any of the OTIB-0054 

cases. 

Priority Ranking: Medium

20. Fast Spectrum 

Refractory Metals Reactor 

(710)

TAN – WRRTF – LPTF

1962–1968 Located in a shielded test cell of the LPTF in the 

WRRTF of TAN, this low-power, split-table, critical 

facility collected data for a proposed fast-spectrum, 

compact, refractory-metal reactor concept for 

generating power in space.

It is not clear without a more in-depth study if 

any of the nine OTIB-0054 cases would 

envelope a split-table, fast reactor. The power 

level is low, though. 

Priority Ranking: Medium

26. Hot Critical 

Experiment (HOTCE)

TAN – WRRTF – LPTF

1958–1961 Three low-power reactors supported the ANP 

program by testing various components and 

collecting nuclear physics data: the CET, the 

HOTCE, and the STPF Reactor (SUSIE). They were 

located in the LPTF, which was part of the WRRTF

of TAN. 

HOTCE was a low-power, high-temperature critical 

reactor.

It is not clear without a more in-depth 

investigation whether the HOTCE would be 

adequately enveloped by any of the OTIB-0054 

cases. 

Priority Ranking: Medium
Effective Date:

March 2, 2016

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0 (Draft)

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

34. Nuclear Effects Reactor 

(FRAN)

ARA-IV

1968–1970 FRAN was a small, prompt-burst reactor that could 

go super critical for a short time, producing a 

copious amount of fast neutron and gamma radiation 

inside its annular void, where samples would be 

placed. Expansion of its fuel assembly would 

quickly lower the reactivity, thereby controlling the 

excursion. The bare cylindrical assembly was fueled 

with 93.5% enriched uranium and clad with nickel 

and cadmium. FRAN was used to test new detector 

systems and to provide heat transfer and nuclear 

physics information of materials subjected to intense 

fast-neutron bombardment. The reactor was moved 

to the Lawrence Livermore National Laboratory 

(current name) in 1970. 

Note: FRAN appears in Stacy 2000 and other 

literature that was examined, but not in the INL 

site profile. 

Priority Ranking: Medium

38. Shield Test Pool 

Facility (STPF – SUSIE)

TAN – WRRTF – LPTF

(Bldg. TAN-646)

Early 1960s Three low-power reactors supported the ANP 

program by testing various components and 

collecting nuclear physics data: the CET, the 

HOTCE, and the STPF Reactor (SUSIE). They were 

located in the LPTF, which was part of the WRRTF

of TAN. 

The SUSIE reactor was used for bulk shielding 

experiments that were performed in support of the 

ANP Shielding Experimentation Program. The 

water-moderated, -reflected, and –cooled reactor, 

situated in a water-filled pool at TAN, could be 

operated safely, was adaptable to many forms of 

nuclear research, and was easy to operate at 

minimum cost. The reactor’s nominal power level 

was 2 MWth from aluminum-clad U-235 fuel. 

After the ANP program was discontinued in 1961, 

SUSIE continued in use by other programs.

It is not clear without a more in-depth 

investigation whether SUSIE would be 

adequately enveloped by any of the OTIB-0054 

cases. 

Priority Ranking: Medium
Effective Date:

March 2, 2016

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

49. Thermal Reactor Idaho 

Test Station (THRITS) 

TAN – WRRTF – LPTF

1964 THRITS, located in an LPTF cell, was a low-power 

reactor with a split-table core where the two halves 

were brought together to make a critical assembly. 

Operators mocked up reactor design concepts for 

thermal and fast neutron reactor systems to obtain 

basic physics and design data for such concepts. The 

THRITS fuel consisted of polyethylene interspersed 

with enriched uranium foils, producing a thermal 

neutron spectrum.

It is not clear without a more in-depth study if 

any of the nine OTIB-0054 cases would 

envelope a split-table reactor. The power level is 

low, though. 

Priority Ranking: Medium

LOW PRIORITY RANKING

1. Advanced Reactivity 

Measurement Facility No. 

