Draft Report 

NATIONAL INSTITUTE FOR 

OCCUPATIONAL SAFETY AND HEALTH 

ADVISORY BOARD ON RADIATION AND WORKER HEALTH 

SC&A REVIEW OF NIOSH RESPONSES TO SC&A REVIEW 

OF ORAUT-OTIB-0054, FINDINGS 1–4: 

FISSION AND ACTIVATION PRODUCT ASSIGNMENT FOR 

INTERNAL DOSE-RELATED GROSS BETA 

AND GROSS GAMMA ANALYSES

Contract No. 211-2014-58081 

Prepared by 

Stephen Ostrow 

John Mauro 

SC&A, Inc. 

1608 Spring Hill Road, Suite 400 

Vienna, VA 22182 

February 2015 

Disclaimer

This document is made available in accordance with the unanimous desire of the Advisory Board on 

Radiation and Worker Health (ABRWH) to maintain all possible openness in its deliberations. However, 

the ABRWH and its contractor, SC&A, caution the reader that at the time of its release, this report is predecisional and has not been reviewed by the Board for factual accuracy or applicability within the 

requirements of 42 CFR 82. This implies that once reviewed by the ABRWH, the Board’s position may 

differ from the report’s conclusions. Thus, the reader should be cautioned that this report is for 

information only and that premature interpretations regarding its conclusions are unwarranted. 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

2 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

SC&A, Inc.: 

Technical Support for the Advisory Board on 

Radiation and Worker Health Review of NIOSH 

Dose Reconstruction Program

Document No.

Response to Comments on Findings 1–4 

of SCA-TR-PR2014-0084 Review

Effective Date:

Draft – February 6, 2015

Revision No:

0 – Draft

SC&A Review of NIOSH Responses to SC&A 

Review of ORAUT-OTIB-0054, Findings 1–4:

Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross 

Gamma Analyses

Supersedes:

N/A

Task Manager:

_____________________ Date: ____________

Stephen L. Ostrow, PhD

Reviewers:

 John Stiver

 John Mauro

 Stephen Ostrow

Project Manager:

_____________________ Date: ____________

John Stiver, MS, CHP

REVISION LOG 

Rev. No. Date Revision

0 2/06/2015 Original
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

3 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

TABLE OF CONTENTS

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

1.0 Introduction..........................................................................................................................5

1.1 Background..............................................................................................................5

1.2 Summary of Status of Findings................................................................................6

1.3 Summary of SC&A’s Recommendations................................................................8

2.0 Evaluation of NIOSH Responses to Findings 1–4...............................................................9

2.1 Reactor Modeling report (ORAUT 2014b)..............................................................9

2.2 Finding 1 ................................................................................................................11

2.3 Finding 2 ................................................................................................................16

2.4 Finding 3 ................................................................................................................17

2.5 Finding 4 ................................................................................................................19

3.0 References..........................................................................................................................20

LIST OF TABLES

Table 1. Status of Findings 1–4 on ORAUT-OTIB-0054 (a),(b).......................................................6

Table 2. Summary of Recommendations on the Status of Findings 1–4........................................8

Table 3. Reactor Categories and Representative Reactors ...........................................................10

Table 4. ATR Fuel Loading Comparison (93.2 wt% enrichment) ...............................................12

Table 5. ORIGEN-S Irradiation Parameters for the Representative Reactor Cases.....................18
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

4 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

ABBREVIATIONS AND ACRONYMS

ABRWH Advisory Board on Radiation and Worker Health

Al Aluminum

ATR Advanced Test Reactor

BRS Board Review System

CANDU CANada Deuterium Uranium (reactor)

DOE U.S. Department of Energy

FFTF Fast Flux Test Facility

g gram

GA General Atomics

INL Idaho National Laboratory

kg kilograms

kW kilowatt

lb pound

MW Megawatt

MWth Megawatt thermal

MWd Megawatt Day

MTHM Metric Tons of Heavy Metal

MTU Metric Tons of Uranium

n/(cm

2

-sec) neutrons per centimeter squared per second

NIOSH National Institute for Occupational Safety and Health

ORAUT Oak Ridge Associated Universities Team

ORIGEN Oak Ridge Isotope Generator

ORNL Oak Ridge National Laboratory

OTIB ORAUT Technical Information Bulletin

PWR Pressurized Water Reactor

SC&A S. Cohen and Associates (SC&A, Inc.)

