Page 1 of 9 

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions with the 

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NIOSH Proposal for INL and ANL-W Reactor 

Prioritization for OTIB-0054 Evaluation 

White Paper 

____________________________________

National Institute for Occupational Safety and Health (NIOSH) 

July 28, 2016 

Mitch Findley, Bob Burns, and Brian Gleckler 

Oak Ridge Associated Universities Team 

Reviewed by Tim Taulbee and Stu Hinnefeld 

Division of Compensation Analysis and Support 
Page 2 of 9 

This is a working document prepared by NIOSH’s Division of Compensation Analysis and Support (DCAS) or its contractor for use in discussions with the 

ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or ABRWH (or their technical 

support and review contractors) positions, unless specifically marked as such. This document represents preliminary positions taken on technical issues 

prepared by NIOSH or its contractor. 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. 

Introduction 

In March 2016, SC&A released SCA-TR-2016-SEC002, INL SEC-00219 Reactor Prioritization for Evaluation of 

ORAU-OTIB-0054 Applicability, Revision 0. The report was prepared in response to a request from the Advisory 

Board on Radiation and Worker Health (ABRWH) to screen the rest of the Idaho National Laboratory (INL) 

reactors not already reviewed with respect to ORAUT-OTIB-0054 to ensure that the prescribed approaches were 

bounding for INL reactors. ORAUT-OTIB-0054 is a primary tool used for internal dose reconstruction by assigning 

fission and activation product intakes via ratios tied to an indicator radionuclide. 

In SCA-TR-2016-SEC002 Rev. 0, SC&A reviewed all 52 reactors in the history of the INL site and grouped them 

into high, medium, and low priority categories in regards to performing evaluations on whether OTIB-0054 is 

bounding for INL reactors. The assignment of a reactor to one of the priority categories considered the following 

per SC&A: 

 reactor design factors such as fuel type, enrichment, cladding, 

moderator and coolant 

 operational modes such as steady-state or periodic 

 if the reactor operated within design limits or was intentionally taken 

outside of those limits (INL was a reactor “testing station”)

From the 52 total reactors the following were removed from consideration: 

 12 Argonne National Laboratory-West (ANL-W) reactors 

 4 Naval Reactor Facility reactors (outside of EEOICPA program) 

 3 Test Area North reactors already evaluated 

 3 Test Reactor Area reactors already evaluated 

 2 reactors cancelled before operations 

The 28 INL reactors remaining were grouped into the categories listed in Table 1. 

INL Priority Category Number of Reactors

High 13

Medium 8

Low 7

 Table 1: INL reactor priority in SCA-TR-2016-SEC002 Rev 0 

For the remainder of this report only the high priority category INL reactors will be addressed as they were 

identified as the reactors of most concern in regards to OTIB-0054. NIOSH believes that the high priority 

category recognized in the SC&A report should be addressed first with the medium and low categories reserved 

for evaluation after completion of the high priority category reactors. The following reactors were identified by 

SC&A as being in the high priority category. The thirteen high priority INL reactors are presented in the table 

below: 
Page 3 of 9 

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Reactor # in SC&A Report Reactor Name

11 Cavity Reactor Critical Experiment (CRCE)

29 Loss of Fluid Test Facility (LOFT)

31 Mobile Low Power Reactor (ML-1)

35 Organic Moderated Reactor Experiment (OMRE)

36 Power Burst Facility (PBF)

39 Special Power Excursion Reactor Test I (SPERT-I)

40 Special Power Excursion Reactor Test II (SPERT-II)

41 Special Power Excursion Reactor Test IIII (SPERT-III)

42 Special Power Excursion Reactor Test IV (SPERT-IV)

43 Spherical Cavity Reactor Critical Experiment (SCRCE)

46 Systems for Nuclear Auxiliary Power 10A Transient No. 1 (SNAPTRAN-1)

47 Systems for Nuclear Auxiliary Power 10A Transient No. 2 (SNAPTRAN-2)

48 Systems for Nuclear Auxiliary Power 10A Transient No. 3 (SNAPTRAN-3)

Table 2: High priority category INL reactors in SCA-TR-2016-SEC002 Rev 0 

On June 10, 2016, revision 1 of SCA-TR-2016-SEC002 was released. The primary change in the revision was a 

