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 1 

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

June 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.
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

2 of 44

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.

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.: 1 (Draft)

SUPERSEDES: 0

EFFECTIVE DATE: June 10, 2016

TASK MANAGER: John Stiver, MS, CHP [signature on file]

PROJECT MANAGER: John Stiver, MS, CHP [signature on file]

DOCUMENT 

REVIEWER(S):

John Stiver, MS, CHP [signature on file] 

John Mauro, PhD, CHP [signature on file]

Record of Revisions 

Revision 

Number Effective Date Description of Revision

0 (Draft) 03/02/2016 Initial issue

1 06/10/2016 1.0 – summarized comments received on Rev. 0; 3.0 – added 

Table 3 for convenience in locating reactors in Attachment 1; 

3.0 – revised Table 5 to reflect changed prioritization of 

Attachment 1 and grouped some of the reactors together for joint 

evaluations; 4.0 – added references and SRDB numbers where 

available; Attachment 1 – expanded summary descriptions and 

reclassified some priority rankings; General – editorial revisions 

throughout the document.
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

3 of 44

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.

TABLE OF CONTENTS

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

1.0 Introduction ............................................................................................................................... 8

2.0 ORAUT-OTIB-0054 ............................................................................................................... 11

3.0 Evaluation ............................................................................................................................... 13

4.0 References ............................................................................................................................... 17

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

Investigation.................................................................................................................................. 19
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

4 of 44

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.

ABBREVIATIONS AND ACRONYMS

Advisory Board Advisory Board on Radiation and Worker Health

AEC Atomic Energy Commission 

AFSR Argonne Fast Source Reactor 

AGCRSP Army Gas-Cooled Reactor Systems Program

Al aluminum

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

Be beryllium 

BORAX Boiling Water Reactor Experiment

BWR Boiling Water Reactor 

Ci curie

CET Critical Experiment Tank

CFA Central Facilities Area

CFRMF Coupled Fast Reactivity Measurement Facility

CRCE Cavity Reactor Critical Experiment

Cs cesium

D2O deuterium oxide (“heavy” water)

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

ER Evaluation Report

ESF engineered safety feature
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

5 of 44

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.

ETR Engineering Test Reactor

ETRC Engineering Test Reactor Critical Facility

F Fahrenheit

FFTF Fast Flux Test Facility

FRAN Nuclear Effects Reactor 

ft feet, foot

gpm gallons per minute

GCRE Gas Cooled Reactor Experiment

GW gigawatt

GWth gigawatt thermal

H2O (“light”) water

HOTCE Hot Critical Experiment

Hr hour

HTGR high-temperature gas-cooled reactor

HTRE Heat Transfer Reactor Experiment

Iodine iodine

IET Initial Engine Test

IETF Initial Engine Test Facility

in inch

INL Idaho National Laboratory

kg kilogram

Kr krypton

kWe kilowatt electric

kWth kilowatt thermal

LOCA loss-of-coolant accident

LOFT Loss of Fluid Test Facility

LPTF Low Power Test Facility

LWR light water reactor

MAP mixed activation product

MFP mixed fission product

ML-1 Mobile Low-Power Reactor-1
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

6 of 44

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.

mrad millirad

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

N2 nitrogen

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

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
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

7 of 44

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.

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
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

8 of 44

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.

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), regarding 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 selected 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 several reactors in 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 the range of 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. SC&A distributed that report (Revision 0) on March 2, 2016 

(SC&A 2016).

In prioritizing reactors for further investigation, Revision 0 focused on the degree to which the 

abundance of fission and activation products and actinides relative to the abundance of Cs-137 

 
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

9 of 44

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.

and Sr-90 bear any resemblance to the mix of radionuclides in OTIB-0054. Revision 1 of the 

report addresses some comments that were received. Specifically, in addition to OTIB-0054 

applicability, SC&A was asked to consider in Revision 1 four factors that reflect the scope of the 

population potentially “at risk” of uncontrolled/unmonitored exposures: 

• Duration reactor was in operation

• Frequency/intensity of operation

• Where possible, the approximate number of workers potentially exposed during its 

operation2 

• Incidents or other factors with potential to contribute to the risk of 

unintended/unprotected exposures

These additional considerations that influence prioritization are captured in Attachment 1.

2 This information is not readily available and was not considered in this prioritization (screening process). 

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 on the 

INL site (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 framework to examine the various reactors. Attachment 1 lists all the reactors and, 

for each covered by the SEC petition, 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 OTIB-0054 bounds its makeup and operating 

conditions. 

The priority rankings are divided into three categories: High, Medium, and Low.3

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

Though based 

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

analysis would involve detailed and extensive research for each reactor and actually performing 

the OTIB-0054 applicability analyses themselves, which would go counter to the limited 

objectives of this screening process. The assignment of reactors to priority ranking categories 

considers reactor design factors such as the type of fuel (e.g., solid or gaseous, uranium or 

plutonium-based), enrichment (e.g., low-enriched commercial-type fuel or fully enriched fuel), 

cladding (e.g., aluminum or steel), moderator (e.g., H2O, D2O, or Be), and coolant (e.g., H2O, N2, 

or organic liquid); operational mode (e.g., steady-state or periodic); length of operation; and 

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

outside those limits (e.g., in tests supporting power reactor safety programs). Also considered 

qualitatively is the potential for significant radiation exposure of personnel. These screening 

criteria were selected because they were judged by SC&A to be those criteria that would be best 

indicative of the degree to which the default mix of radionuclides in OTIB-0054 might result in 

 
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

10 of 44

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.

an underestimate of the internal doses to workers or simply result in unrealistic estimates of the 

internal doses to workers at INL who worked in the vicinity of these reactors or worked with 

irradiated fuel from these reactors.

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. 
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

11 of 44

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.

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.

4 The representative reactors are listed in Table 1. 

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

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 performed by NIOSH 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 by NIOSH for each of 

the cases.

