INL Test Reactor Area
 

Nuclear Modeling
 

Stephen L. Ostrow, PhD
 

S. Cohen and Associates
 Contractor to:
 

Advisory Board on Radiation and Worker Health/ABRWH
 

Centers For Disease Control and Prevention
 

November 10, 2015
 
 

 

 

 

Presentation Content 

•	 Background of Investigation: Does the methodology of 

ORAUT-OTIB-0054 (Fission and Activation Product Assignment

for Internal Dose-Related Gross Beta and Gross Gamma 

Analysis) adequately model the reactor characteristics and

operations of the Test Reactor Area (TRA)? 

o Are the reactors’ and operating scenarios adequately enveloped by

the OTIB cases so that the isotopic ratios are valid? 

o Have all off-normal operating scenarios been identified and are they

adequately enveloped by the OTIB methodology? 

•	 Approach: 

o	 Describe the OTIB 

o	 Describe the TRA reactors 

o	 Assess whether the OTIB models the reactors 

(Presentation material taken from SC&A report, NIOSH SEC-00219 

Test Reactor Area Modeling, SCA-SEC-2015-0074-C, Rev. 0, 9/28/15)
 2 
 

 

 

 

 

 

 

 

Background of Investigation 

•	 NIOSH SEC Petition Evaluation Reports 

o	 March 12, 2015: Rev. 00 

o	 July 21, 2015: Rev. 01 

•	 Dose Reconstruction 

o	 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 

o	 This investigation examines one aspect of the dose 

reconstructability assumption for one of the several major site 

areas, the Test Reactor Area (TRA) 

o	 Applicability of ORAUT-OTIB-0054 to the more “exotic” reactors at 

Test Area North (TAN) is addressed in a separate report 

3 
 

 

 

ORAUT-OTIB-0054 Scope
 

•	 This guidance applies to a broad scope of reactor operations 

including: 

o	 plutonium production reactors (low enrichment, low 

burnup, Zircaloy or aluminum cladding) 

o	 research reactors [modest enrichment, modest burnup,

stainless-steel (SS) or Zircaloy cladding; e.g., Training,

Research, Isotope General Atomics (TRIGA) reactors] 

o	 high-enrichment, high-burnup reactors [e.g., Idaho

National Laboratory’s Advanced Test Reactor (ATR), fuel

from naval reactors], and fast breeder reactors [e.g.,

the Hanford Site’s Fast Flux Test Reactor (FFTF),

Argonne National Laboratory–West’s experimental

breeder reactors]. 

(ORAUT-OTIB-0054) 

4 
 

 

 

 

ORAUT-OTIB-0054 Scope
 

• It does not apply to: 

o	 operations involving decay times shorter than 10 

days (e.g., radioactive lanthanum (RaLa) processing) 

o	 determination of intakes if radionuclides have been 

purposely extracted and concentrated, as for heat 

generation sources, medical uses, or waste handling 

operations that caused significant alteration to the 

source term to which workers were exposed 

(ORAUT-OTIB-0054) 

5 
 

 

 

 

Prior Evaluation of ORAUT-OTIB-0054
 

•	 SC&A evaluated the general validity and 

applicability of the OTIB as part of the review 

process of the Subcommittee on Procedures 

Review, and all findings were closed by the 

Advisory Board 

•	 The details of the evaluation and resolution 

process will not be recapitulated here 

6 
 

 

 

 

 

 

ORAUT-OTIB-0054 Development
 

•	 Frequently, air-sampling or urinalysis data on worker 

exposure to mixed fission and activation products 

associated with nuclear reactors or fuel are available 

only in the form of gross beta or gross gamma activity 

unattributed to specific radionuclides. 

•	 The OTIB provides guidance on assigning radionuclidespecific intakes using ratios to Cs-137 or Sr-90 indicator 

radionuclides. 

•	 The goal of the OTIB is to reduce a large amount of 

reactor fuel isotopic data into a simplified, 

representative set that could be used by dose 

reconstructors looking at actual claimant cases. 