change in the prioritization of INL reactors for OTIB-0054 review. The primary change was the reduction of high 

priority category reactors from 13 to 7. Below are tables which summarize the difference in the number of 

reactors categorized for ORAU-OTIB-0054 applicability. 

Priority Category Revision 0 Revision 1

High 13 7

Medium 8 6

Low 7 15

Table 3: INL reactor priority in SCA-TR-2016-SEC002 Rev 1 

Reactor # in SC&A Report Reactor Name

29 Loss of Fluid Test Facility (LOFT)

35 Organic Moderated Reactor Experiment (OMRE)

36 Power Burst Facility (PBF)

39 Special Power Excursion Reactor Test I (SPERT-I)

40 Special Power Excursion Reactor Test II (SPERT-II)

41 Special Power Excursion Reactor Test IIII (SPERT-III)

42 Special Power Excursion Reactor Test IV (SPERT-IV)

Table 4: High priority category INL reactors in SCA-TR-2016-SEC002 Rev 1 

In July 2016, SC&A released SCA-TR-2016-SEC010, Argonne National Laboratory-West SEC-00224 Reactor 

Prioritization for Evaluation of ORAU-OTIB-0054 Applicability, Revision 0. The report evaluated the 12 Argonne 

National Laboratory-West (ANL-W) reactors that were excluded from SCA-TR-2016-SEC002. The evaluation 

followed the same criteria as described earlier but also considered four factors “that reflect the scope of the 

population ‘at risk’ of uncontrolled/unmonitored exposures.” Those four factors are duration of reactor 

operation, frequency/intensity of operation, approximate number of workers potentially exposure during 

operation, and incidents/other factors with potential to contribute to the risk of unintended/unprotected 

exposures. 
Page 4 of 9 

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ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or ABRWH (or their technical 

support and review contractors) positions, unless specifically marked as such. This document represents preliminary positions taken on technical issues 

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ANL-W Priority Category Number of Reactors

High 7

Medium 1

Low 4

Table 5: ANL-W reactor priority in SCA-TR-2016-SEC010 Rev 0 

Again, focusing only on the high priority category the following seven reactors were identified as causing the 

most concern in regards to OTIB-0054. 

Reactor # in SC&A Report Reactor Name

6 Boiling Water Reactor Experiment No. I (BORAX-I)

7 Boiling Water Reactor Experiment No. II (BORAX-II)

8 Boiling Water Reactor Experiment No. III (BORAX-III)

9 Boiling Water Reactor Experiment No. IV (BORAX-IV)

10 Boiling Water Reactor Experiment No. V (BORAX-V)

17 Experimental Breeder Reactor No. I (EBR-I)

18 Experimental Breeder Reactor No. II (EBR-II)

Table 6: High priority category ANL-W reactors in SCA-TR-2016-SEC010 Rev 0 

Discussion 

The following narrative provides a review of each of the high priority category reactors for INL and ANL-W. 

NIOSH believes that combining the reactors from both sites is practical given the completion of both the INL and 

ANL-W 83.13 evaluation reports.

INL: Loss of Fluid Test Facility (LOFT) 

The Loss of Fluid Test Facility (TAN-650) operated from 1973 - 1985. It was part of an international safety 

program for commercial power reactors. From the Historical Engineering Record – Idaho National Engineering 

and Environmental Laboratory – Test Area North, HAER No. ID-33-E (SRDB 83751) page 120, “The first nuclear 

test took place on December 10, 1978.” NIOSH proposes that LOFT be removed from consideration for 

evaluation of OTIB-0054 applicability at this time because nuclear operations began after the INL SEC evaluation 

period. If the workgroup believes a review is necessary, we suggest that it be categorized along with other 

findings arising from SCA’s review of the site profile, rather than with SEC-00219 findings. 