 
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

12 of 44

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.

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

Case Parameters Basis

ATR 1

Specificpower = 2,379.1 MW/MTU

Irradiation time = 132.27 days

Burnup = 314,684 MWd/MTU

Maximumburnup atnominal power.

ATR 2

Specificpower=8,651.2 MW/MTU

Irradiation time = 36.4 days

Burnup = 314,904 MWd/MTU

Maximumburnup atmaximum assembly

power.

ATR 3

Specificpower=2,379.1 MW/MTU

Irradiation time = 56 days 

Burnup =133,230 MWd/MTU

Nominal burnup atnominalpower.

FFTF 1

Specificpower=163.8 MW/MTHM

Irradiation time = 929.4 days 

Burnup =152,230 MWd/MTHM

Maximumburnup atnominal power.

FFTF 2

Specificpower=163.8 MW/MTHM

Irradiation time = 488.3 days 

Burnup =79,984 MWd/MTHM

Nominal burnup atnominalpower.

N Reactor 1

Specificpower=10.4 MW/MTU

Irradiation time = 114.2 days 

Burnup = 1,188 MWd/MTU

Production of weapons-grade plutonium

(nominal 6% Pu-240 content) atnominal

power.

N Reactor 2

Specificpower=10.4 MW/MTU

Irradiation time = 285.6 days 

Burnup = 2,970 MWd/MTU

Production of fuel-gradeplutonium (nominal

12% Pu-240 content) at nominal power.

TRIGA 1

Specificpower=15.57 MWd/MTU

Irradiation time = 730.1 days 

Burnup = 11,368 MWd/MTU

Maximumburnup atnominal power.

TRIGA 2

Specificpower=15.57 MW/MTU

Irradiation time = 115.2 days 

Burnup = 1994 MWd/MTU

Nominal burnup atnominalpower.

Source: Reproduced from ORAUT 2015, Table 5-2.
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

13 of 44

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.

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. As discussed in Section 1, the screening in this report 

examines whether it seems likely that OTIB-0054 adequately envelopes the INL reactors listed 

in Attachment 1 and prioritizes for further investigation those that might not be enveloped by the

OTIB-0054 methodology. Table 3 lists (for convenience in finding a particular reactor) all 52 

reactors alphabetically (which is also numerically) as shown in Stacy 2000. Priority rankings 

from Attachment 1 are also included in the last column.

Table 3. List of all INL Reactors

Reactor Name a,b Location Priority 

Ranking

1. Advanced Reactivity Measurement Facility No. 1 (ARMF-I) TRA (Bldg. TRA-660) Low

2. Advanced Reactivity Measurement Facility No. 2 (ARMFII). Renamed Coupled FastReactivity Measurement Facility 

(CFRMF) in 1968

TRA (Bldg. TRA-660) Low

3. Advanced Test Reactor (ATR) TRA (Bldg. TRA-670) N/Ac

4. Advanced Test Reactor Critical Facility (ATRC) TRA (Bldg. TRA-670) Low

5. Argonne Fast Source Reactor (AFSR) ANL-W N/Ae

6. Boiling Water Reactor Experiment No. 1 (BORAX-I) ANL-W N/Ae

7. Boiling Water Reactor Experiment No. 2 (BORAX-II) ANL-W N/Ae

8. Boiling Water Reactor Experiment No. 3 (BORAX-III) ANL-W N/Ae

9. Boiling Water Reactor Experiment No. 4 (BORAX-IV) ANL-W N/Ae

10. Boiling Water Reactor Experiment No. 5 (BORAX-V) ANL-W N/Ae

11. Cavity Reactor Critical Experiment (CRCE) TAN - WRRTF – LPTF Medium

12. Coupled Fast Reactivity Measurement Facility (CFRMF). 

Formerly named Advanced Reactivity Measurement Facility 

No. 2 (ARMF-II)

