Draft

ADVISORY BOARD ON

RADIATION AND WORKER HEALTH

National Institute for Occupational Safety and Health

INL SEC-00219 AND ANL-W SEC-00224: SC&A RESPONSE TO 

NIOSH REACTOR ANALYSIS PLAN AND CONSOLIDATION 

OF ALL REACTOR MODELING COMMENTS

Contract No. 211-2014-58081 

SCA-TR-2016-SEC012, Revision 0 

Prepared by

Stephen L. Ostrow, PhD

John Mauro, PhD, CHP 

Stephen F. Marschke 

Bob Barton, CHP 

SC&A, Inc.

1608 Spring Hill Road, Suite 400 

Vienna, Virginia, 22182 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755 

December 2016

DISCLAIMER

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

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

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

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

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

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

issues.

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

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

Health Review of NIOSH Dose Reconstruction Program

DOCUMENT TITLE: 

INL SEC-00219 and ANL-W SEC-00224: SC&A Response to 

NIOSH Reactor Analysis Plan and Consolidation of All Reactor 

Modeling Comments

DOCUMENT NUMBER/

DESCRIPTION:

SCA-TR-2016-SEC012 

REVISION NO.: 0 (Draft)

SUPERSEDES: N/A

EFFECTIVE DATE: December 8, 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] 

Record of Revisions

Revision 

Number

Effective 

Date Description of Revision

0 (Draft) 12/08/2016 Initial issue
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TABLE OF CONTENTS

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

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

2.0 NIOSH Proposal ............................................................................................................... 14

3.0 SC&A Response to the NIOSH Proposal ......................................................................... 16

4.0 Test Area North and Test Reactor Area Reactors ............................................................. 19

4.1 TAN Reactors and Associated Irradiated and Spent Fuel .................................... 19

4.2 TRA Reactors........................................................................................................ 21

5.0 References......................................................................................................................... 23

Appendix A: Characterization of Workforce and Exposure Potential for Six Reactors Selected 

for Analysis at INL/ANL-W ......................................................................................................... 25

Appendix B: Evaluation of Applicability of OTIB-0054 to Reconstructing Internal Doses at 

Test Area North............................................................................................................................. 53
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ABBREVIATIONS AND ACRONYMS

ATR Advanced Test Reactor

Advisory Board Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission 

Al aluminum 

ANL-W Argonne National Laboratory-West

ATR Advanced Test Reactor

Be beryllium 

BORAX Boiling Water Reactor Experiment

Bq becquerel

CPP Chemical Processing Plant

Cs cesium

D2O deuterium oxide (“heavy” water)

DCF dose conversion factor

DOE U.S. Department of Energy 

EBR Experimental Breeder Reactor

EPA U.S. Environmental Protection Agency 

ETR Engineering Test Reactor

F Fahrenheit

FGR Federal Guidance Report

H&S Health and Safety

H2O (“light”) water

HTRE Heat Transfer Reactor Experiment

I iodine 

IET initial engine test

INL Idaho National Laboratory

LLI lower large intestine

LOCA loss-of-coolant accident

LOFT Loss of Fluid Test Facility

LWR light-water reactor

mrem millirem

ms millisecond
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MTR Materials Test Reactor

MTU metric ton uranium 

MW megawatt

MWe megawatt electric

MWth megawatt thermal

Na sodium 

NaK sodium-potassium (liquid metal)

Nb niobium 

NRC U.S. Nuclear Regulatory Commission 

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

Pr praseodymium 

psig pounds per square inch gage 

Pu plutonium 

RaLa radioactive lanthanum

Ru ruthenium 

SEC Special Exposure Cohort 

SL-1 Stationary Low-Power Reactor

SNAP Systems for Nuclear Auxiliary Power

SPERT Special Power Excursion Reactor Test(s)

Sr strontium 

SRDB Site Research Database

TAN Test Area North

TBD technical basis document

TRA Test Reactor Area

U uranium 

UO2 uranium dioxide 

WG work group 
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wt% weight percent

Y yttrium

Zr zirconium 
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1.0 INTRODUCTION

This report consolidates and summarizes SC&A’s recommendations to the Advisory Board on 

Radiation and Worker Health (Advisory Board) Idaho National Laboratory (INL) Work Group 

(WG) with respect to prioritizing reactors for further analysis by the National Institute for 

Occupational Safety and Health (NIOSH) as part of two Special Exposure Cohort (SEC) 

evaluations. The report includes all reactors at the INL site that were discussed in four different 

SC&A reports (SC&A 2015a, 2015b, 2016b, 2016c), both those designated as INL reactors and 

those designated as Argonne National Laboratory-West (ANL-W) reactors. This report also 

contains SC&A’s responses to a NIOSH report (NIOSH 2016b) that commented on SC&A’s 

recommendations from two of the four (SC&A 2016b, 2016c) abovementioned SC&A reports 

and, at the WG’s request, information on occupancy and exposure potential in various reactor 

areas. 

To place this report in context, it is necessary first to discuss some background material on how 

the reactor prioritization exercise proceeded and evolved in response to WG requests and 

comments on and discussions about several SC&A and NIOSH reports. 

NIOSH issued its INL SEC Petition SEC-00219 evaluation report (NIOSH 2015a) on March 12, 

2015, and a subsequent revision on July 21, 2015 (NIOSH 2015b). NIOSH issued its ANL-W 

SEC Petition SEC-00224 evaluation report (NIOSH 2016a) on February 18, 2016. In a series of 

meetings and discussions among the INL WG, SC&A, NIOSH, and NIOSH’s technical 

contractor, Oak Ridge Associated Universities (ORAU), the WG directed SC&A to review the 

issue of dose reconstructability for both INL and ANL-W reactors (both sets of reactors and 

facilities are located within the overall INL site). 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 definition1

1 It should be noted that the INL SEC is based on internal alpha exposures only. NIOSH plans to reconstruct internal 

doses from fission and activation products at the Central Processing Plant (CPP) 1963–1974.

and that are not being held in reserve for further 

evaluation by NIOSH. Operations at INL and ANL-W 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. 

A primary tool NIOSH uses for internal dose reconstruction is the guidance in ORAUT-OTIB0054, Fission and Activation Product Assignment for Internal Dose-Related Gross Beta and 

Gross Gamma Analyses, Revision 04, dated August 27, 2015 (ORAUT 2015; hereafter referred 

to as “OTIB-0054”). 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. SCA-SEC-2015-

0074-C, NIOSH SEC-00219: Test Reactor Area Modeling (SC&A 2015a), evaluated whether

OTIB-0054 is applicable to the three large materials-testing reactors located in the INL Test 

Reactor Area (TRA): the Materials Test Reactor (MTR), the Engineering Test Reactor (ETR), 

and the Advanced Test Reactor (ATR). SCA-TR-2015-SEC0074A, Review of NIOSH Strategy 

for Reconstructing Internal Doses to Workers at Test Area North (SC&A 2015b), similarly 

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

(HTRE) reactors located in the INL Test Area North (TAN) that supported the Aircraft Nuclear 

Propulsion program.2 

2 Discussions of the TRA and TAN reactor prioritization recommendations appear in Section 4 of this report. 

At the November 10, 2015, INL WG meeting, the Advisory Board members directed SC&A to 

screen INL reactors other than the six already addressed and create a prioritized list of reactors 

for detailed examination later with respect to OTIB-0054 applicability. Similarly, at its 

March 23–24, 2016, meeting, the Advisory Board directed SC&A to screen the ANL-W reactors. 

SC&A responded to these requests in its INL reactor prioritization report (SC&A 2016a) of 

March 2, 2016, and Revision 1 (SC&A 2016b) of June 10, 2016, that responded to some 

Advisory Board comments, and in its ANL-W reactor prioritization report (SC&A 2016c) of 

July 13, 2016. 

In prioritizing reactors for further investigation, SC&A focused on the degree to which the 

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

and Sr-90 are thought to bear any resemblance to the mix of radionuclides in OTIB-0054. In

addition to OTIB-0054 applicability, the report also considers the following four factors (to the 

extent that they are known) 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 

operation 

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

unprotected exposures

The priority rankings of the INL and ANL-W reactor prioritization reports were 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 were still somewhat subjective because a full analysis would involve detailed and 

extensive research for each reactor and performing the OTIB-0054 applicability analyses 

themselves, which would be counter to the limited objectives of the 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 [Al] or steel), moderator 

(e.g., “light” water [H2O], “heavy” water [D2O], or beryllium), and coolant (e.g., H2O, nitrogen 

gas, or organic liquid); operational mode (e.g., steady-state, periodic, or pulsed); 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.4

4 Information to address exposure potential was not readily available at the time the reactor prioritization reports 

were written but is included here in Appendix A. 