INL: Organic Moderated Reactor Experiment (OMRE) 

The Organic Moderated Reactor Experiment operated from 1957 - 1963. It was a concept reactor to 

demonstrate the technical and economic feasibility of using a liquid hydrocarbon as both coolant and 

moderator. NIOSH agrees that OMRE should be evaluated due to its unique design and that the Piqua Nuclear 

Generating Station, a municipal power reactor that was developed based on OMRE, is already part of the Special 

Exposure Cohort. 
Page 5 of 9 

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ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or ABRWH (or their technical 

support and review contractors) positions, unless specifically marked as such. This document represents preliminary positions taken on technical issues 

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INL: Power Burst Facility (PBF) 

The Power Burst Facility operated from 1972 through 1985 and was located on the former site of SPERT-I. It was 

an extension of the reactor safety studies started with the Special Power Excursion Test reactors. As such, it was 

utilized for transient testing to produce accident conditions. Fuel failure was one outcome of the PBF testing. 

PBF utilized an intermediate enriched fuel with a unique ceramic fuel clad in 304 stainless steel. Due to the 

unique fuel, NIOSH agrees that PBF should be evaluated. 

INL: Special Power Excursion Reactor Tests 

Four SPERT reactors (I-IV) were constructed and tested between 1955 and 1970 in an effort to understand the 

mechanisms that resist power increases and terminate runaway power conditions. These reactors were 

designed and operated beyond their normal safety limits to study a wide range of variables, such as core 

configuration, coolant flow and reflection, moderation, void, and temperature coefficients. NIOSH proposes 

that a model for the most extreme experiment from all of the Special Power Excursion Reactor Tests (SPERT), in 

terms of possible departures from OTIB-0054, be used to represent the “bounding” case to cover all four SPERT 

reactors that are discussed below.

INL: Special Power Excursion Reactor Test I (SPERT-I)

Special Power Excursion Reactor Test I operated from June 1955 until the fall of 1964. It was an open-tank, light 

water-moderated and reflected reactor, originally using 93 % enriched uranium fuel. The reactor tank was four 

feet in diameter and 14 feet tall and water filled to a level two feet above the core. The SPERT-I tests were 

characterized by sudden power spikes arrested by shutdown or the self-limiting tendencies of the reactor 

itself. No provisions for heat removal or coolant circulation through the core were included. Three 

important tests for SPERT-I were: 

1. A deliberate 2,300 MWt power burst on November 5, 1962 that destroyed the 

highly-enriched uranium metal plate core and distorted the reactor vessel. 

SPERT-I was rebuilt and used low-enriched uranium rod fuel afterward. 

2. Uranium oxide fuel (low enrichment - 4%) was for a 17,400 MWt power burst on 

November 12, 1963. 

3. Uranium oxide fuel (low enrichment 4%) was for a 35,000 MWt power burst on 

April 14, 1964. 

INL: Special Power Excursion Reactor Test II (SPERT-II)

Special Power Excursion Reactor Test II operated from March 1960 until August 1964. It continued the program 

of investigating the kinetic behavior of heterogeneous, water-moderated, reactor systems begun with SPERT-I. It 

was a closed, pressurized water reactor with a reactor tank 10 feet in diameter and 24. 5 feet deep. The coolant 

flow system designed for utilized either light water or heavy water. SPERT-II utilized 93% enriched uranium 

aluminum clad fuel plates and was not designed to operate at high temperatures and pressures. 
Page 6 of 9 

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ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or ABRWH (or their technical 

support and review contractors) positions, unless specifically marked as such. This document represents preliminary positions taken on technical issues 

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INL: Special Power Excursion Reactor Test III (SPERT-III) 

Special Power Excursion Reactor Test III operated from December 1958 until June 1968. It provided the widest 

practical range of control over temperature, pressure, and coolant flow. Pressures from atmospheric to 

2,500 psi, water temperatures from 68° to 670°F, and coolant flow rates ranging from zero to 20,000 

gallons per minute with heat-removal capacities up to 60,000 kilowatts for durations of 30 minutes were 

attainable. SPERT-III utilized 4% enriched uranium stainless steel clad fuel plates. 