TRA (Bldg. TRA-660) Low

13. Critical Experiment Tank (CET) TAN - WRRTF – LPTF Low

14. Engineering Test Reactor (ETR) TRA (Bldg. TRA-642) N/Ac

15. Engineering Test Reactor Critical Facility (ETRC) TRA (Bldg. TRA-654) Low

16. Experimental Beryllium Oxide Reactor (EBOR) TAN - WRRTF – LPTF (Bldg. TAN-646) N/Af

17. Experimental Breeder Reactor No. I (EBR-I) ANL-W N/Ae

18. Experimental Breeder Reactor No. II (EBR-II) ANL-W N/Ae

19. Experimental Organic Cooled Reactor (EOCR) CFA (vicinity) N/Af

20. Fast Spectrum Refractory Metals Reactor (710) TAN - WRRTF – LPTF Low

21. Gas Cooled Reactor Experiment (GCRE) ARA-III Low

22. Heat Transfer Experiment No. 1 (HTRE-1) TAN - IETF N/Ad

23. Heat Transfer Experiment No. 2 (HTRE-2) TAN - IETF N/Ad

24. Heat Transfer Experiment No. 3 (HTRE-3) TAN - IETF N/Ad

25. High Temperature Marine Propulsion Reactor (630-A) TAN - WRRTF – LPTF Low

26. Hot Critical Experiment (HOTCE) TAN - WRRTF – LPTF Low

27. Large Ship Reactor A (A1W-A) NRF N/Ag
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

14 of 44

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 a,b Location Priority 

Ranking

28. Large Ship Reactor B A1W-B NRF N/Ag

29. Loss of Fluid Test Facility (LOFT) TAN (Bldg. TAN-650) High

30. Materials Test Reactor (MTR) TRA (Bldg. TRA-603) N/Ac

31. Mobile Low-Power Reactor No. 1 (ML-1) ARA-IV Medium

32. Natural Circulation Reactor (S5G) NRF N/Ag

33. Neutron Radiography Facility (NRAD) ANL-W N/Ae

34. Nuclear Effects Reactor (FRAN) ARA-IV Low

35. Organic Moderated Reactor Experiment (OMRE) East of CFA, between Waste Area 

Groups 4 and 5

High

36. Power Burst Facility (PBF) Near the SPERT-I site High

37. Reactivity Measurement Facility (RMF) TRA (Bldg. TRA-603) Low

38. Shield Test Pool Facility (STPF - SUSIE) TAN - WRRTF – LPTF (Bldg. TAN-646) Low

39. Special Power Excursion Reactor Test No. I (SPERT-I) Separate complex east of CFA High

40. Special Power Excursion Reactor Test No. II (SPERT-II) Separate complex east of CFA High

41. Special Power Excursion Reactor Test No. III (SPERT-III) Separate complex east of CFA High

42. Special Power Excursion Reactor Test No. IV (SPERT-IV) Separate complex east of CFA High

43. Spherical Cavity Reactor Critical Experiment (SCRCE) TAN – WRRTF - LPTF Medium

44. Stationary Low-Power Reactor (Earlier name - Argonne 

Low Power Reactor) (SL-1, ALPR) 

ARA-II Low

45. Submarine Thermal Reactor (S1W, STR). Also known as 

the Submarine Prototype Reactor 

NRF N/Ag

46. Systems for Nuclear Auxiliary Power (SNAP) 10A 

Transient No. 1 (SNAPTRAN–1)

TAN - IETF Medium

47. Systems for Nuclear Auxiliary Power (SNAP) 10A 

Transient No. 3 (SNAPTRAN–3)

TAN - IETF Medium

48. Systems for Nuclear Auxiliary Power (SNAP) 10A 

Transient No. 2 (SNAPTRAN–2)

TAN – IETF Medium

49. Thermal Reactor Idaho Test Station (THRITS) TAN – WRRTF - LPTF Low

50. Transient Reactor Test Facility (TREAT) ANL-W N/Ae

51. Zero Power Physics Reactor (Earlier name - Zero Power 

Plutonium Reactor) ZPPR 

ANL-W N/Ae

52. Zero Power Reactor No. 3 (ZPR-III) ANL-W N/Ae

a The list of 52 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; LPTF = Low Power Test 

Facility; NRF = Naval Reactor Facility; TAN = Test Area North; TRA = Test Reactor Area; WRRTF = Water Reactor 

Research Test Facility

c Already evaluated in SC&A 2015a. d Already evaluated in SC&A 2015b. e ANL-W is not included in SEC-00219 definition. 

f Never operated. g NRF is not in the EEOICPA program. 

The first step in the screening process was to identify which of the 52 reactors listed in Table 3 

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 4 summarizes the results
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

15 of 44

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.

of the winnowing process that includes for further consideration 28 reactors and excludes 24 

reactors.

Table 4. 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, as discussed in Section 1. The 24 reactors that are excluded from the 

prioritization process are included in the last group in Attachment 1 for completeness, but with 

only brief summary descriptions. 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 one-of-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 in a very short time interval, before they shut 

themselves down (due to strongly negative reactivity properties, such as negative void, 

temperature, or expansion coefficients), or in intermittent mode, in which a series of experiments 

were run, then the reactor shut down and possibly modified until the next series of experiments.

Fuel burnups were frequently considerably lower than for the representative reactors of OTIB0054, 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 for 

determining which reactors 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 5, which shows that 7 reactors 

were put in the High category, 6 in the Medium category, and 15 in the Low category. 

Table 5. Priority Class Categorization 

Priority Class Reactor Number from Attachment 1

High (7) 29, 35, 36, 39, 40, 41, 42

Medium (6) 11, 31, 43, 46, 47, 48

Low (15) 1, 2, 4, 12, 13, 15, 20, 21, 25, 26, 34, 37, 38, 44, 49

Total Included Reactors 28
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

16 of 44

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.

The rationale behind the categorizations shown in Table 5 appear in Attachment 1. SC&A 

recommends that the seven reactors in the High priority class be investigated first, with the 

results presented in a single report. For convenience and efficiency, the four SPERT reactors 

(Nos. 39–42) would be grouped together. Also in the High priority class are the LOFT (No. 29), 

OMRE (No. 35), and PBF (No. 36) reactors. 
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

17 of 44

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.

4.0 REFERENCES

AEC 1958. Proceedings of the SRE-OMRE Forum, Los Angeles, CA, February 12 and 13, 1958, 

TID-7553, U.S. Atomic Energy Commission. May 1958.

Aerojet-General 1960. The ML-1 Design Report, IDO-28550. May 16, 1960. Available at 

http://digital.library.unt.edu/ark:/67531/metadc100219/m2/1/high_res_d/metadc100219.pdf

Aerojet-General 1963. Army Gas-Cooled Reactor Systems Program: Final Summary Report of 

the Gas-Cooled Reactor Experiment-I, IDO-28598. October 1963. Available at 

http://digital.library.unt.edu/ark:/67531/metadc100222/m2/1/high_res_d/metadc100222.pdf

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. 