These screening criteria were selected because SC&A judged them to be those that would best 

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

After excluding reactors that either were located in ANL-W (12) or the Naval Reactors Facility 

(4), were never operated (2), or the TRA and HTRE reactors that were previously studied in 

SC&A 2015a and 2015b (6), the initial list of 52 reactors on the site was reduced to 28 for INL. 

Of those remaining, the SC&A INL reactor prioritization report (SC&A 2016b) categorized 

seven reactors in the “High Priority” class. Table 1 (adapted from Table A.1 of SC&A 2016b) 

lists the reactors and provides summary information.

Table 1. INL Reactors: High Priority Rankings 

Reactor Name(a) Operation 

Dates Summary Description

29. Loss of Fluid Test 

Facility (LOFT)

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

U.S. Nuclear Regulatory Commission (NRC), made major contributions 

to the light-water reactor (LWR) safety program for commercial 

nuclear power plants by simulating system behaviors during a loss-ofcoolant accident (LOCA) up to a worst-case, double-ended break in 

the primary coolant system. The reactor had a maximum power of 

50 megawatts thermal (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.

35. Organic Moderated 

Reactor Experiment 

(OMRE)

1957–1963 The OMRE reactor 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 

a civilian nuclear power industry. OMRE used a waxy liquid 

hydrocarbon 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 organic coolant was thought to have some advantages over 

“conventional” coolants because it allows low-pressure operation, 

solidifies at low temperatures, and does not corrode metals. 

36. Power Burst Facility 

(PBF)

1972–1985 The PBF continued the reactor safety program begun with the Special 

Power Excursion Reactor Test (SPERT) series of facilities but was much 

larger than the SPERT reactors. Fuel and cladding combinations were 

varied and tested to failure. Transient testing, including LOCA 

scenarios that modeled design-basis and severe accident conditions at 

a commercial nuclear power plant, led to fuel and cladding damage 

accompanied by the 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. The PBF could produce very high, short-duration 

(millisecond [ms]) 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. 
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Reactor Name(a) Operation 

Dates Summary Description

39. Special Power 

Excursion Reactor Test 

No. 1 (SPERT-I)

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

to large, rapid reactivity excursions 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. 

SPERT-I, the first reactor of the program, was an open-tank, lightwater-moderated and reflected reactor, with the uranium fuel 

enriched up to 93.5%. Some experiments were also conducted with 

fuel enriched only a few percent to better simulate power reactor 

fuel. The fuel consisted of plate type uranium and aluminum fuel 

assemblies 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% UO2 fuel rods), and a deliberate 35,000 MWth excursion on April 

14, 1964 (with 4% UO2 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.

40. Special Power 

Excursion Reactor Test 

No. 2 (SPERT-II)

1960–1964 SPERT-II 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 pressure vessel and 

the coolant system was pressurized. Each fuel plate contained a 

93.5% enriched uranium-aluminum alloy and was clad in aluminum. 

Because the reactor operated in the transient, burst mode, with 

power excursions up to 20 megawatt (MW)-sec, total burnup was 

small.
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Reactor Name(a) Operation 

Dates Summary Description

41. Special Power 

Excursion Reactor Test 

No. 3 (SPERT-III)

1958–1968 SPERT-III accommodated the widest variation in several important 

parameters, such as temperature, pressure, and coolant flow. The 

core sat in a pressure vessel similar to that used in a commercial 

nuclear power plant, and the maximum operating temperature of 

668 degrees Fahrenheit (°F) and pressure of 2,500 pounds per square 

inch gage (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. 

The fuel plates contained 4.8% enriched UO2 clad in stainless steel; 

the reactor used ordinary water as coolant, moderator, and reflector.

42. Special Power 

Excursion Reactor Test 

No. 4 (SPERT-IV)

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.

Note: 

(a) The reactor numbering scheme is taken from Stacy 2000. 

SC&A’s ANL-W reactor prioritization report (SC&A 2016c) analyzed the 12 ANL-W reactors 

in a similar fashion as SC&A 2016b analyzed the 28 INL reactors and categorized seven reactors 

in the High Priority class. Table 2 (adapted from Attachment 1 of SC&A 2016c) lists the reactors 

and provides summary information. 
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Table 2. ANL-W Reactors: High Priority Rankings 

Reactor Name(a) Operation 

Dates Summary Description

6. Boiling Water Reactor 

Experiment No. 1 

(BORAX-I) 

1953–1954 The BORAX series of reactor experiments tested the feasibility and 

safety and explored the operating parameters of direct steam 

production in an LWR. The chronology of the BORAX reactors is: 

BORAX-I: 7/1953–7/1954 (no surrounding structure)

BORAX-II: 10/1954–3/1955 (new facility)

BORAX-III: 7/1955–4/1956 (modification of B-II)

BORAX-IV: 12/1956–6/1958 (modification of B-III)

BORAX-V: 3/1962–8/1964 (new facility). 

BORAX-I: The 1.4 MWth reactor was water moderated and relied on 

natural water cooling and steam formation to remove heat. It was 

housed in a tank open to the atmosphere, which resulted in the 

requirement of a 1⁄2-mile exclusion zone when operating and limited 

operations to the warmer months of the year. BORAX-I was a 

precursor to the stationary low-power reactor (SL-1) plant at INL.

The core was built up from a lower grid and consisted of 26 curved 

plate fuel assemblies of a uranium-235 (U-235)-aluminum alloy clad 

with aluminum. Each assembly contained 18 fuel plates joined to 

aluminum side plates. Operators initially conducted a series of 

nondestructive experiments consisting of steady-state boiling, as well 

as over 70 excursion tests of <25 ms duration, until the reactor was 

deliberately destroyed in its final experiment on July 22, 1954, 

through the rapid withdrawal of its control rods. Calculations done 

prior to the test seriously underestimated the actual damage, where 

much of the core melted and blew pieces of fuel plates, etc. 200–

300 ft away, in a large steam explosion and resulting “geyser.”

7. Boiling Water Reactor 

Experiment No. 2 

(BORAX-II) 

1954–1955 BORAX-II, at 6.4 MWth, was contained in a stainless steel pressure 

vessel located below ground in a concrete-lined pit and operated with 

varying enrichments of uranium in its MTR-type fuel plates. It was 

intentionally destroyed in 1955 by taking it prompt critical. It was 

much larger than BORAX-I and designed to be more representative of 

a future commercial boiling-water reactor; e.g., it operated at a 

design pressure of 300 psig. It released excess energy as steam, 

because it had no turbine generator attached. The reactor was water 

cooled (natural circulation), moderated, and reflected. 

8. Boiling Water Reactor 

Experiment No. 3 

(BORAX-III) 

1955–1956 The BORAX-II reactor was modified (raising its output to 12 MWth) 

and renamed BORAX-III after the addition of a 2.5 megawatt electric 

(MWe) steam turbine-generator to investigate radioactive 

contamination of the turbine from radioactivity in the primary coolant 

and to demonstrate production of electricity. The fuel consisted of 

plate-type fuel elements of a 90% enriched uranium-aluminum alloy 

with aluminum cladding. By the time it went out of service in 1956, 

BORAX-II had operated for a total of 1,170 hrs.
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Reactor Name(a) Operation 

Dates Summary Description

9. Boiling Water Reactor 

Experiment No. 4 

(BORAX-IV) 

1956–1958 BORAX-IV, a modification of BORAX-III, at 20 MWth (and 2.5 MWe) 

and 300 psig primary coolant pressure, tested uranium (U-233 and 

U-235) and thorium ceramic fuel plates (to allow higher temperature 

operations than with uranium fuel plates), some of which 

purposefully contained defects to determine reactor behavior with 

compromised fuel. The tests released some short-lived radionuclides 

to the atmosphere. Operating at full power with a large number of 

fuel elements having cladding defects, it released approximately 

4,565 curies of short-lived radionuclides to the atmosphere in 

March 1958.

10. Boiling Water 

Reactor Experiment 

No. 5 (BORAX-V) 

1962–1964 BORAX-V, a new facility with basically the same configuration as 

BORAX-IV, but with an integral nuclear superheat system, operated at 

40 MWth. The reactor was located in a cylindrical carbon steel 

pressure vessel with ellipsoidal heads that was clad with stainless 

steel in the interior. The reactor was water moderated and water 

cooled, with the superheater section cooled by steam. Cooling could 

be accomplished by natural or forced convection.

17. Experimental 

Breeder Reactor No. I 

(EBR-I)

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

built at INL. It had a maximum power of 1 MWth from its graphitereflected, unmoderated, enriched uranium core. The reactor system 

was cooled by the liquid metal (NaK) (primary and secondary coolant 

systems) and surrounded by a U-238 breeding blanket. In addition to 

producing the first electricity from a nuclear plant, EBR-I also 

demonstrated the feasibility of fuel production (breeding). In fact, at 

full power of 1 MWth, 16% of the total power was generated in the 

blanket. Breeding ratios increased from 1.00 in the first core to 1.27 in 

the last core. The reactor also had several beam holes for 

instrumentation and irradiation experiments. The first three of the 

EBR-I’s four core loadings had highly enriched (94% U-235) uranium 

fuel clad either with stainless steel or aluminum. On November 29, 

1955, the reactor suffered a 40% to 50% inadvertent core meltdown 

due to operator error with control rods during an experiment. 