sec second

SS Stainless Steel

TRIGA Training, Research, Isotopes, General Atomic

U uranium

U.S. United States

UZrH uranium zirconium hydride

wt% Weight Percent
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

5 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

1.0 INTRODUCTION

1.1 BACKGROUND

Frequently, air-sampling or urinalysis data on worker exposure to mixed fission and activation 

products associated with nuclear reactors or nuclear fuel are available only in the form of gross 

beta or gross gamma activity unattributed to specific radionuclides. This is particularly true for 

exposures during the early decades of the U.S. nuclear program. For those cases, ORAUTOTIB-0054, Fission and Activation Product Assignment for Internal Dose-Related Gross Beta 

and Gross Gamma Analyses (hereafter referred to as “the OTIB” or “OTIB-0054”) provides 

guidance and a standard approach to the dose reconstructor on how to assign radionuclidespecific intakes to exposed workers. 

SC&A reviewed Rev. 0 of the Oak Ridge Associated Universities technical information bulletin 

(OTIB) (ORAUT 2007) in 2008 (SC&A 2008) and identified 26 issues, some of which were 

subsequently resolved through work of the Advisory Board on Radiation and Worker Health 

(ABRWH) Subcommittee on Procedures Review, the National Institute for Occupational Safety 

and Health (NIOSH), and SC&A. The technical basis of the OTIB was substantially revised 

from Rev. 0 to Rev. 1 (ORAUT 2013), leading the Subcommittee at its July 18, 2013, meeting to 

authorize SC&A to perform a full de novo review of Rev.1. SC&A presented its comprehensive 

technical review of Rev. 1 of the OTIB in its November 2013 report (SC&A 2013). That review 

produced 10 findings, which NIOSH responded to on February 4, 2014, via entries to the online 

Board Review System (BRS). Those 10 findings were labelled 27–36, but were also numbered 

1–10 for convenience, as they replaced the original group of 26 following SC&A’s de novo

review; the latter numbering will be used throughout this report. 

While SC&A comments were made on Rev. 1 of the OTIB (ORAUT 2013), they apply equally 

to Rev. 2, issued on March 6, 2014 (ORAUT 2014a). The Publication Record for Rev. 2 states: 

Revision initiated to correct an error with the Pm-147 intake fractions in Tables 

7-3b and 7-3c. The values had mistakenly been entered as zeros. No changes 

occurred as a result of formal internal review.

SC&A checked that NIOSH did indeed make that correction in Rev. 2 of those two tables, for 

the ATR (Advanced Test Reactor) 2 and ATR 3, respectively. Based on a cursory comparison of 

Rev. 1 and Rev. 2, SC&A did not notice any other changes to the documents. In addition, the 

changes from Rev. 1 to Rev. 2 do not affect Findings 1–4.

Subsequent to SC&A’s review of the OTIB in SC&A 2013, the Subcommittee on Procedures 

Review teleconferences that discussed the OTIB were held on November 7, 2013; February 13, 

2014; April 16, 2014; August 28, 2014; and November 25, 2014. In addition, SC&A, NIOSH, 

and ORAUT (NIOSH’s subcontractor) participated in technical calls to obtain clarification on 

specific items on May 13, 2014, and October 2, 2014. SC&A issued Rev. 0 of this report on 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

6 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

April 10, 2014 (SC&A 2014), and has now revised it following receipt of ORAUT 2014b, the 

so-called Reactor Modeling Report,1 which responds to Findings 1–4. 

1.2 SUMMARY OF STATUS OF FINDINGS 

The status of, and information about, the 10 findings can be found in the BRS system. The 

current entries for Findings 1–4, the subject of this evaluation, are shown in Table 1. 

(a),(b) Table 1. Status of Findings 1–4 on ORAUT-OTIB-0054 

Finding Board Review System (BRS) Entries Status

1 Ostrow, Stephen (SC&A), 11/5/2013: SC&A is not able to evaluate the 

appropriateness of the input parameters used for the ORIGEN2 runs 

since they are not specified, or references cited in the OTIB.

For more discussion on this finding see the attached SC&A report SCATR-PR-2013-0084 [SC&A 2013]. 

Burns, Bob (ORAUT), 2/4/2014: The ORIGEN2 runs and the process 

of down-selecting to four representative reactors were not affected by 

the revision. A separate report is planned that will document the reactor 

modeling process in detail. 