7 
 

 

 

 

 

ORAUT-OTIB-0054 Development
 

•	 The starting point is the radionuclide mix in spent 

fuel for a number of different reactor types and 

fuel designs operated under a variety of different 

conditions (e.g., specific power, irradiation time, 

and burnup) calculated at several different decay 

times following removal from the reactor 

•	 It is assumed that an actual claimant case 

condition will fit somewhere within the 

parameter space defined by the selected 

representative reactor types 

8 
Initial Representative Reactors 

The OTIB starts with 7 representative reactors
 

9 

Reactor Category 

Hanford N-Reactor 

Hanford Single-Pass Reactors 

Plutonium Production Reactors 

Fast Flux Test Facility (FFTF) Sodium-Cooled Fast Reactors 

Advanced Test Reactor (ATR) High-flux Reactors 

TRIGA® Reactor (Al-clad fuel) 

TRIGA® Reactor (SS-clad fuel) 

Research Reactors 

Pressurized Water Reactor (PWR) Generic Reactor 
 

 

 

Reduction of Representative Reactors
 

•	 The ORIGEN2 isotope generation and depletion code

calculated isotopic inventories for the 7 reactors in 11 

runs, decayed for 7 decay times, which produced activity

data for 879 fission product nuclides and 688 activation 

product nuclides 

•	 The activities in each dataset were normalized to that of 

Cs-137 at 10 days of decay. Four cases were then 

selected as representative for a wide variety of reactors, 

fuel types, and operational parameters: 

Advanced Test Reactor (ATR) – High flux reactors
 

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

Hanford N-Reactor – Pu production reactors
 

TRIGA with stainless steel cladding – Research reactors
 

10 
 

 

 

 

 

Reduction of Representative Reactors
 

•	 ORIGEN-S (SCALE system) runs then produced a 

total of 9 representative cases for the 4 reactors. 

•	 NIOSH customarily considers all 9 reactor cases 

when doing a dose reconstruction and, in the 

absence of individual worker information, might 

apply data for 4 decay times (10 days, 40 days, 

180 days, and 1 year) as well, if required by a 

specific dose reconstruction case. 

11 
ORAUT-OTIB-0054 Representative Cases
 ORIGEN-S Irradiation Parameters for the Nine Representative Reactor Cases 

Case Parameters Basis 

ATR 1 Specific power = 2,379.1 MW/MTU Irradiation Maximum burnup at nominal power. 

time =132.27 days 

Burnup =314,684 MWd/MTU 

ATR 2 Specific power = 8,651.2 MW/MTU Irradiation 

time =36.4days
 

Burnup =314,904 MWd/MTU
 

Maximum burnup at maximum assembly power. 

ATR 3 Specific power = 2,379.1 MW/MTU Irradiation Nominal burnup at nominal power.
 

time =56days
 

Burnup =133,230 MWd/MTU
 

FFTF 1 Specific power = 163.8 MW/MTHM Irradiation Maximum burnup at nominal power. 

time =929.4 days 

Burnup =152,230 MWd/MTHM 

FFTF 2 Specific power = 163.8 MW/MTHM Irradiation 

time =488.3 days
 

Burnup =79,984 MWd/MTHM
 

Nominal burnup at nominal power.
 

N Reactor 1 Specific power = 10.4 MW/MTU Irradiation Production of weapons -grade plutonium (nominal 6% 

time =114.2 days Burnup = 1,188 Pu-240 content) at nominal power. 

MWd/MTU 

N Reactor 2 Specific power = 10.4 MW/MTU Irradiation Production of fuel -grade plutonium (nominal 12% Pu-240 

time =285.6 days Burnup = 2,970 content) at nominal power.
 

MWd/MTU
 

TRIGA 1
 Specific power = 15.57 MWd/MTU Irradiation 

time =730.1 days
 

Burnup =11,368 MWd/MTU
 

Maximum burnup at nominal power. 

TRIGA 2 Specific power = 15.57 MW/MTU Irradiation Nominal burnup at nominal power.
 

time =115.2 days Burnup = 1994 MWd/MTU
 

12 
 

 

 

 

 

 

 

 

Advanced Test Reactor
 

•	 Surrogate for high-flux reactors, as might be encountered in 

materials testing or experimental reactors 

•	 The ATR, which had a maximum power level of 250 MW, 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), and has 

operated for over 40 years to the present 

•	 The ATR is a pressurized, light-water moderated, beryllium-reflected 

reactor, using highly enriched uranium fuel (93.15% nominal 

enrichment) arranged in an unusual curved plate configuration 

•	 The primary reactivity control mechanism consists of a unique design 

of rotating beryllium cylinders with hafnium thermal neutron 

absorber shells 

•	 The OTIB models three cases with different specific powers, 

irradiation times, and burnups 

13 
 

 

 

Fast Flux Test Facility
 

•	 Represents sodium-cooled fast reactors, which use fast neutron

fission to burn both depleted uranium and plutonium. The concept

can also breed plutonium fuel through neutron capture in U-238. 