[SRDB Ref. ID 23569]

INL 2009. Power Burst Facility: U(18)O2CaO-ZrO2 Fuel Rods in Water, INL/EXT-09-15446, 

Idaho National Laboratory, Idaho Falls, ID. September 2009. [SRDB Ref. ID 136135]

Lofthouse and Kunze 1971. Lofthouse, J. H., and J. F. Kunze, Spherical Gas Core Reactor 

Critical Experiment, IN-1443, NASA CR-72781, Idaho Nuclear Corporation for National 

Aeronautics and Space Administration. February 1971. [SRDB Ref. ID 135872] 

Modro et al. 1989. Modro, S. M., S. N. Aksan, V. T. Berta, and A. B. Wahba, Review of LOFT 

Large Break Experiments: OECD LOFT Project, NUREG/IA-0028, U.S. Nuclear Regulatory 

Commission. October 1989. Available at http://www.nrc.gov/reading-rm/doccollections/nuregs/agreement/ia0028/#pub-info 

Montgomery et al. 1957. Montgomery, C. R., J. A. Norberg, and T. R. Wilson, Summary of the 

SPERT-I, -II, and -III Reactor Facilities, AEC Research and Development Report IDO-16418, 

U.S. Atomic Energy Commission. November 1, 1957. [SRDB Ref. ID 136833] 

Nace et al. 1972. Nace, R. L., M. Deutsch, and P. T. Voegeli, Physical Environment of the 

National Reactor Testing Station, Idaho – A Summary: Geology, Hydrology, and Waste 

Management at the National Reactor Testing Station, Idaho, Geological Survey Professional 

Paper 725-A, U.S> Department of the Interior, Geological Survey. 1972. [SRDB Ref. ID 

138052] 

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. 
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

18 of 44

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.

ORAUT 2010. Idaho National Laboratory and Argonne National Laboratory-West – Site 

Description, ORAUT-TKBS-0007-2, Revision 4, Oak Ridge Associated Universities. August 2, 

2010. 

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. 

Reeder and Berta 1979. Reeder, D. L., and V. T. Berta, The Loss-of-Fluid Test (LOFT) Facility, 

14th Annual Intersociety Energy Conversion Conference, Boston, MA. August 5, 1979. 

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. 

SC&A 2016. INL SEC-00219 Reactor Prioritization for Evaluation of ORAUT—OTIB-0054 

Applicability, SCA-TR-2016-SEC0002, Revision 0, SC&A Inc. March 2, 2016. 

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. [SRDB Ref. ID 1152, 135760]

Stillman and Mead 1965. Stillman, D. B., and S. W. Mead, The FRAN Prompt Burst Reactor, 

Lawrence Radiation Laboratory, University of California, Livermore, UCRL-14413. September 

29, 1965. 
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

19 of 44

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.

ATTACHMENT 1. INL REACTOR PRIORITIZATION WITH RESPECT TO ORAUT-OTIB-0054 

APPLICABILITY INVESTIGATION

Table A.1. High Priority Ranking

Reactor Name & 

Location a,b

Operation

Dates

Summary Description c

Comments/Priority Ranking

29. Loss of Fluid Test

Facility (LOFT) 

TAN (Bldg. TAN-650)

1973–1985 The LOFT series of 38 nuclear power experiments (the first five 

experiments were non-nuclear, thermal-hydraulic 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) up to a worst-case, doubleended break in the primary coolant system. The reactor had a 

maximum power of 50 MWth, and the associated components and 

systems were built as a volumetrically scaled model of a commercial, 

four-coolant-loop pressurized water reactor, including its engineered 

safety features (ESFs).

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 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 97-foot high 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 text in this cell continues on the next page.]

LOFT modeled, in 38 tests over a 12-year 

period, a commercial LWR, but often 

operated beyond customary limits to 

explore coupled nuclear and thermal 

hydraulic behavior in design-basis or 

beyond-design-basis accident scenarios. 

In addition, the last test deliberately 

melted 100 fuel rods, resulting in release 

of radioactivity in the containment 

building. Given the facility’s size, long 

operating history, and potential to have 

exposed a significant number of 

personnel, a more detailed examination is 

warranted. 

Priority Ranking: High
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

20 of 44

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 & 

Location a,b

Operation

Dates

Summary Description c

Comments/Priority Ranking

[Row continued from 

previous page; no additional 

text in this cell.]

[Row 

continued 

from 

previous 

page; no 

additional 

text in this 

cell.]

LOFT included six large-break LOCA experiments of increasing 

severity to provide data on a range of accidents with different initial 

and boundary conditions. The 1.68-m-long core (approximately half 

the length of a commercial core) was arranged in four triangular and 

five square fuel bundles, with the same physical, chemical, and 

metallurgical properties as commercial fuel. The primary coolant 

system consisted of an operating loop (steam generator, two primary 

coolant pumps, pressurizer, and piping) that simulated three unbroken 

loops, and a single “broken” loop that simulated a LOCA. 

The last test, in 1985 (the LP-FP-2 experiment), 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. 

As stated in ORAUT 2010 (Section 2.2.4), there was a potential for 

both internal and external exposure associated with LOFT:

Internal exposure was possible from airborne fission product 

activity in the containment soon after shutdown. Entries were 

monitored with a continuous air monitor, and respiratory 

protection was worn as required. On July 9, 1985, after 

completion of the LP-FP-2 test, leakage was discovered from 

the fission product monitoring system and the primary coolant 

system, which allowed fission products to enter the reactor 

building. Over the following 2-month period, 8,780 Ci of noble 

gas (88Kr) and 0.09 Ci of iodine (131I) were released to the 

environment... 

[The text in this cell continues on the next page.]

[Row continued from previous page; no 

additional text in this cell.]
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

21 of 44

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 & 

Location a,b

Operation

Dates

Summary Description c

Comments/Priority Ranking

[Row continued from 

previous page; no additional 

text in this cell.]

[Row 

continued 

from 

previous 

page; no 

additional 

text in this 

cell.]

External exposure occurred to personnel who worked inside 

the containment vessel or on the reactor’s primary system or 

the sample systems. During initial entry after a test, the fields 

in containment were ≥100 mrad/hr beta-gamma. The shortlived fission products would decay rapidly and reduce the 

general fields to ≤10 mrad/hr beta-gamma. 

References

Modro et al. 1989

Reeder and Berta 1979

[Row continued from previous page; no 

additional text in this cell.]

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 (AEC) 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 rather than water or 

a liquid metal as both coolant and moderator. The relatively lowpower (5–10 MWth), critical reactor tested various types and 

configurations of highly enriched uranium dioxide (UO2) fuel 

elements and gathered performance data on the coolant as well as 

nuclear data. The waxy coolant was thought to have some advantages 

over “conventional” coolants since it allows low-pressure operation, 

solidifies at low temperatures, and does not corrode metals. 