Munn, Wanda (Chair, Procedures Subcom.), 2/13/2014: Since NIOSH 

has provided a response to the finding, the status has been changed to In 

Progress and SC&A has been asked to review the NIOSH response. 

Ostrow, Stephen (SC&A), 4/10/2014: SC&A will review the NIOSH 

reactor modeling report when it is available and recommends that the 

status of this finding remain In Progress pending that review.

Ostrow, Stephen (SC&A), 4/16/2014: This finding on OTIB-0054, 

Revision 01 also applies to Revision 02 (03/06/2014). 

In Progress

 

1

 This report will usually refer to ORAUT 2014b by its unofficial name, the Reactor Modeling Report. 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

7 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Table 1. Status of Findings 1–4 on ORAUT-OTIB-0054 (a),(b)

Finding Board Review System (BRS) Entries Status

2 Ostrow, Stephen (SC&A), 11/5/2013: The OTIB does not provide 

sufficient information to allow evaluation of its down-select from the 

initial seven to the final four representative reactors chosen. 

For more discussion on this finding, see SC&A report SCA-TR-PR2013-0084, attached to Finding OTIB-0054-027 [i.e., Finding 1]. 

Burns, Bob (ORAUT), 2/4/2014: See response to Rev. 1, Finding 1. 

Munn, Wanda (Chair, Procedures Subcom.), 2/13/2014: Since NIOSH 

has provided a response to the finding, the status has been changed to In 

Progress and SC&A has been asked to review the NIOSH response. 

Ostrow, Stephen (SC&A), 4/10/2014: SC&A will review the NIOSH 

reactor modeling report when it is available and recommends that the 

status of this finding remain In Progress pending that review.

Ostrow, Stephen (SC&A), 4/16/2014: This finding on OTIB-0054, 

Revision 01, also applies to Revision 02 (03/06/2014). 

In Progress

3 Ostrow, Stephen (SC&A), 11/5/2013: While Rev. 0 of the OTIB 

(Section 5.2) provides extensive discussions of the ORIGEN2 runs for 

each reactor, Rev. 1 does not for the ORIGEN-S runs. For each of the 

nine representative reactor cases, the OTIB (Table 5-2) specifies the 

specific power, irradiation time, and burnup, and includes a basis (e.g., 

“maximum burnup at nominal power” for ATR 1), but does not say how 

the values were selected or cite any reference; Rev. 0 made extensive 

use, for example, of the DOE report, Source Term Estimates for DOE 

Spent Nuclear Fuels, DOE/SNF/REP-078, Rev. 0, March 2003 [DOE 

2003]. SC&A cannot fully evaluate the appropriateness of the values 

chosen for each case without such information.

For more discussion on this finding, see SC&A report SCA-TR-PR2013-0084, attached to Finding OTIB-0054-027 [i.e., Finding 1]. 

Burns, Bob (ORAUT), 2/4/2014: See response to Rev. 1, Finding 1. 

Munn, Wanda (Chair, Procedures Subcom.), 2/13/2014: Since NIOSH 

has provided a response to the finding, the status has been changed to In 

Progress and SC&A has been asked to review the NIOSH response. 

Ostrow, Stephen (SC&A), 4/10/2014: SC&A will review the NIOSH 

reactor modeling report when it is available and recommends that the 

status of this finding remain In Progress pending that review.

Ostrow, Stephen (SC&A), 4/16/2014: This finding on OTIB-0054, 

Revision 01, also applies to Revision 02 (03/06/2014). 

In Progress
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

8 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Table 1. Status of Findings 1–4 on ORAUT-OTIB-0054 (a),(b)

Finding Board Review System (BRS) Entries Status

4 Ostrow, Stephen (SC&A), 11/5/2013: SC&A notes that Table 5-1 of 

the OTIB lists both aluminum and stainless steel-clad TRIGA reactors 

as belonging to the initial set of seven reactors. However, Table 5-2, 

which lists the four reactors chosen as references, as well as the 

accompanying text, do not indicate which cladding was assumed for the 

TRIGA reactor. The OTIB also does not indicate what fuel enrichment 

was chosen, give a source for the specific power or the chosen burnups, 

or provide justification for its assumptions.

For more discussion on this finding, see SC&A report SCA-TR-PR2013-0084, attached to Finding OTIB-0054-027 [i.e., Finding 1]. 

Burns, Bob (ORAUT), 2/4/2014: The cladding type (stainless steel) can 

be specified in Table 5-2.

See response to Rev. 1, Finding 1.

