INL/ANL-W

Reactor Prioritization & 

Other Reactor-Related 

Issues

INL/ANL-W WG Meeting

Stephen L. Ostrow, PhD

SC&A, Inc.

May 16, 2017
Introduction

 This presentation summarizes SC&A report: INL SEC-00219 

and ANL-W SEC-00224: SC&A Response to NIOSH 

Reactor Analysis Plan and Consolidation of all Reactor 

Modeling Comments, SCA-TR-2016-SEC012, Rev. 0 

(12/8/2016) to inform the INL/ANL-W Work Group in 

providing guidance to NIOSH with respect to:

 Prioritizing new reactor and irradiated fuel characterization 

studies.

 Addressing specific SC&A concerns related to modeling Test 

Area North (TAN) and Test Reactor Area (TRA) operations.

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Background

• A primary tool NIOSH uses for internal dose reconstruction is ORAUTOTIB-0054, Fission and Activation Product Assignment for Internal DoseRelated Gross Beta and Gamma Analysis. 

• SC&A performed preliminary assessments in 2015 and 2016 of whether the 

OTIB envelopes, with sufficient accuracy, the important conditions of the 

INL and ANL-W reactors, and prioritized the reactors into High, Medium, 

and Low categories for further detailed investigations. 

• NIOSH responded (7/28/16) with a plan for additional reactor evaluations.

• SC&A’s latest report, INL SEC-00219 and ANL-W SEC-00224: SC&A 

Response to NIOSH Reactor Analysis Plan and Consolidation of all Reactor 

Modeling Comments (12/8/16), addresses the NIOSH report, consolidates 

all SC&A comments related to reactor evaluation prioritization, and adds an 

appendix on workforce and exposure potential. 

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INL/ANL-W

 Operations at the INL and ANL-W sites involving radioactive 

materials were very complex, and many unique nuclear 

reactors and experiments were built and tested, with different 

fuel types, blankets, reflectors, moderators, coolants, operating 

scenarios, and burnups. 

 There were a total of 52 reactors at the INL site.

INL Site Reactors

Location Number

INL 34

ANL-W 12

NRF (Naval Research Facility) 4

Never Operated 2

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ORAUT-OTIB-0054

 NIOSH uses OTIB-0054 to determine internal doses when 

only gross beta or gross gamma measurements are available 

by assigning 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]). 

 The OTIB generated 9 different representative cases based 

on four reactors, which are intended to envelope the range of 

reactor and nuclear fuel types and operating scenarios. 

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Reactor Type 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
SC&A Priority Rankings

 SC&A assigned priority rankings to each applicable reactor for 

further, detailed analyses, considering the factors below, which 

are likely to indicate if the radionuclide mixtures in OTIB-0054 

might result in an under or unrealistic estimate of internal 

doses.

 Fuel type, enrichment, and cladding

 Moderator and reflector

 Coolant 

 Operational mode and whether it was inside or outside design limits

 Length of operation/overall burnup 

 Number (approx.) of potentially exposed workers (Appendix A)

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

unintended/unprotected exposures

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SC&A Priority Rankings

 SC&A categorized seven INL and seven ANL-W reactors in 

the High Priority category for further studies:

 INL: LOFT (Loss of Fluid Test Facility); OMRE (Organic 

Moderated Reactor Experiment); PBF (Power Burst Facility); 

SPERT-I, -II, -III, -IV (Special Power Excursion Reactor Test) 

 ANL-W: BORAX-I, -II, -III, -IV, -V (Boiling Water Reactor 

Experiment); EBR-1, -II (Experimental Breeder Reactor) 

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NIOSH Response

NIOSH proposed (7/28/16):

 Merging the INL and ANL-W high priority category reactors for 

detailed evaluation of OTIB-0054 applicability using the 

ORIGEN isotope generation and depletion code. 

 Eliminating several reactors from the high priority category: 

LOFT, BORAX-I, -II, -III, -V. Reasons are given in the report. 

 Modeling the most extreme experiment from all four of the 

SPERT reactor tests as a “bounding case.”

 Modeling the most bounding case of the last two EBR-I cores. 

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NIOSH Response

Summary: Reactors that NIOSH Proposes to Evaluate

 OMRE Organic Moderated Reactor Experiment

 PBF Power Burst Facility

 SPERT I-IV Special Power Excursion Reactor Tests

 BORAX-IV Boiling Water Reactor Experiments

 EBR-I (Core 4) Experimental Breeder Reactor-I

 EBR-II Experimental Breeder Reactor-II

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SC&A Response to NIOSH

No.(a) NIOSH Recommendation(b) SC&A Response

1 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.

SC&A concurs. Whether a reactor is classified as an INL 

reactor or an ANL-W reactor is immaterial to the reactor 

modeling work. Treating them together and generating a single 

report would reduce repetition in NIOSH’s report-writing and the 

Board’s and SC&A’s reviewing efforts. 

2 NIOSH proposes that the LOFT be removed 

from consideration for evaluation of OTIB0054 applicability at this time because 

nuclear operations did not commence until 

December 1978.