18. Experimental 

Breeder Reactor No. II 

(EBR-II)

1961–1994 The EBR-II, similar to, but a large scale-up from, EBR-I, continued fast 

neutron breeder reactor development at ANL-W, including onsite 

reprocessing of spent fuel into new fuel pins, demonstrating the 

feasibility of a closed fuel cycle. The unmoderated core, with 67% 

enriched U-235 fuel, sat in a tank of 90,000 gallons of liquid sodium 

(Na) primary coolant, had a closed-loop Na secondary coolant system, 

and produced steam in a tertiary system. The entire system was 

placed in a large containment building. The maximum power level was 

62.5 MWth, and the EBR-II could supply 20 MWe of electric power to 

INL facilities. In addition to demonstrating fuel reprocessing and 

electricity production in a liquid metal fast breeder reactor, EBR-II also 

performed irradiation, fuel development, and transient stability 

experiments.

Note: 

(a) The reactor numbering scheme is taken from Stacy 2000. 
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2.0 NIOSH PROPOSAL

NIOSH responded to SC&A’s INL and ANL-W reactor prioritization reports (SC&A 2016b, 

2016c) with its NIOSH Proposal for INL and ANL-W Reactor Prioritization for OTIB-0054 

Evaluation, July 28, 2016 (NIOSH 2016b). The report reviews the SC&A high priority reactor 

lists and makes eight recommendations that inform how NIOSH would proceed with further 

reactor modeling work to determine how well OTIB-0054 envelopes the INL and ANL-W 

reactors. Table 3 reproduces NIOSH’s proposal conclusions. 

Table 3. NIOSH Proposal Conclusions 

No.(a) NIOSH Recommendation(b)

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.

2 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.

3 NIOSH agrees that the Organic Moderated Reactor Experiment (OMRE) should be evaluated for OTIB-0054 

applicability due to its unique moderator and coolant.

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

use of ceramic fuel.

5 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.

6 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 uraniumthorium oxide fuel. NIOSH proposes that BORAX-V be removed from consideration for evaluation of OTIB0054 applicability since its primary function was to evaluate steam superheating with essentially the same 

configuration as BORAX-IV.

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, It is not clear to SC&A that the Mark II core material was ever handled outside the 

SEC period, which would necessitate modeling it. 

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

Source: NIOSH 2016b.

Notes: 

(a) Numbering added here reflects the order in which each point appeared in the original bulleted list in NIOSH 

2016b.

(b) Recommendations copied from the NIOSH 2016b Conclusions section.

The NIOSH report ends with a recommendation for the six reactors (or series of reactors) to 

model with detailed analyses using the ORIGEN (Croff 1980) and SCALE (ORNL 2015) system 

of isotopic buildup and decay codes. Those reactors and the NIOSH-proposed groupings appear 

here in Table 4.
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Table 4. Reactors for NIOSH Proposed Grouping for OTIB-0054 Applicability Evaluation 

Reactor Numbers 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)
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3.0 SC&A RESPONSE TO THE NIOSH PROPOSAL

NIOSH presented its response (NIOSH 2016b) to the two SC&A reactor prioritization studies at 

the August 2, 2016, meeting of the INL/ANL-W WG. The WG members then directed SC&A to 

comment on the NIOSH proposal with respect to OTIB-0054 adequacy considerations, as well as 

the potential for personnel exposure at each of the six reactors. Table 5 presents SC&A’s 

assessment of NIOSH’s proposal, and Appendix A of this report presents SC&A’s detailed 

findings on the potential for personnel exposure at the six reactors. 

As Appendix A shows, the prioritized reactor sites generally employed hundreds of monitored 

workers, except for the PBF. This facility only appears to have assigned 30 workers during most 

badging cycles in the available records. Penetrating doses to workers at the prioritized reactor 

sites were also significant, with some monthly badging cycles averaging on the order of 

hundreds of millirem (mrem) (see Sections A.2 and A.4 on EBR-II and SPERT, respectively). 

While external exposure rates are not necessarily directly correlative to internal exposure 

potential, the magnitude of these accrued 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.

Table 5. SC&A Responses to NIOSH Proposal 

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 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.

SC&A recognizes that the first five LOFT experiments 

were non-nuclear, 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-designbasis 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 Organic Moderated 

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

4 NIOSH agrees that the Power Burst Facility 

(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 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.

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.”

6 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.

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. 

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 the Experimental Breeder 

Reactor No. II should be evaluated for OTIB0054 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. 

Notes: 

(a) Numbering added here reflects the order in which each point appeared in the original bulleted list in NIOSH 

2016b.

(b) Recommendations copied from the NIOSH 2016b Conclusions section.
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Note that SC&A’s recommendations for further analysis for TAN and TRA reactors are 

discussed in Section 4. 
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4.0 TEST AREA NORTH AND TEST REACTOR AREA REACTORS

SC&A verbally presented its draft results, summarized in Table 5 of this report, to an INL WG 

teleconference on September 8, 2016. At the recommendation of NIOSH and at the direction of 

the WG, SC&A includes in this section its observations and recommendations pertaining to the 

TRA and TAN reactors that were the subject of investigations in SC&A’s September 28, 2015, 

review of NIOSH internal dose reconstruction at TRA (SC&A 2015a) and SC&A’s 

September 28, 2015, TAN evaluation report (SC&A 2015b).5

5 Note that while NIOSH commented on the two SC&A reactor prioritization reports (SC&A 2016b, 2016c) in the 

proposal report (NIOSH 2016b), it never commented on the SC&A TRA (SC&A 2015a) and TAN (SC&A 2015b) 

reactor reports. 

 Hence, this document now 

contains discussions of all the INL and ANL-W reactors of interest in the two SEC evaluations. 

4.1 TAN REACTORS AND ASSOCIATED IRRADIATED AND SPENT FUEL

One of the topics addressed in SC&A’s September 28, 2015, TAN evaluation report (SC&A 

2015b) is the applicability of OTIB-0054 and Tables 5-22 and 5-23 of the INL technical basis 

document (TBD) (NIOSH 2010) 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 

HTREs 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 during burnup. 

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, the HTRE spent and irradiated fuel 

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. 

To provide additional information to address these potential concerns, preliminary analyses 

reported in SC&A 2015b included generic ORIGEN-ARP6

 runs that SC&A had made, 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.

7

6 ORIGEN-ARP is a version of the ORIGEN code contained in the SCALE system of nuclear codes (ORNL 2015). 7 It is understood that power levels for the TAN reactors refer to “thermal power,” since none of the reactors 

produced electricity. 

These 

scenarios were intended to provide a general representation of the operating conditions of the 

HTRE tests at INL (e.g., HTRE-1 was operated at 20 MW for 150.8 hours). The results of the 

ORIGEN runs were used to help assess the suitability of using OTIB-0054 reactor configurations 

and operating conditions to represent the TAN reactor operations. 

The first set of calculations assumed a high power level (200 MW) and derived the mix of fission 

products in the fuel relative to Sr-90 after a 10-day cool-down period, which was assumed

instead of the longer cool-down periods in Table 7-3a of OTIB-0054 to minimize the relative 

importance of the longer-lived fission products. Only the ATR1 case of Table 7-3a was 

investigated, because SC&A believes the results would also apply (well enough to inform the 

 
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investigation) to the other reactor types in OTIB-0054 in the preliminary analysis that was 

performed. The second set of calculations assumed a low power level (20 MW). 

For each of the two assumed power levels, SC&A ran ORIGEN for different fuel burnup times. 

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. The SC&A analysis 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. The 

detailed results of these calculations are provided in Tables 1 and 2 of SC&A 2015b. 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 comparable to, 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.

Given the complexity of this subject, SC&A took additional steps in this report to evaluate the 

degree to which OTIB-0054 is, in fact, applicable or claimant favorable to workers at TAN 

facilities, keeping in mind that OTIB-0054 is used to reconstruct the intake rates of individual 

radionuclides by TAN workers who were routinely monitored on a gross beta/gamma urine 

bioassay program. The results of this additional investigation appear in Appendix B. 