1 (ARMF-I)

TRA (Bldg. TRA-660)

1960–1974 The ARMF-I was a very small, water pool reactor, 

operating up to 100 kWth on highly enriched 

uranium, located in a water tank near the MTR. It 

was used to determine nuclear properties, such as 

reactivity, of small samples placed in it. The reactor 

replaced the Reactivity Measurement Facility 

(RMF). 

The reactor used MTR/ETR/ATR-type fuel but 

operated only periodically and at a low power 

level, resulting in low burnup of the fuel. Hence, 

it might not be enveloped by any of the three 

ATR cases of OTIB-0054, but modeling it with 

one of the ATR cases might be claimant 

favorable, since the ATR cases would have a 

greater buildup of longer-lived radionuclides.

Priority Ranking: Low

4. Advanced Test Reactor 

Critical Facility (ATRC)

TRA (Bldg. TRA-670)

1964–present The low power, highly enriched ATRC performs 

functions for the ATR similar to those of the 

Engineering Test Reactor Critical Facility (ETRC)

reactor for the MTR. The full-sized reactor tests fuel 

and experiment configurations destined for the ATR, 

but at low power levels. The ATRC is designed to 

operate at 5 kWth steady state but is usually operated 

at about 500 Wth. 

The reactor uses MTR/ETR/ATR-type fuel but 

operates only periodically and at a low power 

level, resulting in low burnup of the fuel. Hence, 

it might not be enveloped by any of the three 

ATR cases of OTIB-0054, but modeling it with 

one of the ATR cases might be claimant 

favorable, since the ATR cases would have a 

greater buildup of longer-lived radionuclides.

Priority Ranking: Low
Effective Date:

March 2, 2016

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0 (Draft)

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

15. Engineering Test 

Reactor Critical Facility 

(ETRC)

TRA (Bldg. TRA-654)

1957–1982 The ETRC was operated as a full-scale, low-power 

version of the ETR that was used to determine the 

nuclear characteristics of experiments that would 

later be irradiated in the ETR. 

The reactor used MTR/ETR/ATR-type fuel but 

operated only periodically and at a low power 

level, resulting in low burnup of the fuel. Hence, 

it might not be enveloped by any of the three 

ATR cases of OTIB-0054, but modeling it with 

one of the ATR cases might be claimant 

favorable, since the ATR cases would have a 

greater buildup of longer-lived radionuclides.

Priority Ranking: Low

21. Gas Cooled Reactor 

Experiment (GCRE) 

ARA-III

1960–1961 The GCRE, located at the ARA-III site, was the 

Army’s initial effort at developing a low-power, 

nitrogen-cooled, water-moderated mobile nuclear 

power plant. The GCRE was built by Aerojet 

General Nucleonics, went critical on February 23, 

1960, and operated until April 1961. The GCRE 

generated 2.2 MWth from either plate-type or pintype fuel; the former was housed in an aluminum 

calandria and the latter in a stainless steel calandria. 

The GCRE provided nuclear and engineering data, 

as well as served as a training platform for Army and 

civilian personnel. 

Priority Ranking: Low

25. High Temperature 

Marine Propulsion Reactor 

(630-A) 

TAN – WRRTF – LPTF

1962–1964 Located in a shielded test cell of the LPTF in the 

WRRTF of TAN, the 630-A reactor, a low-power 

critical experiment, explored the feasibility of an aircooled, water-moderated system for nuclearpowered merchant ships. 

Priority Ranking: Low
Effective Date:

March 2, 2016

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0 (Draft)

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

37. Reactivity 

Measurement Facility 

(RMF) 

TRA (Bldg. TRA-603)

1954–1962 The RMF, a water pool reactor with a highly 

enriched water-cooled, -moderated, and -reflected 

uranium core (MTR-type fuel) and a power level of 

100 or 200 Wth, was located in a canal in the 

basement of the MTR building. It was used to assay 

new and spent fuel assemblies for the MTR and ETR 

and to measure nuclear properties. It was replaced by 

the ARMF-I, which became operational in 1964. 

The reactor used MTR/ETR/ATR-type fuel but 

operated only periodically and at a low power 

level, resulting in low burnup of the fuel. Hence, 

it might not be enveloped by any of the three 

ATR cases of OTIB-0054, but modeling it with 

one of the ATR cases might be claimant 

favorable, since the ATR cases would have a 

greater buildup of longer-lived radionuclides.