The reactor achieved initial criticality in September 1957 and 

operated almost continuously until April 1963. The Experimental 

Organic Cooled Reactor (EOCR) was designed to follow OMRE, 

which lacked test loops, but was cancelled by the AEC near the end 

of construction in 1962. 

References

Nace et al. 1972

USAEC 1958

OMRE operated at relatively low power 

levels, but did so nearly continuously for 

several years. In addition, it is doubtful 

whether OMRE would be adequately 

enveloped by any of the OTIB-0054 

cases because it used an organic coolant 

and moderator. Hence, it should be 

investigated further. 

Priority Ranking: High
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

22 of 44

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 & 

Location a,b

Operation

Dates

Summary Description c

Comments/Priority Ranking

36. Power Burst Facility 

(PBF) 

At 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 and was built on the site of 

SPERT-I. Fuel and cladding combinations were varied and tested to 

failure. Transient testing, including LOCA scenarios that modeled 

design-basis and severe accident conditions at a commercial nuclear 

power plant, led to fuel and cladding damage accompanied by the 

subsequent evolution of hydrogen and the release of fission products 

to the reactor containment. Simulated LOCAs and other severe 

accident tests were performed in an experimental loop within the 

reactor core. Test data were used by the NRC to develop and test 

reactor transient codes. 

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 MWth for a short period of time 

before initiating a very short super-critical power burst. 

The reactor was water-cooled and reflected, and used uranium-oxidefuel 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. At least 

one experimental series had an enrichment of 18.31%. Some of the 

experiments used previously irradiated fuel (e.g., 38,000 MWd/MTU) 

to better simulate accident conditions. 

[The text in this cell continues on the next page.]

The PBF operated for over 10 years,

tested a variety of new and previously 

irradiated fuel often to fuel failure, and, 

for at least part of the time, operated with 

fuel enrichments greater than used in a 

commercial nuclear power plant, thus 

putting it outside the range of the nine 

representative cases of OTIB-0054. 

Priority Ranking: High
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

23 of 44

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 & 

Location a,b

Operation

Dates

Summary Description c

Comments/Priority Ranking

[Row continued from 

previous page; no additional 

text in this cell.]

[Row 

continued t

from 

previous 

page; no 

additional 

text in this 

cell.]

ORAUT 2010 (Section 2.9) recognizes the potential of exposure from 

he PBF:

Internal exposure was possible based on releases from 

operations at the SPERT reactors and PBF. Maintenance 

activities and other work with radioactive material (especially 

from PBF loop experiment) resulted in airborne MFPs and 

MAPs, which made internal exposure possible; 137Cs was the 

primary radionuclide.

External exposure resulted from experiment changes and 

maintenance activities. Cesium-137 was a primary nuclide for 

direct radiation exposure from fission products in the 

transport lines and in the loops at the PBF during severe fuel 

damage tests when radiation levels were measured up to 

50 rad/hr. Other radiological work activities resulted in much 

lower exposure rates from the MFPs and MAPs.

References

INL 2009

Nace et al. 1972

[Row continued from previous page; no 

additional text in this cell.]

39. Special Power

Excursion Reactor Test No. 

I (SPERT-I) 

Separate complex east of 

CFA

1955–1964 The four reactors in the SPERT program were deliberately subjected 

to large, rapid reactivity excursions in order to gather data on coupled 

neutronic and thermal-hydraulic responses as part of an AEC 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 flows, 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. [The text in this cell continues on next page.]

The SPERT series modeled a commercial 

reactor but often operated beyond usual 

limits to explore coupled nuclear and 

thermal-hydraulic behavior in designbasis or beyond-design-basis accident 

scenarios (which included very high 

power levels during excursions). They

also varied fuel and reactor component 

materials and designs. 

[The text in this cell continues on the next 

page.]
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

24 of 44

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 & 

Location a,b

Operation

Dates

Summary Description c

Comments/Priority Ranking

[Row continued from 

previous page; no additional 

text in this cell.]

[Row 

continued 

from 

previous 

page; no 

additional 

text in this 

cell.]

SPERT-I, the first reactor of the program, was an open-tank, light 

water moderated and reflected reactor, with the uranium fuel enriched 

up to 93.5% (much higher than in a commercial reactor). 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 inches

long) in a 4-foot diameter and 14-foot deep carbon steel tank, clad

with aluminum. 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 deliberate 2,300 MWth excursion on November 

5, 1962, resulted in an explosion that completely melted 

approximately 8% and partially melted about 35% of the plate-type 

core and even distorted the reactor vessel. 

Subsequently, SPERT-I was rebuilt, and low-enriched fuel rods 

replaced the high-enriched fuel plates. A deliberate 17,400 MWth

excursion on November 12, 1963 (with 4% uranium oxide fuel rods), 

and a deliberate 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 them. 

SPERT-I underwent about 1,300 kinetic tests with six different cores 

in its 10-year lifetime. 

References

Montgomery et al. 1957

Nace et al. 1972

They were major experiments that had a 

variety of fuel types and enrichments, 

lasted for about 15 years, and involved 

many workers. Hence, they should be 

evaluated further. 

Priority Ranking: High
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

25 of 44

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 & 

Location a,b

Operation

Dates

Summary Description c

Comments/Priority Ranking

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 in a reactor that modelled a commercial 

reactor. Several different types of fuel assemblies were used, both 

light and heavy water were tested as moderators and coolants, and 

different reflectors were also used. Unlike SPERT-I, SPERT-II was 

placed in a closed 24.5 ft high × 10 ft inside diameter pressure vessel

and the coolant system was pressurized.