For actinide activation products (the results are in Tables 3 and 4 of SC&A 2015b), SC&A found 

that, 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 Tables 5-22 and 5-23 of the TBD 

applied to reconstructing internal doses to workers at TAN 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 

Tables 5-22 and 5-23 of the TBD.8

8 It should be noted that a similar situation might also apply to other reactors in the INL complex. 

Although SC&A was unable to locate specific information concerning workforce and exposure 

potential for HTRE personnel during the years of its operation, SC&A compiled data for the 

TAN area in the years immediately following the operation of the HTRE. This information is 

described in Section A.6 of Appendix A. For the years for which information was available, 

there were generally 250–300 regular badges and 50–150 “other” badges issued in the TAN area 

 
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per badging cycle. The maximum penetrating dose for TAN was generally between 200 and 800 

mrem per badging cycle, with the highest value (1.15 rem) during a 2-week period in 1965. 

4.2 TRA REACTORS

SC&A’s September 28, 2015, review of NIOSH internal dose reconstruction at TRA (SC&A 

2015a) examined the three major TRA reactors and their operations to determine if they are 

adequately enveloped by OTIB-0054 methodology and data: the MTR (1952–1970), the ETR 

(1957–1981), and the 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 40 MW for the MTR, 175 MW for the ETR, and 250 MW for the ATR.9

9 It is understood that power levels for the TRA reactors refer to “thermal power,” since none of the reactors 

produced electricity.

Their 

high-flux capability allows for the accelerated simulation of long-term irradiation of reactor 

materials. Each is a pressurized, light-water-moderated, beryllium-reflected reactor, primarily 

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 (a very effective thermal neutron absorber) shells. 

SC&A 2015a found that OTIB-0054 adequately enveloped the three reactors (the OTIB 

explicitly models the ATR) for uranium fuel operations but noted that the MTR also ran for a 

period with plutonium fuel. Although the MTR was initially intended to use uranium fuel, in 

1958 it became the first reactor run with a plutonium-239 (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 weight percent (wt%) Pu and 79 wt% Al. The reactor reached initial power on 

January 28, 1970, and operated at power levels up to 24 MW until April 23, 1970, at which time 

it had accumulated a burnup of 923 megawatt days. This compared to the standard U-235 core, 

which operated at 40 MW. 

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 (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 the SC&A 2015a report. 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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Another “unusual” operating scenario for the MTR was the radioactive lanthanum (RaLa) 

extraction campaign from 1956 through 1963. 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 because, 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).” 

SC&A compiled available data on the workforce and exposure potential in the TRA area, 

summarized in Section A.7 of Appendix A. In general, the TRA area issued approximately 1,000 

regular badges per exchange period from mid-1958 through mid-1967, with spikes greater than 

4,000 badges for some periods. In addition, most badging periods also issued several hundred 

“other” badges, which are issued to visitors or non-routine workers. Construction workers were 

issued separate dosimeters aside from the “other” category. In general, the number of 

construction dosimeters issued ranged between 100 and 200 per badging period. 

Maximum penetrating exposures were approximately 250 mrem for weekly badging exchanges 

(1956–1958), 500–1,000 mrem for bi-weekly exchanges (1959–1969), and averaged about 1,000 

mrem for monthly badging cycles (1970–1974). Construction worker exposures were generally 

in the 100–200 mrem range. Several badging cycles for construction workers showed no accrued 

penetrating external dose.
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5.0 REFERENCES

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

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

Tennessee. July 1980. [SRDB Ref. ID 23569] 

EPA 1999. Federal Guidance Report No. 13: Cancer Risk Coefficients for Environmental 

Exposures to Radionuclides, EPA 402-R-99-001, U.S. Environmental Protection Agency, 

Washington, DC. September 1999. 

NIOSH 2010. Idaho National Laboratory and Argonne National Laboratory-West – 

Occupational Internal Dose, ORAUT-TKBS-0007-5, Revision 3, National Institute for

Occupational Safety and Health, Cincinnati, Ohio. March 2, 2010. 

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

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

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

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

NIOSH 2016a. SEC Petition Evaluation Report: Petition SEC-00224, Revision 0, National 

Institute for Occupational Safety and Health, Cincinnati, Ohio. February 18, 2016. 

NIOSH 2016b. NIOSH Proposal for INL and ANL-W Reactor Prioritization for OTIB-0054 

Evaluation, National Institute for Occupational Safety and Health, Cincinnati, Ohio. July 28, 

2016. 

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

and Gamma Analysis, ORAUT-OTIB-0054, Revision 4, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee. August 27, 2015. 

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

Analysis and Design, Oak Ridge National Laboratory, Oak Ridge, Tennessee. Accessed 

November 18, 2016, at http://scale.ornl.gov. 

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

Revision 0, SC&A, Inc., Vienna, Virginia, and Saliant, Inc., Jefferson, Maryland. 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., Vienna, Virginia, and Saliant, 

Inc., Jefferson, Maryland. September 28, 2015. 

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

Applicability, SCA-TR-2016-SEC002, Revision 0, SC&A, Inc., Vienna, Virginia, and Saliant, 

Inc., Jefferson, Maryland. March 2, 2016. 
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SC&A 2016b. INL SEC-00219 Reactor Prioritization for Evaluation of ORAUT-OTIB-0054 

Applicability, SCA-TR-2016-SEC002, Revision 1, SC&A, Inc., Vienna, Virginia, and Saliant, 

Inc., Jefferson, Maryland. June 10, 2016. 

SC&A 2016c. Argonne National Laboratory-West SEC-00224 Reactor Prioritization for 

Evaluation of ORAUT-OTIB-0054 Applicability, SCA-TR-2016-SEC010, Revision 0, SC&A, 

Inc., Vienna, Virginia, and Saliant, Inc., Jefferson, Maryland. July 13, 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, Idaho. 2000. [SRDB Ref. IDs 1152] 
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APPENDIX A: CHARACTERIZATION OF WORKFORCE AND 

EXPOSURE POTENTIAL FOR SIX REACTORS SELECTED FOR 

ANALYSIS AT INL/ANL-W 

Prepared by

Bob Barton, SC&A 

The National Institute for Occupational Safety and Health (NIOSH) presented its response 

(NIOSH 2016b) to SC&A’s reactor prioritization reports (SC&A 2016a, 2016b, 2016c) during 

the Idaho National Laboratory (INL)/Argonne National Laboratory-West (ANL-W) Work Group 

(WG) meeting on August 2, 2016, and recommended that six reactors be considered for the 

initial analyses:

(1) Experimental Breeder Reactor No. I (EBR-I) MARK IV: 1962–1963 

(2) Experimental Breeder Reactor No. II (EBR-II): 1961–1994 

(3) Power Burst Facility (PBF): 1972–1985 

(4) Boiling Water Reactor Experiment No. 4 (BORAX-IV): 1958 

(5) Special Power Excursion Reactor Tests (SPERT): 1958–1970 

(6) Organic Moderated Reactor Experiment (OMRE): 1958–1963 

At the August 2016 WG meeting, the members asked SC&A to research and compile data on the 

number and exposure potential of workers at these six reactor facilities during relevant time 

periods. Subsequently, at the September 8, 2016, meeting, the WG asked SC&A to perform 

additional data compilation for the TRA and TAN areas. SC&A conducted a thorough search of 

available documents on the Site Research Database (SRDB) for both INL and ANL-W sites. The 

documents used to compile the available data in this appendix are listed in Section A.8. It should 

be noted that individual documentation about the workforce at the BORAX-IV experiment in 

1958 could not be located and, therefore, is not discussed further in this appendix. 

SC&A encountered several different formats of Health and Safety (H&S) reports containing 

information about the number of individuals and visitors present, the number of regular versus 

non-routine badges, exposure distributions among the monitored workforce, and average 

penetrating exposures by time period. SC&A observed that the different formats often contained 

unclear and/or conflicting information. Additionally, some reports covered time periods that

overlapped the badging periods covered in other sources, which obscured the actual worker and 

dose totals. SC&A used its judgment in interpreting the differing reporting practices to obtain the 

summary totals displayed in the figures of this appendix and considers this assessment to be a 

reasonably accurate characterization of total workforce and exposure potential at each reactor 

site. 

In interpreting the reporting practices reflected in the available documentation, it is important to 

note that for the purposes of this analysis: 

• “Regular” refers to individuals who routinely worked and were monitored in a given 

area. 
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• “Visitor” refers to individuals who did not routinely work in an area but were assigned to 

the area for a given badging period. 

• “Other” refers to the category of dosimeters issued to visitors, as well as any nonroutine, health physics-specific, and/or incident-driven temporary badges issued per area. 

The distinction between “visitor” and “other” is simply that some routine (or “regular”) workers 

appear to have also been issued non-routine and/or incident badges, which are in addition to the 

routine worker totals. However, it appears the incidence of such additional badging is likely 

minimal, and so “other” is considered a fair representation of the visitors entering a reactor area.