Priority Ranking: Low

44. Stationary Low-Power 

Reactor (Earlier name:

Argonne Low Power 

Reactor) (SL-1, ALPR) 

ARA-II

1958–1961 The SL-1 was a low-power boiling water reactor, 

with a design power of 3 MWth. It was designed by 

Argonne National Laboratory, based on Boiling 

Water Reactor Experiment (BORAX) experience, as 

a prototype of a reactor that could be used by the 

Army in geographically remote locations where 

fossil fuel was difficult to obtain, such as the Arctic. 

The fuel, which was 93% enriched, was in the form 

of aluminum-uranium alloy plates, and light water 

served as coolant and moderator. 

A criticality accident occurred on January 3, 1961,

due to the central control rod being manually 

withdrawn too far in an attempted startup after a 

maintenance period. The withdrawal of the control 

rod, which has a large reactivity worth, caused the 

power level to spike at about 20 GW in a few 

milliseconds. This resulted in a violent steam 

explosion and core meltdown that killed three 

workers in the building at the time and released 

fission products in the building and to the 

atmosphere, as well as substantial contamination to 

the area around the building; the reactor building 

The ATR cases might adequately envelope the 

SL-1 (other than its meltdown). 

Priority Ranking: Low
Effective Date:

March 2, 2016

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

was fabricated out of steel plate and was not 

designed as a containment structure. 

EXCLUDED FROM PRIORITIZATION PROCESS (ANL-W, NRF, ALREADY EVALUATED, OR NEVER OPERATED)

3. Advanced Test Reactor 

(ATR)

TRA (Bldg. TRA-670)

1967–present ORAUT-OTIB-0054 uses the ATR as a surrogate for 

high-flux reactors (see Table 1), as might be 

encountered in materials testing or experimental 

reactors. The ATR, which has a design power level 

of 250 MWth, is the latest and largest of three 

materials testing reactors at INL (the other two are 

the MTR and the ETR, both of which no longer 

operate). The ATR is a pressurized, light-water 

moderated, beryllium-reflected reactor, using highly 

enriched uranium fuel (93.15% nominal enrichment) 

arranged in an unusual curved plate configuration, 

and employing a unique design of rotating beryllium 

cylinder shells as the primary reactivity control 

mechanism. The reactor’s four-lobed design supports 

nine main test spaces, or loops.

Reviewed in SC&A report (SC&A 2015a): 

NIOSH SEC-00219: Test Reactor Area 

Modeling, SCA-SEC-2015-0074-C, Revision 0, 

September 28, 2015: “ORAUT-OTIB-0054 

explicitly models the ATR, so it is expected that 

any workers exposed to ATR fuel during reactor 

operations or when the fuel is out of the reactor, 

but before the isotopic constituents might have 

been separated, would be adequately treated by 

the methodology in the OTIB in order to 

determine internal exposures. SC&A in its 

investigations did not find instances of the ATR 

operating outside of its design envelope.”

Priority Ranking: N/A

5. Argonne Fast Source 

Reactor (AFSR) 

ANL-W

1959–late 70s The AFSR, with a design power level of 1 kWth, 

fueled with high-enriched uranium, and with a 

blanket of solid depleted uranium, supplied neutrons 

to test and develop equipment (e.g., radiation 

detectors) for the ANL-W fast reactor program.

ANL-W is not included in the SEC-0219 

definition. 