INL: Special Power Excursion Reactor Test IV (SPERT-IV) 

Special Power Excursion Reactor Test IV operated from July 1962 until August 1970. Unlike the other SPERT 

reactors, it was a twin-pool facility (open-tank) for detailed studies of reactor stability as affected by varying 

conditions, including forced coolant flow, variable height of water above the core, hydrostatic head, and 

other hydrodynamic effects. Most specifically, SPERT-IV made detailed studies of phenomena of instability 

first observed in the 1,300 excursions conducted at SPERT-I. SPERT-III utilized 93% enriched uranium 

aluminum clad fuel plates. 

ANL-W: Boiling Water Reactor Experiment No. I (BORAX-I)

Boiling Water Reactor Experiment No. I operated from July 1953 to July 1954 but only during the summer and 

fall months because it was not housed in a building. It was the first of a series of five separate reactors designed 

to investigate the safety of boiling water reactors. Fuel assemblies were made of curved fuel plates containing 

a U-235-aluminum-alloy core that was clad with aluminum. In July 1954, it was subjected to a final 

deliberate destructive excursion to determine the inherent safety under extreme conditions. It operated at 

1.4 MWt. NIOSH proposes that BORAX-I be removed from consideration for evaluation of OTIB-0054 

applicability, as nuclear operations of this reactor are during an existing SEC time period. NIOSH has already 

determined it is infeasible to reconstruct the fission product doses from this reactor due to limited to nonexistent bioassay. 

ANL-W: Boiling Water Reactor Experiment No. II (BORAX-II)

Boiling Water Reactor Experiment No. II operated from October 1954 to March 1955. With BORAX-II, ANL 

tested new core combinations using different enrichments of U-235 in the fuel plates. It operated at 6.4 

MWt. NIOSH proposes that BORAX-II be removed from consideration for evaluation of OTIB-0054 applicability 

as nuclear operations of this reactor are during an existing SEC time period. NIOSH has already determined it is 

infeasible to reconstruct the fission product doses from this reactor due to limited to non-existent bioassay. 

ANL-W: Boiling Water Reactor Experiment No. III (BORAX-III)

Boiling Water Reactor Experiment No. III operated from July 1955 to April 1956. In 1955, ANL replaced the 

reactor core and added a turbine-generator to BORAX-II to test power distribution, and whether turbine 

contamination would be a significant problem in a boiling water reactor. The revised reactor configuration 

was designated BORAX-III. The core consisted of 90% enriched uranium aluminum alloyed fuel plates clad in 

aluminum. It operated at 12 MWt. NIOSH proposes that BORAX-III be removed from consideration for 

evaluation of OTIB-0054 applicability as nuclear operations of this reactor are during an existing SEC time period. 

NIOSH has already determined it is infeasible to reconstruct the fission product doses from this reactor due to 

limited to non-existent bioassay. 
Page 7 of 9 

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ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or ABRWH (or their technical 

support and review contractors) positions, unless specifically marked as such. This document represents preliminary positions taken on technical issues 

prepared by NIOSH or its contractor. 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. 

ANL-W: Boiling Water Reactor Experiment No. IV (BORAX-IV)

Boiling Water Reactor Experiment No. IV operated from July 1956 to June 1958. ANL-W replaced the BORAX-III 

core with a core of a maximum seventy-two ceramic uranium-thorium oxide fuel elements to test and 

demonstrate the feasibility of stable operation with a fuel that: (1) could operate at higher temperatures 

than the uranium core; and (2) was considered less reactive with water coolant in case of cladding rupture. 