Munn, Wanda (Chair, Procedures Subcom.), 2/13/2014: Since NIOSH 

has provided a response to the finding, the status has been changed to In 

Progress and SC&A has been asked to review the NIOSH response. 

Ostrow, Stephen (SC&A), 4/10/2014: SC&A will review the NIOSH 

reactor modeling report when it is available and recommends that the 

status of this finding remain In Progress pending that review.

Ostrow, Stephen (SC&A), 4/16/2014: This finding on OTIB-0054, 

Revision 01, also applies to Revision 02 (03/06/2014).

In Progress

Notes:

(a) BRS entries are current as of January 7, 2015.

(b) The findings apply equally to Rev. 1 and Rev. 2 of the OTIB

1.3 SUMMARY OF SC&A’S RECOMMENDATIONS

Sections 2.2 through 2.5 present SC&A’s review of NIOSH’s response in the Reactor Modeling 

Report to Findings 1–4, respectively. SC&A’s recommendations for the disposition of those 

findings are shown in Table 2. 

Table 2. Summary of Recommendations on the Status of Findings 1–4

Finding Current Status Recommended Status

1 In Progress Closed

2 In Progress In Progress

3 In Progress Closed

4 In Progress Closed
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

9 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

2.0 EVALUATION OF NIOSH RESPONSES TO FINDINGS 1–4 

As seen in Table 1, Findings 1–4 were designed in the BRS as In Progress pending receipt and 

evaluation of the so-called NIOSH Reactor Modeling Report (ORAUT 2014b). This report, 

entitled, Supporting Calculations for ORAUT-OTIB-0054 and ORAUT-RPRT-0047:

...summarizes the modeling of the reactor fuel and core configurations for the 

development of the following Oak Ridge Associated Universities (ORAU) Team 

work products:

 ORAUT-OTIB-0054, “Fission and Activation Product Assignments for 

Internal Dose-Related Gross Beta and Gross Gamma Analyses,” Revision 2...

 ORAUT-RPRT-0047, Assignment of Fission and Activation Product 

Radionuclides for Non-Specific Bioassays at Savannah River Site –

Comparison of Methods, Revision 00... 

The current SC&A evaluation reviews only those portions of the Reactor Modeling Report 

pertinent to OTIB-0054 (i.e., it doesn’t examine the Savannah River portions). 

2.1 REACTOR MODELING REPORT (ORAUT 2014b)

The Reactor Modeling Report (ORAUT 2014b) is a very detailed and technical description of 

how ORAUT modeled fuel in a number of representative nuclear reactors, and developed 

accompanying cross section and other datasets in order to determine isotopic compositions under 

a variety of conditions and assumptions. It is not the intention of this report to repeat what is 

presented in the very detailed Reactor Modeling Report or elsewhere, but a brief summary would 

be instructive.2

Although modern nuclear power reactor and fuel designs in the United States have been reduced 

to primarily a few different types since their first appearance in the late 1940s and early 1950s, 

many different experimental, training, and special purpose reactors (e.g., plutonium production, 

tritium production, or materials testing reactors) were developed and operated, especially at the 

weapons complex laboratories, such as Idaho National Laboratory, Savannah River Site, and 

Hanford. Recognizing that, out of practical considerations, a dose reconstructor would not be 

able to model every reactor and fuel type and operating history that a claimant might have

encountered, it would be highly desirable to reduce the large number of possible combinations to 

a more manageable, representative few that would encompass most cases. 

The primary goal of the modeling in the Reactor Modeling Report, which is an essential part of 

OTIB-0054, is to reduce a large amount of representative reactor fuel isotopic data into a form 

easily usable by a dose reconstructor looking at actual claimant cases. The starting point is the 

radionuclide mix in spent fuel for a number of different reactor types and fuel designs operated 

under a variety of different conditions (e.g., specific power, irradiation time, and burnup) 

calculated at several different decay times following removal from the reactor. It is assumed that 

an actual claimant case condition will fit somewhere within the parameters defined by the 

 

2

Parts of the following description are adapted from SC&A 2013 and ORAUT 2014b. 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

10 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

selected representative reactor types. As stated in the introduction of the Reactor Modeling 

Report:

The first step served to narrow the reactor and fuel designs under consideration 

to a manageable number. Once the representative reactor systems were selected, 

depletion modeling was performed to develop case-specific cross-section sets for 

each system over a range of operating conditions. Last, the case-specific crosssection sets were used in radioisotope generation and depletion calculations to 

obtain the fission and activation product inventories (by activity) for each reactor 

system as a function of time after reactor shutdown. Multiple inventory 

calculations were performed for the representative reactor systems to encompass 

a range of operating conditions. 