A.1. EXPERIMENTAL BREEDER REACTOR I – MARK IV CORE (1962–1963) 

Figure A.1-1 plots the number of “regular” individuals reported in available H&S documentation 

from 1963 to 1965. SC&A was not able to find specific information for EBR-I for 1962. In 

general, there were approximately 150–200 workers during the period of operation for the 

MARK IV core, with a spike in September of 1963 in which both regular workers and visitors 

increased markedly. Although the MARK IV experimentation ended in December 1963, the 

number of regular workers and visitors was also compiled for 1964 and 1965, as there may have 

been exposure to residual materials present at EBR-I after the last experiment. 

Figure A.1-1. Number of Regular Individuals and Visitors to the EBR-I Area: 1963–1965 

0

50

100

150

200

250

300

350

400

450

Number of Individuals

Monthly Badging Cycle End Date

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Figure A.1-2 plots the average penetrating dose to regular workers at EBR-I from 1963 to 1965. 

For the majority of badging periods in which the average penetrating dose was reported (17 of 22 

periods), the average worker dose at EBR-I was 15 millirem (mrem) or less. The highest average 

worker doses were observed in March and September 1963 (26 and 27 mrem, respectively). It is 

worth noting that a spike in the total number of regular workers and visitors was observed in 

September 1963. No reports of the maximum external exposure to specific individuals per 

badging period was available in EBR-I documentation. 

Figure A.1-2. Average Penetrating Dose to Regular Workers at EBR-I: 1963–1965 

A.2. EXPERIMENTAL BREEDER REACTOR II (1961–1994) 

Figure A.2-1 shows the number of routine workers at EBR-II from 1963 through 1975.10

10 SC&A was unable to locate additional information for EBR-II after July 1975. 

From 

1963 until 1966, the regular work force at EBR-II was in the range of about 400 workers. For the 

remaining badging periods with records after 1966, documentation generally reported regular 

workers in the 200-to-300 range for two out of every three months. Interestingly, the number of 

regular workers reported for EBR-II approximately doubled every third month beginning in 

1967. The apparent gap from mid-1973 to mid-1974 is not the result of a lack of records. During 

this gap, the available records had “cut off” the listing of actual work locations in the scanned 

electronic versions of the hardcopy file. 

Figure A.2-2 plots the number of reported visitors to EBR-II from 1963 to 1972. Unlike the 

regular worker totals presented in Figure A.2-1, the available badging reports did not indicate the 

 

0

5

10

15

20

25

30

Average Regular Worker Dose (mrem)

Monthly Badging Cycle End Date
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number of visitors after June 1972. It is apparent that visitor totals could vary significantly 

depending on the timeframe. Totals ranged from a low of 32 visitors in May 1965 to a high of 

691 visitors in June 1968. In general, the number of visitors increased somewhat during 1968 

and 1969. 

Figure A.2-1. Number of Regular Individuals at the EBR-II Area: 1963–1975 

0

100

200

300

400

500

600

700

800

1963 1964 1965 1966 1967 1968 1969 1970 1971 1972 1973 1974 1975 1976

Number of Regular Individuals

Monthly Badging Cycle End Date
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Figure A.2-2. Number of Visitors to the EBR-II Area: 1963–1972 

Figure A.2-3 displays the average worker exposure per monthly badging cycle from 1963 to

1971. As a general trend, it appears that average worker doses increased somewhat during this 

timeframe. The highest observed average penetrating dose was approximately 122 mrem in 

February 1967, with a similarly high value observed in November 1969. The average over the 

entire period was approximately 25 mrem per worker per monthly badging period. 

Monthly badging reports often provided information on how many regular workers fell into 

certain dose ranges (0–299, 300–1,199, and 1,200+ mrem). SC&A did not observe any reported 

workers falling into the highest dose category (1,200+ mrem); however, some workers did fall 

into the intermediate category for some months. Figure A.2-4 displays the percentage of regular 

workers who had monthly penetrating doses that were between 300 and 1,199 mrem. The highest 

observed percentage of workers falling into this dose range was approximately 13% in 

November 1969. Not surprisingly, many of the “spikes” observed in in Figure A.2-4 correspond 

to the higher monthly averages shown in Figure A.2-3. Figure A.2-5 overlays the average worker 

doses from Figure A.2-3 with the percentage of workers receiving doses greater than 300 mrem 

in Figure A.2-4. As expected, the two figures show significant temporal agreement. 

0

100

200

300

400

500

600

700

1963 1964 1965 1966 1967 1968 1969 1970 1971 1972 1973

Number of Visitors

Monthly Badging Cycle End Date
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Figure A.2-3. Average Worker Penetrating Dose in the EBR-II Area per Monthly Badging 

Cycle: 1963–1971 

Figure A.2-4. Percentage of Workers at EBR-II Exceeding 300 mrem per Monthly Badging 

Cycle: 1963–1971 

0

10

20

30

40

50

60

70

80

90

100

110

120

130

Average Regular Worker Penetrating Dose 

(mrem)

Monthly Badging Cycle End Date

0%

2%

4%

6%

8%

10%

12%

14%

Percent of Regular Workers with Monthly 

Penetrating Doses Exeeding 300 mrem

Monthly Badging Cycle End Date
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Figure A.2-5. Overlay of Figures A.2-3 and A.2-4 Demonstrating the Correlation between 

Average Observed Penetrating Doses and the Number of Workers with Doses Greater than 

300 mrem 

A.3. POWER BURST FACILITY (1972–1985) 

SC&A could not locate any badging reports after 1975, so the analysis of regular workforce and 

exposure potential was limited to the first few years of operation. In contrast to EBR-I and 

EBR-II, the PBF had significantly fewer regular workers during the early years of its operation. 

As seen in Figure A.3-1, the number of regular workers assigned to PBF during these years was 

between 25 and 30 for most periods for which records were available. The maximum observed 

number of workers during this period was 77 in January 1975. It is worth noting that the large 

observed gap from May 1973 to April 1974 was also observed for EBR-II (see Figure A.1-1) and 

was the result of the specific area designation being cut off in scanned version of the badging 

reports for those months (i.e., an artifact of the record-keeping system, rather than an actual 

occurrence). 

Interestingly, the number of visitors often outnumbered the number of regular workers, as shown 

in Figure A.3-2. The greatest number of visitors at PBF was 140 in February 1973. Note that the 

absence of a visitor total in Figure A.3-2 is not indicative of there being no visitors to PBF for a 

particular badging cycle, but rather the lack of available records indicating the actual visitor 

totals. 

0%

2%

4%

6%

8%

10%

12%

14%

0

10

20

30

40

50

60

70

80

90

100

110

120

130

Average Worker Penetrating Dose (mrem)

Percentage of Workers >300 mrem in Cycle

Monthly Badging Cycle End Date

Average Regular Worker Dose (mrem) % Regular Workers >300 mrem
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Figure A.3-1. Number of Regular Individuals at PBF per Monthly Badging Cycle 

0

10

20

30

40

50

60

70

80

Number of Regular Individuals

Monthly Badging Cycle End Date
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Figure A.3-2. Comparison of Regular Worker Totals to Visitors at PBF per Monthly 

Badging Cycle 

Only 7 of the 46 badging cycle entries identified among available documentation also provided 

the maximum and/or average penetrating dose to regular workers at PBF. Nonetheless, 

Figure A.3-3 plots these average and maximum penetrating doses by monthly badging period. 

Average worker doses ranged from 2 to 92 mrem, while the maximum observed monthly dose 

was 460 mrem. 

0

25

50

75

100

125

150

Number of Workers

Monthly Badging Cycle End Date

Regular Individuals Visitors
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Figure A.3-3. Comparison of Average Doses to Maximum Doses for Regular Workers at 

PBF

A.4. SPECIAL POWER EXCURSION REACTOR TESTS (1958–1970) 

Available records for SPERT do not indicate a total number of regular workers or visitors at the 

area until 1963. However, the total number of regular (or routine) badges versus non-routine 

badges issued is a reliable indicator of the approximate number of workers potentially exposed at 

SPERT during a given badging cycle. Figure A.4-1 plots the number of regular and non-routine 

badges reported for SPERT from 1958 to 1962. 

Beginning in 1963, the H&S reports provided the actual number of regular workers versus 

visitors. These totals are presented in Figure A.4-2. As seen in the figure, the number of regular 

workers generally decreased from 1963 through 1966, at which point the number fluctuates 

between 20–30 workers for some periods and over 150 workers for others. The number of 

visitors also shows similar fluctuations during the period after 1966, ranging from around 100 

visitors to over 300. 