•	 The FFTF was a 400 MW, liquid sodium-cooled reactor located at the 

DOE Hanford site that operated from April 1982 to April 1992, 

primarily to explore reactor designs and operations that might be 

applicable to a commercial fast breeder reactor. 

•	 The fuel was mixed oxide with plutonium (86 wt% Pu-239) amounting 

to 29 wt% of the heavy metal (uranium plus plutonium); the driver

assemblies were clad with 316 SS (stainless steel); and the assemblies 

were subjected to high burnups. 

•	 The OTIB models two cases with different specific powers, irradiation

times, and burnups 

14 
 

 

 

 

 

Hanford N-Reactor 

•	 Represents Pu production reactors, where it is desired to 

have low enrichment fuel (maximizing U-238, which 

absorbs a neutron and produces Pu-239 following two beta

decays) and short irradiation time (to minimize the

undesirable “contaminant” Pu-240). 

•	 The N-Reactor, which operated from 1963 to 1987, was a 

low-enrichment, graphite-moderated, pressurized water- cooled reactor that initially produced Pu for weapons and 

was later modified to produce steam to generate electricity 

• The core of the N-Reactor was a 1800-ton graphite block,

33 feet high by 33 feet wide by 39 feet (12 meters) long
 

•	 The OTIB models two cases with different specific powers,

irradiation times, and burnups 

15 
 

 

 

 

TRIGA Reactor
 •	 The TRIGA reactor with stainless steel clad fuel represents research and 

radiographic reactors. 

•	 The low-power, water-cooled, pool-type, uranium zirconium hydride-fueled

reactors are manufactured in a variety of versions by General Atomics (GA) 

primarily for research and educational purposes, and are considered inherently 

safe due to their very large prompt negative temperature coefficient of 

reactivity that quickly drops the reactor subcritical in response to any power

excursions. This feature also allows the TRIGA to be pulsed to a high power

level, followed by a quick automatic self-shutdown. 

•	 GA has supplied over 65 reactors throughout the world over the past 45 years,

ranging in power from 20 kW to 16 MW and containing 60–100 fuel elements.

The models have progressed from Mark I through Mark III, with a number of

specially modified units also produced. The early models used highly enriched 

uranium (70%), but the later models reduced the enrichment to 20% and the 

earlier models were modified to accept lower enrichment fuel. 

•	 The OTIB models two cases with different specific powers, irradiation times,

and burnups 

16 
 

 

 

 

 

 

 

 

INL Test Reactor Area
 

Table 5-1 of the Evaluation Report lists the eight TRA 

reactors, only the first three of which were (or are, in the 

case of the ATR) high-power, high-flux reactors used for 

material testing. The remaining critical facilities were 

low-power reactors used for reactivity measurements and 

are not considered in this analysis. 

SC&A examined whether ORAUT-OTIB-0054 adequately 

envelopes the parameters of the three materials testing 

reactors, including the fuel and cladding composition, 

power level profile, burnup, and decay time following 

removal from the core. 

17 
 

 

 

INL Test Reactor Area
 

Facility Name Building No. First Used Last Used Facility Type 

Materials Test Reactor (MTR) TRA-603 1952 1970 Reactor 

Engineering Test Reactor (ETR) TRA-642 1957 1981 Reactor 

Advanced Test Reactor (ATR) TRA-670 1967 In use Reactor 

Reactivity Measurement Facility (RMF) TRA-603 1954 1962 Reactor 

Advanced Reactivity Measurement Facility No. I (ARMF-I) TRA-660 1960 1974 Reactor 