The active length of the flat-plate fuel assemblies were about 24 

inches. Each fuel plate contained a 93.5% enriched uraniumaluminum alloy and was clad in aluminum. Since the reactor operated 

in the transient, burst mode, with power excursions up to 20 MW-sec, 

total burnup was small. 

References

Montgomery et al. 1957

Nace et al. 1972

The SPERT series modeled a commercial 

reactor but often operated beyond usual 

limits to explore coupled nuclear and 

thermal-hydraulic behavior in designbasis or beyond-design-basis accident 

scenarios (which included very high 

power levels during excursions). They

also varied fuel and reactor component 

materials and designs. They were major 

experiments that had a variety of fuel 

types and enrichments, lasted for about 

15 years, and involved many workers. 

Hence, they should be evaluated further. 

Priority Ranking: High

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, such as temperature, pressure, and coolant flow. 

Although SPERT-III was planned to be the third in the series, it 

actually was the second built and operated. 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 nuclear 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, which had a 

maximum coolant flow of 20,000 gpm. 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, moderator, and reflector. 

References

Montgomery et al. 1957, Nace et al. 1972

The SPERT series modeled a commercial 

reactor but often operated beyond usual 

limits to explore coupled nuclear and 

thermal-hydraulic behavior in designbasis or beyond-design-basis accident 

scenarios (which included very high 

power levels during excursions). They

also varied fuel and reactor component 

materials and designs. They were major 

experiments that had a variety of fuel 

types and enrichments, lasted for about 

15 years, and involved many workers. 

Hence, they should be evaluated further. 

Priority Ranking: High
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

26 of 44

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 & 

Location a,b

Operation

Dates

Summary Description c

Comments/Priority Ranking

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 open pool design allowed direct observation 

of reactor performance under different hydrodynamic conditions. The 

fuel consisted of a 93.5% enriched uranium-aluminum matrix in a 

plate-type configuration. 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. 

References

Nace et al. 1972

The SPERT series modeled a commercial 

reactor but often operated beyond usual 

limits to explore coupled nuclear and 

thermal-hydraulic behavior in designbasis or beyond-design-basis accident 

scenarios (which included very high 

power levels during excursions). They

also varied fuel and reactor component 

materials and designs. They were major 

experiments that had a variety of fuel 

types and enrichments, lasted for about 

15 years, and involved many workers. 

Hence, they should be evaluated further. 

Priority Ranking: High

a The list and numbering scheme of the 52 INL reactors were taken from Stacy 2000.

b Location acronyms (current names are used in most cases): CFA = Central Facilities Area; TAN = Test Area North. c The primary sources of information for the summary descriptions are Stacy 2000, ORAUT 2010, and NIOSH 2015b. Other sources specific to particular 

reactors are listed at the end of each source description.
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

27 of 44

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.

Table A.2. Medium Priority Ranking

Reactor Name & 

Location a,b

Operation 

Dates

Summary Description c Comments/Priority Ranking

11. Cavity Reactor Critical 

Experiment (CRCE) 

TAN - WRRTF – LPTF

1967–early 

1970s

CRCE was located in a shielded test cell of the Low Power Test 

Facility (LPTF) in the Water Reactor Research Test Facility 

(WRRTF) area of Test Area North (TAN). It was a National 

Aeronautics and Space Administration (NASA)-sponsored 

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 flowing in the annular 

region between the core and the moderator. The hydrogen could be 

heated to a very high temperature, producing a high specific impulse

(compared to a conventional chemical rocket engine) that might be 

used, for example, on a voyage to Mars. The CRCE core material was 

uranium hexafluoride (UF6), operating at the relatively low 

temperature of about 200° F. Typical of the reactors in the LPTF, 

CRCE operated at a maximum power of less than100 Wth. CRCE was 

followed at INL by the SCRCE gaseous core reactor. 

References: 

Nace et al. 1972

The reactor operated at low power and 

low temperature and only intermittently 

during experimental runs, potentially 

exposing relatively few people, and the 

UF6 gaseous core did not undergo the 

deleterious effects of “burnup.” However, 

the reactor lies totally outside the 

envelope of any of the OTIB-0054 

representative reactors and ought to be 

investigated further along with the other 

gaseous core reactors. 

Priority Ranking: Medium
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

28 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c 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 high-temperature, 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, generated 330 kWe. 

The ML-1, built by Aerojet General, went critical on March 31, 1961, 

and operated (from a remote control cab 500 feet away – outside the 

exclusion zone) 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 × 22" H cylindrical core

arrangement, moderated by light water and cooled by nitrogen. The 

reactor and lead shield/fast neutron reflector were suspended in a 

nine-foot diameter tank of borated water. 

The reactor went critical on March 30, 1961, and operated for less 

than 1,000 hours before it was shut down in 1964. 

References

Aerojet-General 1960

The GCRE operated for less than 1,000 

hours and was relatively small. However, 

none of the representative reactor cases of 

OTIB-0054 are HTGRs, and the INL gascooled reactors ought to be investigated 

further.

Priority Ranking: Medium 
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

29 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c 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, which followed the CRCE reactor, was the last experiment 

in the NASA-sponsored program to determine the feasibility and 

explore the neutronics characteristics of a reactor with a core of 

highly-enriched (93.2%) gaseous uranium (UF6). SCRCE employed a 

deuterium oxide (D2O) reflector and moderator and had a spherical 

geometry rather than the cylindrical geometry of previous 

experiments, which facilitated the development and use of 1-D 

neutronics codes.Heat from the reactor would be used to heat a gas 

that would propel a space vehicle. The low-power (up to 500 Wth for 

two hours) experiment was located in a shielded cell of the LPTF. 

Three configurations were tested: (1) clean, spherical geometry; (2) 

Styrofoam added to cavity propellant region; and (3) same as #2 with 

stainless steel added to the cavity wall as a thermal neutron absorber.

References

Lofthouse and Kunze 1971.