0

50

100

150

200

250

300

350

400

450

500

PENETRATING DOSE (MREM)

MONTHLY BADGING CYCLE END DATE

Average Dose for Regular

Workers

Maximum Dose for

Regular Workers
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Figure A.4-1. Number of Routine and Non-Routine Dosimeters Issued for SPERT: 1958–

1962 

0

50

100

150

200

250

Number of Badges

Bi-Weekly Badging Cycle End Date

Regular Badges Other Badges
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Figure A.4-2. Number of Regular Workers and Visitors at SPERT by Monthly Badging 

Period

Figure A.4-3 shows the maximum reported dose by monitoring period for workers at SPERT 

from 1958 through 1971. It should be noted that until 1962, SPERT workers were on a biweekly 

badging exchange schedule, after which they switched to a monthly schedule. As shown in the 

chart, one uncharacteristically high value was observed in the beginning of 1959 (2,120 mrem). 

Inspection of H&S reports subsequent to this badging period indicates that this exposure was 

likely related to exposure incurred by the worker while at the Chemical Processing Plant (CPP). 

Figure A.4-4 plots the maximum reported penetrating dose without the outlier from CPP 

included. As shown, the maximum exposures to SPERT workers varied significantly by badging 

period. Many of the biweekly badging results ranged from 100 to 200 mrem, with spikes 

exceeding 400 mrem. After 1963, the dose totals generally increased per badging period, which 

is to be expected when switching from biweekly to monthly badging periods. The greatest 

observed penetrating dose during the biweekly period was 940 mrem in 1960, while during the 

monthly badging periods, the highest observed dose was 820 mrem in 1971. 

0

50

100

150

200

250

300

350

Number of Individuals

Monthy Cycle End Date

Regular Individuals Visitors
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Figure A.4-3. Maximum Penetrating Dose by Badging Cycle for Regular Workers at 

SPERT (All Data)

0

500

1000

1500

2000

2500

Badging Cycle End Date

Maximum Penetrating Dose per Badge Cycle (mrem)

Maximum Value Likely Related to CPP Exposure
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Figure A.4-4. Maximum Penetrating Dose by Badging Cycle for Regular Workers at 

SPERT (Outlier Removed)

Badging Cycle End Date

Like the reporting of the number of individuals and visitors in available H&S reports, average 

worker doses were not reported until 1963, when SPERT appears to have switched to a monthly 

dosimeter exchange cycle. Figure A.4-5 plots the average penetrating dose to workers from 1963 

to 1971. Until 1967, the average dose to regular workers was generally less than 20 mrem. After 

this time, average doses generally increased, with marked spikes in August 1968 (162 mrem), 

February 1970 (180 mrem), and September 1971 (172 mrem). 

0

100

200

300

400

500

600

700

800

900

Maximum Penetrating Dose (mrem)

1000
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Figure A.4-5. Average Regular Worker Dose per Monthly Badging Cycle: 1963–1971 

A.5. ORGANIC MODERATED REACTOR EXPERIMENT (1958–1963) 

Documentation concerning the workforce and exposure potential specific to the OMRE project 

was only located for 1963 (the last year of operation). The total number of regular workers, as 

well as the number of visitors to the OMRE area, are shown in Figure A.5-1. Because data were

only available for the last year of experimental operation, it is not surprising that the number of 

regular workers decreased consistently from the start of 1963 through October 1963. OMRE 

appeared on H&S reports after October 1963; however, these reports simply listed the number of 

regular workers and visitors as zero. 

0

20

40

60

80

100

120

140

160

180

200

Monthly Badging Cycle End Date

Average Penetrating Dose per Worker (mrem)
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Figure A.5-1. Number of Regular Workers and Visitors to the OMRE Area by Monthly 

Badging Cycle in 1963 

Figure A.2-2 plots the average doses to regular workers by monthly badging cycle. The data are 

limited to just five monthly badging cycles in 1963 and ranged from an average penetrating dose 

of 2 mrem to 38 mrem. Interestingly, the highest observed average value occurred in the last 

month for which the H&S department reported workers for OMRE (five regular workers and six

visitors). 

0

20

40

60

80

100

120

140

160

180

Number of Individuals

200

1/1963 2/1963 3/1963

Monthly Badging Cycle End Date

4/1963 5/1963 6/1963 7/1963 8/1963 9/1963 10/1963

Regular Individuals Visitors
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Figure A.5-2. Average Penetrating Dose to Regular Workers at the OMRE Site per 

Monthly Badging Cycle in 1963 

A.6. TEST AREA NORTH REACTOR AREA

As noted in Section 4.1, SC&A was unable to locate specific information about the HTRE during 

its operational period (approximately 1955–1961). Nonetheless, SC&A compiled available data 

for the TAN area beginning in subsequent years when records became available. Figure A.6-1 

plots the number of regular and “other” badges for the TAN area beginning in November 1962. 

As seen in the figure, the number of badges was relatively low for the first few badging cycles 

available before increasing to approximately 300–400 regular badges per cycle. It should be 

noted that all badges were issued on a 2-week exchange schedule until late February 1965. 

0

5

10

15

20

25

30

35

40

1/1963 2/1963 3/1963

Monthly Badging Cycle End Date

4/1963 5/1963 6/1963 7/1963 8/1963 9/1963 10/1963

Average Regular Worker Penetrating Dose (mrem)
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Figure A.6-1. Number of Regular and “Other” Badges Issued at TAN Beginning in 

November 1962 for Workers on 2- and 4-Week Exchange Schedules

1964 1965 1966

Date

In addition to statistics concerning the TAN area as a whole, H&S reports also contained badging 

totals for the Initial Engine Test (IET) area of TAN where the HTREs were located. These totals 

are shown in Figure A.6-2. It should be noted that while the HTREs were located at the IET area, 

data were only available starting in the fourth quarter of 1963, when one of the main activities at 

IET was the Systems for Nuclear Auxiliary Power (SNAP) program. It is not clear to what extent 

exposure potential would still have existed for material from the HTRE program.

0

50

100

150

200

1962 1963

250

300

350

400

450

Number of Badges

TAN Regular - Four Week Exchange TAN Regular - Two Week Schedule

TAN Other Badges - Four Week Schedule TAN Other - Two Week Schedule
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Figure A.6-2. Number of Regular and “Other” Badges Issued at IET Beginning in October 

1963 for Workers on 2- and 4-Week Exchange Schedules

1964 1965 1966

Date

Figure A.6-3 plots the maximum penetrating dose observed for any given badging period at both 

IET and TAN as a whole. As the figure shows, the maximum penetrating dose at TAN generally 

bounded the dose at IET. The highest observed penetrating dose was 1.15 rem during a 2-week 

badging period at TAN in August 1965. In general, the maximum 2-week penetrating exposure 

was greater than 200 mrem at TAN. 

0

20

1963

40

60

80

100

120

140

160

180

Number of Badges

IET Regular - Four Week Exchange

200

IET Regular - Two Week Schedule

IET Other Badges - Four Week Schedule IET Other - Two Week Schedule
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Figure A.6-3. Maximum Penetrating Dose (mrem) per Badging Cycle at TAN and IET for 

Workers on 2- and 4-Week Exchange Schedules

0

1962 1963

In addition to regular and “other” badges at TAN and IET, a separate category for construction 

work at TAN was used (dosimetry code Ax). The total number of Ax badges issued and the 

maximum penetrating dose are shown in Figure A.6-4. As seen in the figure, data concerning 

construction badging at TAN were sparse. For most badging cycles, only 10 or fewer badges 

were issued for TAN construction. The sole badging cycle higher than these totals was 168 

badges in December 1963. In addition, the exposure potential for TAN construction activities 

appears consistently low (often zero mrem), with only two data points that were non-zero. The 

highest penetrating dose was 50 mrem, which coincided with the highest number of badges 

observed. 

200

400

600

800

1000

1200 Maximum Penetrating Dose (mrem)

1400

1964 1965 1966

Date

IET Four Week Schedule IET Two Week Schedule

TAN Four Week Schedule TAN Two Week Schedule
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Figure A.6-4. Total Number of Ax Badges and Maximum Penetrating Dose (mrem) per 

Badging Cycle at TAN for Construction

1964

Date

A.7. TEST REACTOR AREA (1956–1981) 

Figure A.7-1 displays available records tabulating the number of badges issued in TRA

beginning in October 1956. In general, 1,000 regular badges were issued per badging period 

from 1957 to 1962. After this time, there were spikes in the number of regular badges issued, 

which ranged as high as 4,000 badges per period. The number of “other badges” generally 

mirrored the trend seen in the regular badges up through 1966. No data were available on regular 

badging for 1968 and 1969. The number of regular and “other” badges were comparable after 

1969. 