Advanced Reactivity Measurement Facility No. 2 (ARMF-II); renamed Coupled 

Fast Reactivity Measurement Facility (CFRMF) TRA-660 1962 1991 Reactor 

Engineering Test Reactor Critical Facility (ETRCF) TRA-654 1957 1982 Reactor 

Advanced Test Reactor Critical Facility (ATRC) TRA-670 1964 In use Reactor 

Hot Cell Facility TRA-632 1954 In use Research 

Gamma Facility TRA-641 1955 ND Research 

Radiation Measurement Laboratory (RML) TRA-604 1952 In use Sampling/ Research 

Radiochemistry Laboratories TRA-604 1950 In use Sampling/ Research 

Alpha Laboratories TRA-652 1961 1970 Sampling/ Research 

HB-4 Crystal Spectrometer TRA-603 1952 1970 Research 

Fast (Neutron) Chopper TRA-665 1952 1970 Research 

18 
 

 

 

 

 

SC&A Evaluation – ATR
 

Advanced Test Reactor 

•	 OTIB-0054 explicitly models the fuel characteristics 

and operations of the ATR, so it is expected that 

any workers exposed to ATR fuel during reactor 

operations or out of the reactor, but before any 

isotopic separation, would be adequately treated 

by the methodology in the OTIB in order to 

determine internal exposures. 

•	 SC&A did not find material instances of the ATR 

operating outside of its design envelope. 

19 
 

ATR Illustrations
 

(Wikipedia) 

(Stacy, S., Proving the Principle: A History of the Idaho National Engineering 

and Environmental Laboratory 1949-1999, DOE/ID-10799, 2000) 

20 
 

 

 

SC&A Evaluation – MTR
 

Materials Test Reactor 

•	 The MTR, designed by Argonne National Laboratory and 

Oak Ridge National Laboratory and sited at INL, was the

second reactor built at that location. 

•	 It achieved first criticality on March 31, 1952, and 

attained full power of 30 MW on May 22, 1952. The 

power level was later increased to 40 MW in September

1955. 

•	 It was last used in 1970 with a plutonium core, and was

succeeded by the larger and higher-flux ETR and ATR in 

the same TRA facility complex. 

21 
 

 

 

 

 

SC&A Evaluation – MTR
 

•	 The MTR was cooled and moderated with light water, and 

used Al-clad, curved plate, enriched uranium fuel the majority 

of the time. Beryllium, graphite, and light-water neutron 

reflectors surrounded the core and control rods entered from 

above. 

•	 The core was quite small by modern standards, measuring 9 in 

× 28 in in cross section × 24 in high and containing only a 

maximum total of 4.9 kg of U-235. 

•	 The MTR provided a high neutron and gamma flux for the

accelerated testing of materials for reactor development, and

had about 100 beam holes that penetrated into the core and 

its surroundings from several different directions. Samples for

irradiation and experiments were provided by several different 

laboratories. 

22 
 

 

 

 

 

SC&A Evaluation – MTR
 

•	 The MTR operated for > 125,000 hrs and performed

> 19,000 irradiations. 

•	 The MTR fuel enrichment, cladding, and plate design

were similar in important nuclear aspects to those of

the ATR, and both reactors were light-water cooled

and moderated, with beryllium and aluminum

reflectors. Its power level and burnup, however,

were considerably less than those of the ATR. Hence, 

the OTIB-0054 methodology, which explicitly models

the ATR, should also adequately envelope the MTR in

considering internal exposures when the later 

reactor operated with uranium fuel. 

23 
 

 

 

 

SC&A Evaluation – MTR
 

Radioactive Lanthanum (RaLa) 

•	 The RaLa extraction campaign from 1956 through 1963 

was an “unusual” operating scenario. The fuel was 

irradiated for only 17 days, then quickly removed and

dissolved to recover the short half-life fission product, 

which was used at Los Alamos National Laboratory to test

implosion technologies (explosive lensing) to compress 

plutonium pits. 

•	 This campaign does not have to be considered here since, 

as stated in Section 3.0 of OTIB-0054: “It [OTIB-0054] does 

not apply to operations involving decay times shorter than

10 days (e.g., early radioactive lanthanum (RaLa) 

processing).” 

24 
 

 

 

 

 

SC&A Evaluation – MTR 

Plutonium Core 

•	 Although the MTR initially used uranium 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, thereby extending

the lifetime of the fuel and avoiding loading the core with excessive 

reactivity and neutron absorbers at startup as the Pu-240 neutron

absorber was slowly replaced by the fissile Pu-241. 

•	 The experiment used curved fuel plates similar in configuration to

the standard MTR fuel plates, where each Al-clad plate contained 21

wt% Pu and 79 wt% Al. 