Section 5 of the OTIB and Section 3 of the Reactor Modeling Report begin by choosing seven 

actual reactors to represent five different general categories. Table 3 summarizes the reactors 

and cases:

Table 3. Reactor Categories and Representative Reactors

Reactor Category

Plutonium Production Reactors

Hanford N-Reactor

Hanford Single-Pass Reactors

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

High-flux Reactors Advanced Test Reactor (ATR)

Research Reactors

TRIGA® Reactor (Al-clad fuel)

TRIGA® Reactor (SS-clad fuel)

Generic Reactor Pressurized Water Reactor (PWR)

Source: ORAUT 2014a, Table 5-1 and ORAUT 2014b Table 3-1. 

The ORIGEN2 Version 2.1 isotope generation and depletion code (Croff 1980) was then used to 

calculate isotopic inventories for the seven representative reactors in Table 3 in 11 runs, decayed 

for 7 decay times after irradiation had ceased: 10, 40, 60, 90, and 180 days, and 1 and 3 years. 

The code contains nuclear data libraries for the FFTF, ATR, and PWR, but not for the two 

Hanford production reactors and the two TRIGA reactors. For each of those cases lacking an 

ORIGEN2 library, ORAUT ran the code twice with different available cross-section and fission 

product yield libraries chosen to attempt to bound the irradiation results.

ORIGEN2 is a widely-used, one energy group, industry-standard code, developed by Oak Ridge 

National Laboratory (ORNL), to calculate the isotopic inventory after buildup and decay of 

isotopes in a nuclear reactor under conditions specified by the user. The code uses a matrix 

exponential method to solve coupled linear, first-order differential equations with constant 

coefficients. The code contains data libraries, including cross sections, fission product yields, 

decay data, and decay photon data. SC&A believes that the use of ORIGEN2 is appropriate and

consistent with common practice in the nuclear community.
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

11 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

The 11 ORIGEN2 runs produced activity data for 879 fission product nuclides and 688 

activation product nuclides, with a large overlap of nuclides between the two categories, and 

including some stable nuclides in addition to the radioactive ones. The activities in each dataset 

were normalized to that of Cs-137 at 10 days of decay. 

The Reactor Modeling Report states:

Based on comparison of the fission and activation product relative activity data 

for the 11 ORIGEN2 runs, four cases were selected as having isotopic inventories 

that would be representative for the variety of reactors and fuel combinations: 

the ATR, the FFTF, the N-Reactor (run with a PWR cross-section library), and 

the TRIGA reactor (run with stainless-steel (SS) cladding and a PWR crosssection library).

The fission and activation product relative inventory data were summed to produce tables of 

relative activity (normalized to Cs-137 at 10 days of decay) for 1,192 nuclides for the 7 decay 

times. 

ORAUT then switched from the ORIGEN2 to the ORIGEN-S code for further fission and 

activation product inventory calculations for the four representative reactors remaining. 

ORIGEN-S (which uses three neutron energy groups: thermal, resonance, and fast) is a more 

modern and capable version of ORIGEN2 and is part of the SCALE code system (ORNL 2015)

for nuclear safety analysis and design, developed and maintained by ORNL. The ORIGEN-ARP 

preprocessor prepared data and the TRITON code created case-specific cross-section libraries for 

ORIGEN-S; both codes are modules in the SCALE system. Multiple sets of irradiation 

parameters were defined for each of the four representative reactors. SC&A believes that use of 

these SCALE system modules is appropriate.

2.2 FINDING 1

The BRS entry associated with this finding is shown in Table 1, and the background for that

finding is found in Section 1.1 of SC&A 2013. To repeat the finding here for convenience: 

“SC&A is not able to evaluate the appropriateness of the input parameters used for the 

ORIGEN2 runs, since they are not specified, or references cited in the OTIB.”

Section 2.1 of the Reactor Modeling Report discusses the use of ORIGEN2:

...used to perform the fission and activation product inventory calculations used 

to narrow down the broad set of reactor systems initially selected for OTIB-0054 

to the four ultimately considered (ORAUT 2014b).