It operated at 20 MWt. NIOSH agrees that BORAX-IV should be evaluated for OTIB-0054 applicability due to the 

use of uranium-thorium oxide fuel during the operational period. 

ANL-W: Boiling Water Reactor Experiment No. V (BORAX-V)

Boiling Water Reactor Experiment No. V operated from March 1962 to August 1964. It had essentially the same 

arrangement as BORAX-IV but had a nuclear superheat system added to determine the safety aspects and 

potential steam cycle efficiency increases. It operated at 40 MWt. NIOSH proposes that BORAX-V be 

removed from consideration for evaluation of OTIB-0054 applicability since the core is the same as BORAX IV 

and the BORAX V primary function was to evaluate steam superheating.

ANL-W: Experimental Breeder Reactor No. I (EBR-I)

Experimental Breeder Reactor No. I operated from 1951 to 1963. It was an unmoderated and heterogeneous 

design with a thermal output of approximately 1 MWt at full power. To permit the efficient removal of heat 

and to minimize neutron moderation a sodium-potassium (NaK) coolant was used. There were four cores 

utilized during operation: 

1. Mark I (1951 – 1954): This core was 93% enriched unalloyed uranium in stainless-steel 

jackets. 

2. Mark II (1954 – 1955): This core was 98% enriched uranium with a 2% zirconium alloy. 

On November 29, 1955, this core experienced a partial meltdown. 

3. Mark III (1957-1962): This core was also composed of 98% enriched uranium with a 2% 

zirconium alloy and was used to prove that there was nothing intrinsically unsafe about 

a breeder reactor after the partial meltdown of the Mark II core. 

4. Mark IV (1962 – 1963): This core was composed of 98.6% plutonium, 1.25% aluminum, 

and some impurities. The fuel was fabricated and canned at Rocky Flats in a joint effort 

with Argonne National Laboratory (Lemont, IL). 

NIOSH proposes that the most bounding of the last two EBR-I cores be used. The first two cores are during the 

current SEC time period, where NIOSH has already determined it is infeasible to reconstruct internal doses to 

fission products. While it is initially believed the plutonium core will likely be bounding, some preliminary 

modeling would need to be performed on both cores to confirm this. 

ANL-W: Experimental Breeder Reactor No. II (EBR-II)

Experimental Breeder Reactor No. II operated from September 1961 to September 1994. EBR-II was an 

unmoderated, sodium-cooled, fast-neutron power reactor, which was designed to produce 62.5 MWt. 
Page 8 of 9 

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ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or ABRWH (or their technical 

support and review contractors) positions, unless specifically marked as such. This document represents preliminary positions taken on technical issues 

prepared by NIOSH or its contractor. 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. 

Until September 1970, EBR-II consisted of a high-enriched uranium core surrounded on all sides by a 

blanket of depleted uranium. In September 1970, some of the EBR-II Reactor’s depleted-uranium blanket 

subassemblies were replaced with stainless-steel reflector subassemblies. It utilized 67% enriched uranium 

fuel pins. Since the EBR-II fuel loading is significantly different from the Fast Flux Test Facility modeled in 

ORAU-OTIB-0054, NIOSH agrees that EBR-II should be evaluated for OTIB-0054 applicability.

Conclusions 

 NIOSH proposes merging the INL and ANL-W high priority category reactors for evaluation of OTIB-0054 

applicability. NIOSH also proposes that after the evaluation of the high priority category reactors is 

completed, any concerns regarding the medium and low priority category reactors can then be 

addressed. 

 NIOSH proposes that the Loss of Fluid Test Facility (LOFT) be removed from consideration for evaluation 

of OTIB-0054 applicability at this time, due to nuclear operations not commencing until December 1978.

 NIOSH agrees that the Organic Moderated Reactor Experiment (OMRE) should be evaluated for OTIB0054 applicability due to its unique moderator and coolant. 