SC&A recognizes that the first five LOFT experiments were nonnuclear, thermal-hydraulic experiments and that the potential for 

radiation exposure did not occur until December 1978, which is 

after the INL SEC-00219 period. SC&A believes that, given the 

facility’s size, long operating history, beyond-design-basis 

operating scenarios, and potential to have exposed a significant 

number of personnel, the LOFT reactor merits a more detailed 

examination with respect to whether it can be adequately 

modelled by OTIB-0054. Such an examination could be 

conducted as a site profile exercise. 

3 NIOSH agrees that the OMRE should be 

evaluated for OTIB-0054 applicability due to 

its unique moderator and coolant.

SC&A agrees with NIOSH’s characterization with respect to 

OTIB-0054 and notes (Section A.5) that, based on the limited 

data available (only for the last year of operation), there appears 

to be a significant potential for exposure of hundreds of regular 

workers and visitors. 

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SC&A Response to NIOSH (cont.)

No.(a) NIOSH Recommendation(b) SC&A Response

4 NIOSH agrees that the PBF should be 

evaluated for OTIB-0054 applicability due 

to the use of ceramic fuel.

SC&A agrees with NIOSH’s characterization with respect to 

OTIB-0054 and notes (Section A.3) that, based on the 

limited data available (only for the first few years of 

operation), there appears to be a potential for exposure of 

typically less than 100 regular workers and visitors per 

month.

5 NIOSH proposes that a model for the most 

extreme experiment from all of the SPERT, 

in terms of possible departures from OTIB0054, be used to represent the “bounding” 

case to cover all four SPERT reactors.

SC&A disagrees with NIOSH’s recommendation. Although 

the four SPERT reactors were all part of the same series of 

reactor experiments that subjected the reactor systems to 

large reactivity excursions, as seen in the summaries of 

Table 1, they differed significantly from each other and 

should be examined separately, perhaps by choosing the 

“worst case” scenario for each reactor. NIOSH should justify 

in its report its choice, perhaps by performing some 

preliminary calculations to determine the “bounding case”

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SC&A Response to NIOSH (cont.)

No.(a) NIOSH Recommendation(b) SC&A Response

6 BORAX Nos. I, II, and III all ceased 

operations before the end of the approved 

SEC period for ANL-W. NIOSH proposes 

BORAX Nos. 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 are 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 because its primary function 

was to evaluate steam superheating with 

essentially the same configuration as 

BORAX-IV.

SC&A concurs with NIOSH’s assessment about OTIB-0054 

and notes (Appendix A) that individual documentation 

concerning the workforce at the BORAX-IV experiment in 

1958 could not be located and, therefore, is not discussed 

further. 

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SC&A Response to NIOSH (cont.)

No.(a) NIOSH Recommendation(b) SC&A Response

7 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.

SC&A concurs with NIOSH about OTIB-0054 and expects 

that the resulting NIOSH report will make a compelling case 

for which core is bounding. In addition, SC&A notes (Section 

A.1) that several hundred workers and visitors were present 

during the period of operation for the MARK IV core. 

8 NIOSH agrees that EBR-II should be 

evaluated for OTIB-0054 applicability.

SC&A concurs with NIOSH about OTIB-0054 and notes 

(Section A.2) that hundreds of workers and visitors could 

have been exposed each year; in some years, the average 

worker penetrating doses were greater than 100 mrem. 

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Notes: 

The above table is taken from: INL SEC-00219 and ANL-W SEC-00224: SC&A Response to NIOSH 

Reactor Analysis Plan and Consolidation of all Reactor Modeling Comments,12/8/16. “Bolding” added

here. 

(a) Numbering added here from bulleted list in NIOSH 7/28/16 report.

(b) Recommendations copied from the NIOSH 7/28/16 Conclusions section.
SC&A Response to NIOSH

 Appendix A of the 12/8/16 SC&A report examines exposure 

potential:

 Prioritized reactor sites generally employed 100s of monitored 

workers, except for the PBF, which appears to have assigned 

only about 30 workers during most of its badging cycles. 

 Penetrating doses at the prioritized reactor sites were significant, 

with some monthly badging cycles averaging 100s of mrem.

 While external exposures do not necessarily imply internal 

exposure potential, the magnitude of these external doses is 

indicative of the source terms being considered. 

 Coupled with the extensive internal dosimetry program at INL for 

fission products, an adequate characterization of the mix of 

source term contaminants appears warranted. 

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SC&A Response to NIOSH

 SC&A issued targeted reports on 9/28/15 examining the 

applicability of OTIB-0054 to reactors in the INL TAN and 

TRA facilities. Comments from these two reports are collected 

in the SC&A 12/8/16 report for completeness.

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Test Area North (TAN)

 SC&A examined the applicability of OTIB-0054 and Tables 5-22 and 5-23 

(Actinide to Sr-90 and Actinide to Cs-137 Ratios, respectively) of the INL 

Internal Dosimetry TBD to the performance of internal dose reconstruction 

for facilities that handled and stored spent and irradiated fuel at TAN. 