0

10

1965 1966

Total Ax Badges

Maximum

Penetrating Dose 20

30

40

50

60

Maximum Penetrating Dose (mrem)

0

20

40

60

1963

80

100

120

140

160

Number of Monthly Badges

180
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Figure A.7-1. Total Number of Regular and “Other” Badges Issued per Badging Period at 

TRA from Late 1956 through 1974 

Date

Figure A.7-2 plots the maximum observed penetrating dose (in units of mrem) for TRA in 

available documentation. As shown in the figure, three different badging exchange schedules 

were noted based on the time period. For the earlier time periods, badges were on a weekly 

exchange schedule; this changed to a bi-weekly schedule from approximately 1959 into early 

1970. After this time, badges were exchanged on a monthly schedule. 

The highest observed penetrating dose was 11 rem during a 2-week period in January 1961. 

However, information provided in the source documentation indicates that this exposure was 

likely related to the Stationary Low-Power Reactor (SL-1) incident. Figure A.7-3 plots the 

maximum penetrating dose with this data point removed. As seen in both Figure A.7-2 and 

Figure A.7-3, the maximum penetrating dose was approximately 250 mrem during the weekly 

badging cycles, 500–1,000 mrem for bi-weekly exchanges, and 500–1,500 mrem for monthly 

exchanges.

0

500

1000

1500

2000

2500

3000

3500

4000

Number of Badges

4500

Regular Badges

Other Badges
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Figure A.7-2. Maximum Penetrating Dose (mrem) for TRA

Date

0

2000

4000

6000

8000

10000

Maximum Penetrating Dose (mrem)

Bi-Weekly Exchange

Monthly Exchange

Associated with SL-1 Incident

12000

Weekly Exchange
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Figure A.7-3. Maximum Penetrating Dose (mrem) for TRA with Suspected SL-1 Result 

Removed 

Date

0

500

1000

1500

2000

2500

3000

3500

Maximum Penetrating Dose (mrem)

4000

Weekly Exchange Bi-Weekly Exchange Monthly Exchange

4500

Figure A.7-4 plots the average penetrating dose (mrem) for TRA. It should be noted that while 

TRA workers were mainly on a bi-weekly schedule during this period, averages were most often 

reported on a monthly basis. Monthly averages from 1963 through mid-1966 gradually increased 

but were generally on the order of 50 mrem. After this time, monthly averages ranged from 50 

mrem to approximately 250 mrem. The highest observed monthly average was 389 mrem in 

February 1968. 
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Figure A.7-4. Average Penetrating Dose (mrem) Evaluated on Monthly and Bi-Weekly 

Intervals

Date

0

50

100

150

200

250

300

350

400

Average Penetrating Dose (mrem)

450

Bi-Weekly Average Monthly Average

In addition to regular and “other” badges issued to TRA personnel, construction personnel were 

tabulated under a separate dosimetry code (Mx or MTX). The number of construction badges 

issued in TRA and the maximum observed penetrating dose are shown in Figure A.7-5. As seen 

in the figure, the number of construction badges issued varied significantly based on the specific 

time period. In general, the number of construction badges could be generally characterized as 

100–150 badges per exchange cycle, with a range from a low of 16 to a high of 225. 

Likewise, the maximum penetrating dose among construction workers also varied widely 

depending on time period. Available documentation suggests that the two highest observed 

values (2.88 and 1.60 rem) were associated with the SL-1 incident. Figure A.7-6 displays the 

same data with these two data points removed. As seen in Figure A.7-6, maximum penetrating 

exposures were generally about 100–200 mrem, with only a few occasions exceeding this range 

and several badge cycles in which no external penetrating dose was accrued by construction 

personnel. 
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Figure A.7-5. Total Number of Construction Badges Issued and Maximum Observed 

Penetrating Dose (mrem)

0

1962 1963 1964

500

1000

1500

2000

2500

3000

3500

0

1959 1960 1961

Date

50

100

150

200

250

Maximum Penetrating Dose (mrem)

Maximum Doses Associated 

with SL-1 Incident

Total Badges

Total Construction Badges

Maximum Penetrating Dose (mrem)
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Figure A.7-6. Total Number of Construction Badges Issued and Maximum Observed 

Penetrating Dose (mrem) with Suspected SL-1 Exposures Removed

0

100

200

300

1964

A.8. SITE RESEARCH DATABASE REPORTS USED IN DATA COMPILATION

SRDB 

# Document Title

125789 Health and Safety Records Section Monthly Reports 1962 

125790 Health and Safety Records Section Monthly Reports 1963 

125920 Dosimetry Branch Reports 1972–1975 

125924 Dosimetry Branch Activity Monthly Reports 1963 

125926 Dosimetry Branch Activity Monthly Reports 1964 

125927 Dosimetry Branch Activity Monthly Reports 1965 

125928 Dosimetry Branch Activity Monthly Reports 1966 

125930 Dosimetry Branch Activity Monthly Reports 1967–1970 

125934 Dosimetry Branch Activity Monthly Reports 1970–1972 

125935 Health and Safety Records Section Monthly Reports 1959 

125938 Health and Safety Records Section Monthly Reports 1960 

400

500

600

Maximum Penetrating Dose (mrem)

0

50

100

1959

150

200

250

Total Badges

1960 1961 1962 1963 Date

Total Construction Badges

Maximum Penetrating Dose

(mrem)
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SRDB 

# Document Title

125948 Health and Safety Records Section Monthly Reports 1961 

125950 Health and Safety Records Section Monthly Reports 1962 

125959 Health and Safety Records Section Monthly Reports 1964 

125961 Health and Safety Records Section Monthly Reports 1965 

125967 Health and Safety Records Section Monthly Reports 1966 

125974 Health and Safety Records Section Monthly Reports 1969 

125975 Health and Safety Records Section Monthly Reports 1970 

125976 Health and Safety Records Section Monthly Reports January and February 1971 

125992 Health and Safety Records Section Monthly Reports January and June 1972 

125993 Health and Safety Records Section Monthly Reports March through December 1971 

126015 Health and Safety Records Section Yearly Summary for 1963

126023 Health and Safety Records Section Yearly Summary for 1965

138710 SPERT Health Physics Monthly Report January – November 1963 

139205 SPERT HP Reports 1958 

153691 Monthly Dosimetry Badge Reports 1973 

153692 Monthly Dosimetry Badge Reports 1974 
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APPENDIX B: EVALUATION OF APPLICABILITY OF OTIB-0054 TO 

RECONSTRUCTING INTERNAL DOSES AT TEST AREA NORTH 

Prepared by

Michael W. Mallett, SC&A

Stephen F. Marschke, SC&A

John Mauro, SC&A

In the various reports and discussions leading up to this report, it was SC&A’s impression that 

the methodology in ORAUT-OTIB-0054, Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross Gamma Analyses, Revision 04 (ORAUT 2015; 

hereafter referred to as “OTIB-0054”), is generally claimant favorable as applied to the Test 

Area North (TAN) reactors, especially the Heat Transfer Reactor Experiment (HTRE) reactors. 

This judgment assumed that the burnup times for the HTRE reactors were much shorter than 

those of the reactors that were used to develop the radionuclide mixes in the irradiated fuel used 

in OTIB-0054. Thus, the mix of radionuclides used in OTIB-0054 would be composed of a 

greater fraction of the long-lived radionuclides (i.e., radionuclides that should have higher 

inhalation dose conversion factors [DCFs] simply because of their generally longer residence 

times in the body). Figure B-1 plots the limiting effective e(50)1

1 e(50) refers to the committed effective dose per unit acute intake over 50 years following the exposure. 

doses in Federal Guidance 

Report (FGR) No. 13 (EPA 1999; hereafter “FGR 13”) as a function of the radioactive half-lives 

of the radionuclides and demonstrates the general validity of this premise. 

Figure B-1. FGR 13 Adult Inhalation e(50) Dose Factor versus Half-Life

However, given the complexity of this subject, SC&A took additional steps to evaluate the 

degree to which OTIB-0054 is, in fact, applicable or claimant favorable to workers at TAN 

 
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facilities, keeping in mind that OTIB-0054 is used to reconstruct the intake rates of individual 

radionuclides by TAN workers who were routinely monitored on a gross beta/gamma urine 

bioassay program.

B.1. RADIONUCLIDE INVENTORY IN THE IRRADIATED FUEL

The first step was to use the ORIGEN code to calculate the number of radionuclide atoms in the 

irradiated fuel per metric ton uranium (MTU) at 10 days after shutdown as a function of power 

level and duration of burnup. These were then converted to activities by multiplying the number 

of atoms by the radioactive decay coefficient of each radionuclide (i.e., AA = λλλλ). These 

calculations were performed for burnup times ranging from 20 hours to 200 days for both lowand high-power cases (i.e., 20 and 200 MW, respectively). The results of the low-power 

ORIGEN run are shown in Figure B-2 for several key radionuclides. As shown, the activities for 

many of the radionuclides (e.g., strontium-90 [Sr-90], cesium-137 [Cs-137], and ruthenium-106 

(Ru-106) continue to build up even after 200 days of burnup time, while other radionuclides 

(most notably iodine-131 [I-131]) have reached equilibrium, and their activities begin to level 

off. 