SCRCE was small and operated for only 

a short period of time. However, the 

reactor, with a gaseous core of UF6 and a 

D2O reflector, lies totally outside the 

envelope of any of the OTIB-0054 

representative reactors and, like the other 

gaseous core reactors, ought to be 

investigated further. 

Priority Ranking: Medium
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

30 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c Comments/Priority Ranking

46. Systems for Nuclear 

Auxiliary Power (SNAP) 

10A Transient No. 1 

(SNAPTRAN–1)

TAN – IETF

Early 1960s The AEC’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 auxiliary power applications by testing the SNAP 10A/2 

reactor under extreme conditions. The reactor was a compact 

cylinder, with a core about 9 inches in diameter and 12 inches long, 

with 37 1.212-inch diameter stainless steel-clad fuel rods. The reactor

was fully enriched (93%), zirconium hydride-uranium fueled with 

4.75 kg of U-235, liquid sodium-potassium (NaK) cooled, and 

beryllium reflected. The SNAPTRAN reactors rode on railcars to and 

from the TAN Hot Shop where they were stored, inspected, and 

worked on when not in operation. 

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 of experiments 

often operated well beyond design basis 

to explore coupled nuclear and thermalhydraulic behavior. It also utilized fuel 

and reactor component materials and 

designs that were not similar to the 

reactors of OTIB-0054 and, therefore, 

should be investigated further. 

Priority Ranking: Medium

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 nuclear-powered 

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 reached 30 GW for 

1.5 milliseconds before it destroyed itself.

The SNAPTRAN series of experiments 

often operated well beyond design basis 

to explore coupled nuclear and thermalhydraulic behavior. It also utilized fuel 

and reactor component materials and 

designs that were not similar to the 

reactors of OTIB-0054 and, therefore, 

should be investigated further. 

Priority Ranking: Medium
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

31 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c Comments/Priority Ranking

48. Systems for Nuclear 

Auxiliary Power (SNAP)

10A Transient No. 2 

(SNAPTRAN–2)

TAN – IETF

1965–1966 SNAPTRAN-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 

and released 54 MW-s of energy before it destroyed itself. 

The SNAPTRAN series of experiments 

often operated well beyond design basis 

to explore coupled nuclear and thermalhydraulic behavior. It also utilized fuel 

and reactor component materials and 

designs that were not similar to the 

reactors of OTIB-0054 and, therefore, 

should be investigated further. 

Priority Ranking: Medium

a The list and numbering scheme of the 52 INL reactors were taken from Stacy 2000.

b Location acronyms (current names are used in most cases): ARA = Auxiliary Reactor Area; IETF = Initial Test Engine Facility; LPTF = Low Power Test 

Facility; TAN = Test Area North; WRRTF = Water Reactor Research Test Facility. c The primary sources of information for the summary descriptions are Stacy 2000, ORAUT 2010, and NIOSH 2015b. Other sources specific to particular 

reactors are listed at the end of each source description.

Table A.3. Low Priority Ranking

Reactor Name & 

Location a,b

Operation 

Dates

Summary Description c

Comments/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 in a 

specially-constructed building 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) and was itself replaced by an improved version, ARMF-II, 

located in the same tank.

References

Nace et al. 1972

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
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

32 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/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 and to perform neutron radiography. 

The ARMF-II 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).

References

Nace et al.1972

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 enveloped by 

either an ATR case or a Fast Reactor Test 

Facility (FFTF) (fast reactor) case. 

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 ETR. The full-sized reactor tests fuel and 

experiment configurations destined for the 250 MWth ATR, but at 

low power levels, thereby optimizing ATR experimental time. The 

ATRC is designed to operate at 5 kWth steady state but is usually 

operated at about 500 Wth. 

References

Nace et al.1972

The reactor uses 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
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

33 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

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 high-energy neutron flux that was useful in comparing 

calculated and observed results. This tool provided physics 

information about the behavior of fast (i.e., unmoderated fission) 

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, but modeling it 

with one of the ATR cases might be 

claimant favorable.

Priority Ranking: Low

13. Critical Experiment 

Tank (CET)

TAN – WRRTF – LPTF

1958–1960 Three low-power reactors (less than 100 Wth) 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, beryllium-reflected CET was a low-power 

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. However, the reactor 

was low power and operated for only a 

short period of time.

Priority Ranking: Low
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

34 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

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, thereby 

making maximum use of the ETR time for experiments.

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

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 (less than 100 Wth), split-table, critical facility 

collected data for a proposed fast-spectrum, compact, refractorymetal 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: Low
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

35 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

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, direct 

and closed cycle mobile nuclear power plant. As such, it was part of 

the Army Gas-Cooled Reactor Systems Program (AGCRSP), 

provided nuclear and other data for the ML-1 reactor, and served as a 

training platform for Army and civilian personnel. The GCRE was 

built by Aerojet General Nucleonics, went critical on February 23, 

1960, and placed on standby on April 6, 1961, after accomplishing 

proof-of-principle. 

The GCRE generated 2.2 MWth from either plate-type or pin-type, 

fully enriched uranium fuel arranged in a heterogeneous pattern; the 

former fuel type was housed in an aluminum calandria and the latter 

in a stainless steel calandria holding moderator water. The coolant 

was nitrogen with 0.5% oxygen added. The GCRE accumulated 

2,989 MW-hr with approximately 1,000 hr at full power. 

References

Aerojet-General 1963

Nace et al. 1972

The GCRE was gas cooled but operated 

for only a bit over one year before it was 

taken out of service. 

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 air-cooled, water-moderated system for nuclearpowered merchant ships. 

The 630-A reactor operated at a low 

power level for only a short period of 

time. 