 The spent and irradiated fuel from the Heat Transfer Reactor 

Experiment (HTRE) tests was of particular interest because the reactor, 

fuel, and operational combinations that underpin the OTIB-0054 

methodology reflect situations in which burnup often occurred over 

protracted periods of time (hundreds of days) and the fuel maintained its 

integrity. In contrast, the fuel at HTRE had very short burnup times and the 

reactors operated at high temperatures, allowing the fuel to melt. 

 In addition, HTRE employed highly enriched uranium, as opposed to the 

enrichment levels of the uranium in the fuel used to derive the mix of 

fission and activation products provided as default mixes in OTIB-0054 

and TBD Tables 5-22 and 5-23.

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Test Area North (TAN)

 To explore these potential concerns, SC&A performed 

ORIGEN-ARP runs, where the isotopic mixtures of fission and 

activation products were compared at different lengths of 

continuous operation at 20 MW (low power) and at 200 MW 

(high power): 20 hours and 200 days for both power levels, 

and an additional 20 day-run for the high-power case. 

 These scenarios were intended to provide a general 

representation of the operating conditions of the HTRE tests 

(e.g., HTRE-1 was operated at 20 MW for 150.8 hours). 

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Test Area North (TAN)

 The 200-day cases are indicative of the long burnup times 

used to derive the relative mixes of radionuclides in Tables 

7‐3a through 7-3i (the nine characteristic reactor cases) of 

OTIB-0054, while the shorter burnup times (20 days and 20 

hours) are typical of many of the TAN experiments. 

 SC&A analyses assumed a fuel cool-down period of 10 days, 

and then multiplied the relative amounts of fission products 

produced by selected organ dose conversion factors (DCFs), 

which yielded a relative “index of harm” for each fission 

product. SC&A then summed the indices of harm for the 

fission products for each of the burnup durations. 

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Test Area North (TAN)

 In summary, the results of these investigations revealed the 

following for fission products:

• For the high power level (200 MW), the indices of harm for the 

20-day burnup and the 20-hour burnup were about the same as, 

slightly higher than, or slightly lower than (i.e., claimant 

favorable), the 200-day burnup for all organs of concern, except 

for the thyroid, for which the relative index of harm was 

substantively higher (i.e., 8.29).

• For the low power level (20 MW), the derived indices of harm for 

the 20-hour burnup compared to the 200-day burnup for all 

organs of concern were not claimant favorable.

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Test Area North (TAN)

 For actinide activation products, for all cases analyzed, the 

ratios of the inventories of all actinide activation products to the 

inventories of Cs-137 and Sr-90 were grossly overestimated 

compared to the ratios in Tables 5-22 and 5-23 of the TBD. 

SC&A’s scoping analyses stopped at this point.

 SC&A recommends that NIOSH continue these types of 

investigations to better understand the applicability and 

limitations of OTIB-0054 and TBD Tables 5-22 and 5-23 

applied to reconstructing internal doses to TAN workers 

involved in irradiated and spent fuel operations, where the 

power levels and burnup durations were significantly different 

from those upon which the isotopic mixes were derived in 

OTIB-0054 and TBD Tables 5-22 and 5-23.

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Test Reactor Area (TRA)

 Another 9/28/15 SC&A report examined the three TRA reactors and their 

operations to determine if they are adequately enveloped by OTIB-0054: 

Materials Test Reactor (MTR; 1952–1970), Engineering Test Reactor 

(ETR; 1957–1981), and ATR (1967–present). 

 All are material testing reactors with similar designs, but with size, power 

level, and capabilities increasing from the first to the last; the maximum 

power level was MTR - 40 MW; ETR - 175 MW; and ATR - 250 MW. 

 Their high-flux capability allows for the accelerated simulation of long-term 

irradiation of reactor materials. Each is a pressurized, light-watermoderated, beryllium-reflected reactor, primarily using highly enriched 

uranium fuel arranged in an unusual curved plate configuration. The 

primary reactivity control mechanism consists of rotating beryllium 

cylinders with hafnium shells. 

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Test Reactor Area (TRA)

 SC&A found that OTIB-0054 adequately envelopes the three 

TRA reactors (the OTIB explicitly models the ATR) for uranium 

fuel operations, but noted that the MTR also ran for a period 

with plutonium (Pu) fuel; in 1958 it became the first reactor 

run with a Pu-239 core. The MTR-Phoenix experiment was a 

demonstration project for a potential high-power, compact 

reactor to convert the fertile Pu-240 to the fissile Pu-241 

through neutron capture. 

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Test Reactor Area (TRA)

 It is not clear which, if any, of the nine OTIB-0054 

representative reactor cases would adequately envelope the 

MTR with plutonium fuel. Although the MTR plutonium fuel 

operations used fuel plates physically similar to those used in 

the uranium fuel, and the rest of the reactor configuration was 

not significantly modified, the nuclear properties of plutonium 

differ from those of uranium, and the fission product 

abundance distribution and core neutron spectrum (and, 

hence, activation product yield) would be different. The issue 

of whether OTIB-0054 adequately envelopes the MTR when 

fueled with plutonium merits further investigation and 

discussion by NIOSH. 

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