It is noteworthy that at 20 hours burnup at low power levels, the activity of Cs-137 and Sr-90 are 

about the same (i.e., 2×1012 becquerel [Bq]/MTU),2

2 Note: 1 Bq = 1 disintegration per second. 

which is expected, because they both have 

about the same cumulative fission product yield (about 6%),3

3 The term “cumulative fission product yield” is used because many of the radionuclides in the irradiated fuel result 

from the direct fission product yield in combination with the decay of parent fission products, each of which has its 

own fission product yield and decay rate. Of course, sufficient burnup time must have occurred to allow for the 

buildup of the parent fission products; otherwise, use of the “cumulative fission product yield” will overestimate the 

radionuclide’s inventory.

and the rate of increase of the 

activity of these radionuclides is about the same because their half-lives are similar (i.e., about 

30 years). Thus, their absolute activities continue to increase at the same rate over time and do 

not flatten out over the period examined in this ORIGEN run. However, other radionuclides start 

out at a much higher level because they have relatively shorter half-lives than Cs-137 and Sr-90, 

and, therefore, even if a relatively small number of atoms of these radionuclides are produced 

relative to Cs-137 and Sr-90 during the fission process, their activity is often initially high

because of their higher decay rates (shorter half-lives) as compared to Sr-90 and Cs-137. 

Niobium-95 (Nb-95) is interesting because the combination of its small independent fission yield 

and the relatively long half-life of its parent, zirconium-95 (Zr-95) (64 days) results in very little 

Nb-95 being produced at 20 hours of burnup time, yielding approximately the same activity as 

Sr-90 and Cs-137 in the irradiated fuel. However, the Nb-95 activity climbs rapidly once a 

sufficient Zr-95 activity has been established and approaches an equilibrium level at 200 days of 

buildup (as shown in Figure B-2). This occurs because its rate of production due to Zr-95 decay 

is relatively high (atoms produced per unit time). However, its rate of decay is also high, thereby 

resulting in a rapid rise in activity as a function of time relative to the activity of Cs-137 and 

Sr-90 in the irradiated fuel. Figure B-2 is the foundation upon which subsequent calculations 

related to the low-power case index of harm are based. A similar figure could be made for the 

high-power case. The results of the index of harm calculations are not always intuitively obvious 

because the inventory of a given radionuclide associated with a given burnup time is the result of 

 
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a combination of factors, including the fission yield (atoms produced per fission), the decay rate 

of the radionuclide and its parents, and the assumed cooldown time. 

Figure B-2. ORIGEN-Calculated Low-Power Radionuclide Activity

B.2. NORMALIZATION OF RADIONUCLIDE INVENTORY TO THE INVENTORY 

OF SR-90 

Because the bioassay monitoring results are often provided as gross beta, SC&A normalized the 

ORIGEN results to Sr-90, a beta emitter and a prominent fission product. Figure B-3 shows the 

results of the Figure B-2 low-power activities normalized to the Sr-90 activity. The I-131 curve 

bends downward, not because the I-131 activity is decreasing, but because the Sr-90 activity is 

increasing at a faster rate, making for a decreasing ratio between the two radionuclides. Also, for 

praseodymium-143 (Pr-143), and to a lesser extent yttrium-91 (Y-91), the curves in Figure B-3 

initially increase and then decrease, forming a kind of arch. This behavior indicates that the 

activities for these radionuclides are initially increasing more rapidly than for Sr-90, but then the 

Pr-143 and Y-91 activities reach equilibrium, while the Sr-90 activity continues to increase. 
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Figure B-3. Low-Power Radionuclide Activity Normalized to Sr-90 

B.3. DOSE CONVERSION FACTORS NORMALIZED TO SR-90 DOSE 

CONVERSION FACTOR 

The next step in the calculation was to obtain dose conversion factors for each of the organs of 

interest from FGR 13. For this exercise, adult inhalation DCFs were used. For each radionuclide 

and age group, FGR 13 provides multiple inhalation DCFs, depending on the chemical nature of 

the material inhaled. For most radionuclides, FGR 13 provides three inhalation DCFs to account 

for slow, medium, and fast clearance of the material from the lungs. For a few radionuclides 

(e.g., I-131), FGR 13 provides additional inhalation DCFs. For this study, all the FGR 13 adult 

inhalation DCFs for each radionuclide were examined, and the largest factors were used in the 

calculation.

Figure B-4 shows the FGR 13 e(50) DCFs that were selected for use in this study. Because each 

radionuclide’s activity has been normalized to the Sr-90 activity, it is also necessary to normalize 

the individual radionuclide DCFs to the Sr-90 DCF. These normalized e(50) DCFs are also 

shown in Figure B-4. This procedure was employed to obtain Sr-90 normalized DCFs for each 

organ of interest for this study (i.e., bone surface, breast, lower large intestine (LLI) wall, red 

bone marrow, skin, lung, liver, kidneys, and thyroid); figures like Figure B-4 could be 

developed, if desired, for each organ. 
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Figure B-4. FGR 13 Inhalation e(50) Dose Conversion Factors and Sr-90-Normalized Dose 

Conversion Factors 

B.4. RELATIVE INDEX OF HARM

The Sr-90 normalized activities (e.g., Figure B-3 for the low-power case) were multiplied by the 

Sr-90 normalized inhalation DCFs (e.g., Figure B-4 for e(50) dose) to obtain an Sr-90 relative 

hazard measure for each radionuclide. For the low-power case, Figure B-5 presents e(50) hazard 

for individual radionuclides relative to that of Sr-90. Notice that the e(50) hazard index for I-131 

relative to that for Sr-90 decreases over time. This occurs because the amount of I-131 remains 

constant in the irradiated fuel over time, while the amount of Sr-90 continues to increase. The 

e(50) hazard index for other radionuclides, such as Nb-95, starts low and rises as a function of 

time relative to that of Sr-90. 
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Figure B-5. Low-Power e(50) Dose Relative Hazard Measures

The final step in the SC&A analysis was to compare the hazard measures at different burnup 

times (e.g., 20 hours versus 200 days). Table B-1 presents the calculated relative hazard 

measures for the low-power case at burnup times of 20 hours and 200 days, as well as the ratio of 

the two. Notice in Table B-1 that the e(50) organ 20-hour and 200-day relative hazard measures 

are 70.6 and 27.6, respectively, and that these are the same values shown by the Total curve in 

Figure B-5. 

Table B-1. Relative Hazard Measures for Low Power Level (20 MW)

Organ

Relative Hazard Ratio 

200 d 20 hr 20hr/200d

e(50) 2.76E+01 7.06E+01 2.55

Bone surface 3.25E+01 5.50E+01 1.69

Breast 7.57E+02 1.19E+03 1.57

LLI wall 1.94E+02 3.78E+02 1.95

Red bone marrow 2.94E+01 4.29E+01 1.46

Skin 5.79E+02 8.66E+02 1.50

Lung 2.41E+01 4.05E+01 1.68

Liver 2.62E+04 3.36E+04 1.28

Kidneys 9.22E+02 1.37E+03 1.49

Thyroid 1.27E+04 1.56E+05 12.29
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Likewise, Table B-2 presents the calculated relative hazard measures for the high-power case at 

burnup times of 20 hours, 20 days, and 200 days, as well as the ratio of the two shorter times to 

200 days.

Table B-2. Relative Hazard Measures for High Power Level (200 MW)

Organ

Relative Hazard Ratios

200 d 20 d 20 hr 20d/200d 20hr/200d

e(50) 3.65E+01 6.63E+01 7.07E+01 1.82 1.93

Bone surface 3.66E+01 5.88E+01 5.51E+01 1.61 1.50

Breast 1.21E+03 1.23E+03 1.19E+03 1.02 0.98

LLI wall 2.71E+02 4.58E+02 3.79E+02 1.69 1.40

Red bone marrow 3.38E+01 4.94E+01 4.29E+01 1.46 1.27

Skin 9.70E+02 9.16E+02 8.68E+02 0.94 0.89

Lung 3.17E+01 4.64E+01 4.05E+01 1.47 1.28

Liver 3.07E+04 3.80E+04 3.37E+04 1.24 1.10

Kidneys 1.40E+03 1.91E+03 1.38E+03 1.37 0.98

Thyroid 1.89E+04 9.68E+04 1.57E+05 5.12 8.29

Those relative indices of harm (shown in the Ratios columns) that are substantially greater than 

1.0 indicate those organs for which OTIB-0054, with the assumption of longer burnup times than 

the TAN reactors, might underestimate internal doses. However, for all organs other than the 

thyroid, the ratios that exceed 1.0 do so only modestly. These or similar tables could be used as 

adjustment factors to the doses derived using OTIB-0054 to apply to those workers for whom it 

might be important to increase the calculated assigned dose.