•	 The reactor reached initial power on January 28, 1970, and 

operated at various power levels up to 24 MW until April 23, 1970, 

at which time it had accumulated a burnup of 923 MWd. This 

compared to the standard U-235 core, which operated at 40 MW. 

25 
 

 

 

 

SC&A Evaluation – MTR
 Plutonium Core 

•	 Although the MTR plutonium fuel operations used fuel

plates similar to those used with the uranium fuel and

the rest of the reactor configuration was not 

significantly modified, the nuclear properties of

plutonium (e.g., cross sections) differ from those of

uranium, and the fission product abundance 

distribution and core neutron spectrum (and, hence 

activation product yield) would be different. 

•	 How much different, and whether the differences 

would be radiologically significant, would require 

detailed comparative ORIGEN runs, which were not 

done for this report. 

26 
 

 

 

SC&A Evaluation – MTR
 

Plutonium Core 

•	 The Hanford N Reactor cases, although containing plutonium, are 

likely not applicable for several reasons. One obvious difference is 

that the MTR was water-moderated, while the Hanford N Reactor 

was graphite-moderated. In addition, the N-Reactor 1 case 

(production of weapons-grade Pu with nominal 6% Pu-240) and the

N-Reactor 2 case (production of fuel-grade Pu with nominal 12%

Pu-240) started out with a low-enrichment uranium core and 

sought to breed plutonium as the end product, while the MTR

started out with a plutonium core and used fertile Pu-240 (23%) to

breed fissile Pu-241. 

•	 It is not clear which, if any, of the nine OTIB-0054 cases would

adequately envelope the case of the MTR with a plutonium core. 

Hence, whether the MTR with a plutonium core can be adequately

modeled with ORAUT-OTIB-0054 has not been determined at this 

time. 

27 
 

 

 

 

 

Engineering Test Reactor – ETR
 •	 The ETR, located south of, and adjacent to, the MTR, was 

designed taking advantage of information gathered and 

lessons learned from the operations and limitations of the

MTR. 

•	 The ETR was larger than the MTR, with four times the

neutron flux (both thermal and fast), with its main control

rods entering from the bottom (decluttering the operating

floor), without troublesome beam holes (thereby lessening

the potential for neutron streaming out of the reactor), and 

with placement of test materials and experiments directly in

the core. 

•	 The experimental spaces were sufficiently large to 

accommodate experiments with their own closed coolant

loops. 

28 
 

 

 

 

 

 

Engineering Test Reactor – ETR
 •	 The ETR achieved first criticality on September 19, 1957, 

and operated at a maximum power level of 175 MW until

December 1981. 

•	 The reactor was light-water cooled and moderated and 

uranium fueled. The reactor core was in a square

configuration with an average diameter of 32 inches and an 

active height of 36 inches. Beryllium and aluminum neutron 

reflectors were placed outside the core. Initially, a new core

contained about 14 kg of U-235, which was increased to

23.7 kg in January 1963. 

•	 The average thermal flux in the core exceeded

1014n/(cm2-s). The 49 fuel assemblies (increased to 52 in

January 1963) contained 19 Al-clad fuel plates per assembly. 

The core life before refueling was initially 3,500 MWd,

which increased to 4,500 MWd in January 1963. 

29 
 

 

 

 

Engineering Test Reactor – ETR 

•	 A notable extension of the reactor occurred in 

1972, when a Sodium Loop Safety Facility with 

accompanying cooling and other systems was 

added. This was in support of the DOE’s 

breeder reactor program. 

•	 As with the MTR operating with uranium fuel,

the OTIB-0054 methodology, which explicitly

models the ATR, should also adequately

envelope the ETR in considering internal 

exposures. 

30 
 

 

 

 

 

 

 

 

Summary and Path Forward
 

• Resolve potential issues of applicability of ORAUTOTIB-0054 to MTR operating with plutonium fuel.
 

•	 The analysis of applicability of ORAUT-OTIB-0054 to 

more “exotic” reactors at Test Area North (TAN), with 

different fuel compositions and arrangements and 

operations than the OTIB representative reactors, 

under normal and off-normal conditions, is 

addressed in the SC&A report on TAN. Other 

experimental reactors located in several different 

areas on the INL site are yet to be addressed. 

31 
Comments and Questions?
 

32