Priority Ranking: N/A
Effective Date:

March 2, 2016

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Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

6. Boiling Water Reactor 

Experiment No. 1 

(BORAX-I)

ANL-W

1953–1954 The BORAX series of reactor experiments, 

beginning with the operation of BORAX-I in 1953 

(1.4 MWth), tested the feasibility and safety and 

explored the operating parameters of direct steam 

production in an LWR. The final, planned 

destructive test of BORAX-I in 1954 melted the 

core, destroyed the reactor, and released fuel and 

fission products. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

7. Boiling Water Reactor 

Experiment No. 2 

(BORAX-II) 

ANL-W

1954–1955 BORAX-II, at 6 MWth, continued the boiling water 

reactor testing program. It operated with varying 

enrichments of uranium in its fuel plates. It was 

intentionally destroyed in 1955 by taking it prompt 

critical. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

8. Boiling Water Reactor 

Experiment No. 3 

(BORAX-III) 

ANL-W

1955–1956 BORAX-III, at 15 MWth, was connected to a 2 MWe

steam turbine-generator to investigate radioactive 

contamination of the turbine and to demonstrate the 

production of electricity. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

9. Boiling Water Reactor 

Experiment No. 4 

(BORAX-IV) 

ANL-W

1956–1958 BORAX-IV, at 20 MWth, tested uranium and 

thorium ceramic fuel plates, some of which 

purposefully contained defects to determine reactor 

behavior with compromised fuel. The tests released 

some short-lived radionuclides to the atmosphere. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

10. Boiling Water Reactor 

Experiment No. 5 

(BORAX-V) 

ANL-W

1962–1964 BORAX-V, at 40 MWth, added an integral, nuclear 

superheat system to the BORAX experiments to 

improve thermal efficiency of the boiling water 

reactor plant

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A
Effective Date:

March 2, 2016

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NOTICE: This report has been reviewed to identify and redact any information that is protected by the Privacy Act 5 USC §552a and has been cleared for 

distribution.

Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

14. Engineering Test 

Reactor (ETR)

TRA (Bldg. TRA-642)

1957–1981 The ETR, which was larger and had higher flux than 

the MTR and was located next to it, was designed to 

take advantage of information gathered and lessons 

learned from its predecessor materials test reactor. It 

operated at a maximum power level of 175 MWth. It 

eliminated the beam holes of the MTR, with all 

experiments taking place within the core. 

Reviewed in SC&A report (SC&A 2015a): 

NIOSH SEC-00219: Test Reactor Area 

Modeling, SCA-SEC-2015-0074-C, Revision 0, 

September 28, 2015: “the OTIB-0054 

methodology, which explicitly modes the ATR, 

should also adequately envelope the ETR in 

considering external exposures.”

Priority Ranking: N/A

16. Experimental 

Beryllium Oxide Reactor 

(EBOR)

TAN – WRRTF – LPTF 

(Bldg. TAN-646)

Never operated Construction began in May 1963, but the project was 

canceled in 1966 before construction was complete. 

Never operated

Priority Ranking: N/A

17. Experimental Breeder 

Reactor No. I (EBR-I)

ANL-W

1951–1963 The EBR-I fast breeder reactor demonstration was 

the first reactor built at INL. It had a maximum 

power of 1 MWth from its enriched uranium core, 

cooled by the liquid metal NaK, and surrounded by a 

U-238 breeding blanket. In 1951, it became the first 

reactor to generate electricity from an attached steam 

turbine-generator, which lit a few lightbulbs adjacent 

to the reactor. The EBR-I also demonstrated the 

feasibility of breeding plutonium in a reactor. About 

half of the core inadvertently melted in 1955 during 

a coolant flow test. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A
Effective Date:

March 2, 2016

Revision No.

0 (Draft)

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NOTICE: This report has been reviewed to identify and redact any information that is protected by the Privacy Act 5 USC §552a and has been cleared for 

distribution.

Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

18. Experimental Breeder 

Reactor No. II (EBR-II) 

ANL-W

1961–1994 The EBR-II continued fast neutron breeder reactor 

development at ANL-W, including onsite 

reprocessing of spent fuel into new fuel pins. The 

core, with 67% enriched U-235 fuel, sat in a tank of 

liquid sodium (Na) coolant, had a closed-loop Na 

coolant system, and produced steam in a tertiary 

system. The reactor was unmoderated. The entire 

system was placed in a containment building. The 

maximum power level was 62.5 MWth, and the EBRII could supply 20 MWe of electric power to INL 

facilities. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

19. Experimental Organic 

Cooled Reactor (EOCR) 

CFA (vicinity)

Never operated The EOCR was intended to test the organic-coolant 

concept beyond the Organic Moderated Reactor 

Experiment (OMRE) but was placed in standby in 

1962 and never operated. 