 NIOSH agrees that the Power Burst Facility (PBF) should be evaluated for OTIB-0054 applicability due to 

the use of ceramic fuel. 

 NIOSH proposes that a model for the most extreme experiment from all of the Special Power Excursion 

Reactor Tests (SPERT), in terms of possible departures from OTIB-0054, be used to represent the 

“bounding” case to cover all four SPERT reactors.

 Boiling Water Reactor Experiment (BORAX) No. I, II, and III all ceased operations before the end of the 

approved SEC period for ANL-W. NIOSH proposes BORAX I-III be removed from consideration for 

evaluation of OTIB-0054 applicability as their operating years are covered by the SEC period when 

bioassay data is known to be incomplete and an infeasibility to reconstruct doses has already been 

established. NIOSH agrees that BORAX-IV should be evaluated for OTIB-0054 applicability due to the use 

of uranium-thorium oxide fuel. NIOSH proposes that BORAX-V be removed from consideration for 

evaluation of OTIB-0054 applicability since its primary function was to evaluate steam superheating with 

essentially the same configuration as BORAX-IV. 

 NIOSH proposes that the most bounding of the last two EBR-I cores be used. While it is initially believed 

the plutonium core would be bounding, some preliminary modeling would need to be performed on all 

four cores to confirm this. 

 NIOSH agrees that the Experimental Breeder Reactor No. II should be evaluated for OTIB-0054 

applicability. 

NIOSH recommends the review for the high priority category INL and ANL-W reactors for OTIB-0054 applicability 

be categorized as presented below based on the reasons provided. 

Reactor #s for NIOSH-Proposed Grouping 

for OTIB-0054 Applicability Evaluation Reactor Name

35 Organic Moderated Reactor Experiment (OMRE)

36 Power Burst Facility (PBF)

39, 40, 41, 42 Special Power Excursion Reactor Tests I-IV (SPERT I-IV))

9 Boiling Water Reactor Experiment No. IV (BORAX-IV)

17 Experimental Breeder Reactor No. 1 (EBR-I) Core 4

18 Experimental Breeder Reactor No. II (EBR-II)
Page 9 of 9 

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ABRWH or its Working Groups or Subcommittees. Draft, preliminary, interim, and White Paper documents are not final NIOSH or ABRWH (or their technical 

support and review contractors) positions, unless specifically marked as such. This document represents preliminary positions taken on technical issues 

prepared by NIOSH or its contractor. 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. 

References 

SC&A 2016. INL SEC-00219 Reactor Prioritization for Evaluation of ORAUT-OTIB-0054 Applicability, SCA-TR-2016-

SEC002, Revision 0, March 2016

SC&A 2016. INL SEC-00219 Reactor Prioritization for Evaluation of ORAUT-OTIB-0054 Applicability, SCA-TR-2016-

SEC002, Revision 1, June 2016

SC&A 2015. Argonne National Laboratory-West SEC-00224 Reactor Prioritization for Evaluation of ORAUT-OTIB0054 Applicability, SCA-TR-2016-SEC010, Revision 0, July 2016 

SC&A 2015. NIOSH SEC-00219 Test Reactor Area Modeling, SCA-SEC-201509974A, Revision 0, September 2015 

SC&A 2015. Review of NIOSH Strategy for Reconstructing Internal Doses to Workers at Test Area North, SCA-TR2015-SEC0074A, Revision 0, September 2015 

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

Analyses, ORAU-OTB-0054, Revision 4, August 2015 

NIOSH 2015. SEC Petition Evaluation Report – Petition SEC-00219 – Idaho National Laboratory Revision 01, July 

2015 

NIOSH 2016. SEC Petition Evaluation Report – Petition SEC-00224 – Argonne National Laboratory – West, 

Revision 00, February 2016

Historical Engineering Record – Idaho National Engineering and Environmental Laboratory – Test Area North, 

HAER No. ID-33-E (SRDB 83751)