Priority Ranking: Low
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

36 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

26. Hot Critical 

Experiment (HOTCE)

TAN – WRRTF – LPTF

1958–1961 Three low-power (less than 100 Wth) 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

experiment containing 50.4 kg of 93.2%-enriched UO2 in 1/8 inchthick stainless steel wire. It had a hydrided zirconium reactor and a 

beryllium reflector. The experiments were intended to measure 

temperature coefficients in solid moderated reactors. HOTCE 

typically operated at a power level of 1 Wth for up to one to three 

hours, but could operate at 100 Wth for short periods. 

It is not clear without a more in-depth 

investigation whether the CET would be 

adequately enveloped by any of the 

OTIB-0054 cases. However, the reactor 

was low power and operated for only a 

short period of time.

Priority Ranking: Low

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, which consisted of stacked, cylindrical fuel rings with an 

internal void for sample irradiation, 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. Power burst were self-limiting and produced very 

large, short-duration fast neutron bursts. About 180 bursts were 

produced over the lifetime of the facility, which was located in a 

specially-designed underground bunker. 

FRAN was designed and built at the Lawrence Livermore National 

Laboratory (current name) and brought back there in 1970. 

References

Stillman and Mead 1965

FRAN appears to lie outside the reactors 

of OTIB-0054, but it was quite small and 

operated for only a short period of time. 

Note: FRAN appears in Stacy 2000 and 

other literature that was examined, but 

not in the INL site profile. 

Priority Ranking: Low
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

37 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

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 such as 

reactivity changes in materials irradiated in the MTR or ETR. It was 

replaced by the ARMF-I, which became operational in 1964. 

References

Nace et al. 1972

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

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 STPF (SUSIE was the name of the reactor for the first 

experimental program) 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, 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 containing 4 kg of U-235. 

After the ANP program was discontinued in 1961, SUSIE continued 

in use by other programs and, after modification, became known as 

EBOR, which was never fueled. 

It is not clear without a more in-depth 

investigation whether SUSIE would be 

adequately enveloped by any of the 

OTIB-0054 cases, but it was low power 

and operated for only a short period of 

time. 

Priority Ranking: Low
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

38 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

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 was fabricated out of steel plate and 

was not designed as a containment structure.

The ATR cases might adequately 

envelope the SL-1 (other than its 

meltdown). The accident and its 

aftermath have been extensively 

discussed in many sources. 

Priority Ranking: Low

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.

References

Nace et al. 1972

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 and the reactor 

operated for only a short period of time,

though. 

Priority Ranking: Low

a The list and numbering scheme of the 52 INL reactors were taken from Stacy 2000.

b Location acronyms (current names are used in most cases): ARA = Auxiliary Reactor Area; LPTF = Low Power Test Facility; TAN = Test Area North; 

TRA = Test Reactor Area; WRRTF = Water Reactor Research Test Facility. c The primary sources of information for the summary descriptions are Stacy 2000, ORAUT 2010, and NIOSH 2015b. Other sources specific to particular 

reactors are listed at the end of each source description.
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

39 of 44

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.

Table A.4. Excluded from Prioritization Process (ANL-W, NRF), Already Evaluated, or Never Operated

Reactor Name & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

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 unusualcurved 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 2015a.

Priority Ranking: N/A

5. Argonne Fast Source 

Reactor (AFSR) 

ANL-W

— — ANL-W is not included in the SEC-0219 

definition. 

Priority Ranking: N/A

6. Boiling Water Reactor 

Experiment No. 1 

(BORAX-I)

ANL-W

— — 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 

— — ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

40 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

8. Boiling Water Reactor 

Experiment No. 3 

(BORAX-III) 

ANL-W

— — 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

— — 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

— — ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

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 2015a.

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
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

41 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

17. Experimental Breeder 

Reactor No. I (EBR-I)

ANL-W

— — ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

18. Experimental Breeder 

Reactor No. II (EBR-II) 

ANL-W

— — 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

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 nonnuclear, 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-chromium-oxide 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.

Priority Ranking: N/A
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

42 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

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.

Priority Ranking: N/A

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.

Priority Ranking: N/A

27. Large Ship Reactor A 

(A1W-A)

NRF

— — NRF is not included in the EEOICPA 

program.

Priority Ranking: N/A

28. Large Ship Reactor B 

A1W-B 

NRF

— — NRF is not included in the EEOICPA 

program.

Priority Ranking: N/A
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

43 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

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 2015a.

Priority Ranking: N/A

32. Natural Circulation 

Reactor (S5G) 

NRF

— — NRF is not included in the EEOICPA 

program.

Priority Ranking: N/A

33. Neutron Radiography 

Facility (NRAD)

ANL-W

— — ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

45. Submarine Thermal 

Reactor (S1W, STR). Also 

known as the Submarine 

Prototype Reactor 

NRF

— — NRF is not included in the EEOICPA 

program.

Priority Ranking: N/A

50. Transient Reactor Test 

Facility (TREAT)

ANL-W 

— — ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A
Effective Date:

June 10, 2016

Revision No.

1 (Draft)

Document No./Description:

SCA-TR-2016-SEC002

Page No.

44 of 44

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 & 

Location a,b

Operation 

Dates

Summary Description c

Comments/Priority Ranking

51. Zero Power Physics 

Reactor (Earlier name:

Zero Power Plutonium 

Reactor) (ZPPR)

ANL-W

— — ANL-W is not included in the SEC-0219 

definition.

Priority Ranking: N/A

52. Zero Power Reactor 

No. 3 (ZPR-III)

ANL-W

— — 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 were taken from Stacy 2000.

b Location acronyms (current names are used in most cases): ANL;-W = Argonne National Laboratory-West; CFA = Central Facilities Area; IETF = Initial Test 

Engine Facility; LPTF = Low Power Test Facility; NRF = Naval Reactor Facility; TAN = Test Area North; TRA = Test Reactor Area; WRRTF = Water Reactor 

Research Test Facility.

c The primary sources of information for the summary descriptions are Stacy 2000, ORAUT 2010, and NIOSH 2015b. Other sources specific to particular 

reactors are listed at the end of each source description.