Never operated.

Priority Ranking: N/A
Effective Date:

March 2, 2016

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0 (Draft)

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NOTICE: This report has been reviewed to identify and redact any information that is protected by the Privacy Act 5 USC §552a and has been cleared for 

distribution.

Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

22. Heat Transfer 

Experiment No. 1 

(HTRE-1)

TAN – IETF

1955–1959 The three HTRE reactors (designed by General 

Electric) and the associated 26 Initial Engine Tests 

(IETs), some of which were non-nuclear, were an 

important part of the Air Force’s ANP program. 

They explored the characteristics of direct-cycle heat 

transfer engineering applied to a turbojet engine 

intended for an aircraft. Fuel utilized highly enriched 

uranium in the form of very thin concentric ribbons 

to maximize heat transfer from the compact reactors. 

The reactors were located in the IETF of TAN and, 

due to the nature of the HTRE design, released 

substantial amounts of radionuclides to the 

environment through a 150-ft exhaust stack.

HTRE-1 used enriched metallic nickel-chromiumoxide fuel elements and was water moderated and 

cooled. The reactor produced 20 MWth and was 

operated for 150.8 hours at full power. 

SC&A (2015b) evaluated the three HTRE 

reactors and concluded that OTIB-0054 does not 

appear to adequately envelope them. 

Priority Ranking: N/A

23. Heat Transfer 

Experiment No. 2 

(HTRE-2)

TAN – IETF

1957–1961 The HTRE-2 core was similar to the HTRE-1 core, 

but with provisions for operating as a materials 

testing reactor for experimental fuel sections 

introduced into a hexagonal center hole. The reactor 

operated for 1,299 hours at power levels up to 14

MWth and temperatures of 2,800 °

F for extended 

periods. Total burnup amounted to about 

5,000 MW-hr. 

SC&A (2015b) evaluated the three HTRE 

reactors and concluded that OTIB-0054 does not 

appear to adequately envelope them. 

Priority Ranking: N/A
Effective Date:

March 2, 2016

Revision No.

0 (Draft)

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NOTICE: This report has been reviewed to identify and redact any information that is protected by the Privacy Act 5 USC §552a and has been cleared for 

distribution.

Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

24. Heat Transfer 

Experiment No. 3 

(HTRE-3)

TAN – IETF

1958–1960 The HTRE-3 was built in a full-scale aircraft reactor 

configuration, where the reactor and other 

components were arranged horizontally to simulate 

their locations in an actual airplane. The reactor was 

water cooled and water moderated, with uranium 

fuel and nickel-chromium cladding. The hydrided 

zirconium moderated was air cooled. The reactor 

underwent an unplanned nuclear excursion on 

November 18, 1958, releasing radionuclides up the 

stack and depositing contamination on the site from 

fallout. The system operated for a total of 126 hours.

SC&A (2015b) evaluated the three HTRE 

reactors and concluded that OTIB-0054 does not 

appear to adequately envelope them. 

Priority Ranking: N/A

27. Large Ship Reactor A 

(A1W-A)

NRF

1958–1994 The large ship reactor facility (A1W) contained two 

Westinghouse pressurized water reactors (designated 

A and B), placed inside a section of a steel hull, that 

were prototypes for those installed in the USS 

Enterprise aircraft carrier. 

NRF is not included in the EEOICPA program.

Priority Ranking: N/A

28. Large Ship Reactor B 

A1W-B 

NRF

1959–1987 [See A1W-A description] NRF is not included in the EEOICPA program.

Priority Ranking: N/A
Effective Date:

March 2, 2016

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0 (Draft)

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NOTICE: This report has been reviewed to identify and redact any information that is protected by the Privacy Act 5 USC §552a and has been cleared for 

distribution.

Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

30. Materials Test Reactor 

(MTR)

TRA (Bldg. TRA-603) 

1952–1970 The MTR was the first materials test reactor at INL 

(it was succeeded by the ETR and the currently 

operating ATR). It was cooled and moderated with 

light water and used Al-clad, curved plate, highly 

enriched uranium fuel most of the time. Beryllium, 

graphite, and light-water neutron reflectors 

surrounded the relatively small core. Experiments 

were conducted external to the core, utilizing 

neutron fluxes exiting through beam holes. Its 

maximum power level increase from 30 MWth to 

40 MWth in 1955. In 1958, it became the first reactor 

to operate with a plutonium-239 (Pu-239) core at up 

to 30 MWth in a demonstration project. 

Reviewed in SC&A report (SC&A 2015a): 

NIOSH SEC-00219: Test Reactor Area 

Modeling, SCA-SEC-2015-0074-C, Revision 0, 

September 28, 2015: “the OTIB-0054 

methodology, which explicitly models the ATR, 

should also adequately envelope the MTR in 

considering internal exposures when the latter 

reactor operated with uranium fuel.”

“The MTR also ran for a period of time with 

plutonium rather than uranium fuel, and at a 

lower power level and lower burnup. It is not 

clear which, if any, of the nine OTIB-0054 

representative cases...would adequately 

envelope this situation.”

Priority Ranking: N/A

32. Natural Circulation 

Reactor (S5G) 

NRF

1965–1995 The S5G, designed by General Electric, was the 

prototype of a pressurized water reactor for USS 

Narwhal. It could operate in a natural circulation 

cooling mode, which greatly reduced the noise 

produced, thereby making it more difficult to detect 

the submarine. 

NRF is not included in the EEOICPA program.

Priority Ranking: N/A

33. Neutron Radiography 

Facility (NRAD)

ANL-W

1977–present NRAD is a 250-kWth, uranium-fueled, water pooltype, steady-state General Atomics TRIGA reactor. 

It is used primarily for non-destructive neutron 

radiography of irradiated fuel specimens. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A
Effective Date:

March 2, 2016

Revision No.

0 (Draft)

Document No./Description:

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NOTICE: This report has been reviewed to identify and redact any information that is protected by the Privacy Act 5 USC §552a and has been cleared for 

distribution.

Reactor Name & 

Locationa,b Operation Dates Summary Description Comments/Priority Ranking

45. Submarine Thermal 

Reactor (S1W, STR). Also 

known as the Submarine 

Prototype Reactor 

NRF

1953–1989 The Westinghouse S1W (Ship 1, Westinghouse) 

reactor was a prototype of the reactor that was 

installed in the USS Nautilus submarine. The S1W 

was located in a simulated submarine hull section 

and was used for testing and training purposes. 

NRF is not included in the EEOICPA program.

Priority Ranking: N/A

50. Transient Reactor Test 

Facility (TREAT)

ANL-W 

1959–1994 The TREAT experiment continued research and 

development for fast breeder reactors by simulating 

accident conditions leading to fuel damage. It was an 

air-cooled, graphite-moderated, uranium-fueled 

reactor, operated at a steady power level of 

100 kWth, and could produce pulses of up to 19 GWth

for experimental purposes. The fuel consisted of a 

93% enriched graphite-uranium matrix, with 

zirconium clad fuel assemblies. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

51. Zero Power Physics 

Reactor (Earlier name:

Zero Power Plutonium 

Reactor) (ZPPR)

ANL-W

1969–1992 ZPPR was a very low-power, air-cooled, split-table 

critical facility that was used to mock up other fast 

neutron spectrum reactors. It was the largest in the 

series of split-table assemblies that were designed to 

support the plutonium-fueled liquid metal fast 

breeder reactor program. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

52. Zero Power Reactor 

No. 3 (ZPR-III)

ANL-W

1955–1970 ZPPR-III was the smallest of the ZPPR series of 

split-table critical facilities. 

ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

a The list and numbering scheme of the 52 INL reactors was taken from Stacy 2000. b Location acronyms (current names are used in most cases): ANL-W = Argonne National Laboratory-West; ARA = Auxiliary Reactor Area; CFA = Central 

Facilities Area; IETF = Initial Engine Test Facility; NRF = Naval Reactor Facility; LPTF = Low Power Test Facility; TAN = Test Area North; TRA = Test 

Reactor Area; WRRTF = Water Reactor Research Test Facility.