Subsections of Section 3.2 of the Reactor Modeling Report presents, in turn, the parameters used 

for each of the 7 reactors considered in the 11 ORIGEN2 runs. SC&A reviewed the chosen 

parameters and examined some of the copious technical literature available on each of the 

reactors to inform an opinion on the appropriateness of ORAUT’s parameter choices. The 

different reactors will be examined below.
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

12 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

Advanced Test Reactor (ATR)

The ATR is a materials testing reactor that has operated at Idaho National Laboratory (INL) for 

over 40 years (it is still in operation). Its high-flux capability allows simulation of long-term 

irradiation in less than real time. An INL website states the following:

The ATR is a water-cooled, high-flux test reactor, with a unique serpentine design 

that allows large power variations among its flux traps. The reactor’s curved fuel 

arrangement places fuel closer on all sides of the flux trap positions than is 

possible in a rectangular grid. The reactor has nine of these high-intensity 

neutron flux traps and 68 additional irradiation positions inside the reactor core 

reflector tank, each of which can contain multiple experiments. Experiment 

positions vary in size from 0.5" to 5.0" in diameter and all are 48" long. The peak 

thermal flux is 1×1015 n/cm2-sec and fast flux is 5×1014 n/cm2-sec when operating 

at full power of 250 MW. There is a hydraulic shuttle irradiation system, which 

allows experiments to be inserted and removed during reactor operation, and 

pressurized water reactor (PWR) loops, which enable tests to be performed at 

prototypical PWR operating conditions. Learn more by viewing the ATR User’s 

Guide (INL 2015).

The ATR User’s Guide (INL 2008), referenced above, contains much detailed information about 

the ATR; considerably more than is required for the ORIGEN2 modeling.

Section 3.2.1 of the Reactor Modeling Report summarizes the important features of the ATR 

model developed by ORAUT. The section contains data, including the uranium isotopic masses 

and aluminum and impurity masses of a single fuel element, and the assumed fuel burnup. Most 

of the data is taken from Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2003). 

This data source was prepared by the Department of Energy (DOE) to support Yucca Mountain 

safety analysis studies by characterizing source terms for a wide range of spent nuclear fuel held 

by DOE at its various facilities. The DOE report was examined by SC&A, which found that the 

assumptions made and the methodologies used appear reasonable and should produce 

conservative (i.e., high) radioactivity estimates. DOE 2003 explicitly estimates the ATR spent 

fuel source terms as Template 12. 

SC&A examined whether the DOE report data were correctly transferred to the Reactor 

Modeling Report. 

Table 4. ATR Fuel Loading Comparison (93.2 wt% enrichment)

Isotope

Fuel Loading, g/fuel element

DOE 2003 (p. A-103) ORAUT 2014b (Section 3.2.1)

U-235 1,075.0 1,073.0

U-238 69.93 63.64

U-234 13.87 12.29

U-236 8.09 2.40

Total 1,166.89 1,151.33

While not an exact match, the differences are not significant for nuclear fuel modeling. 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

13 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

The Reactor Modeling Report uses the DOE report’s upper-end burnup of 314,683 MWd/

MTHM, which corresponds to 362.3 MWd for the assumed fuel element, and states, “The 

composition was therefore assumed to be irradiated for 36.23 days at a constant power of 

10 MW in increments of 10 days” (ORAUT 2014b). DOE 2003 (p. A-104) gives 10 MWth as 

the maximum power level for a fuel element. Finally, the Reactor Modeling Report notes that,

“the cross-section and fission product yield library was developed specifically for the ATR by

the Idaho National Laboratory,” but since the INL data do not include ATR-specific activation 

product cross sections, the Reactor Modeling Report took them from the standard-burnup PWR 

library accompanying ORIGEN2. Actinide and fission product cross sections are ATR-specific, 

however. SC&A believes that the process employed in the Reactor Modeling Report is valid. 

Fast Flux Test Facility (FFTF)

The FFTF was a 400 MWth, liquid sodium-cooled, fast-flux test reactor located at the DOE 

Hanford site that operated from April 1982 to April 1992 (DOE 2014). The Reactor Modeling 

Report draws most of its information on this reactor from DOE 2003, where it is modeled 

explicitly as Template 3. DOE 2003 states that the fuel was mixed oxide with plutonium 

(86 wt% Pu-239) amounting to 29 wt% of the heavy metal (uranium plus plutonium); the driver 

assemblies were clad with 316 SS; and the assemblies were subjected to high burnups. The

Reactor Modeling Report lists the heavy metal (uranium, plutonium, and americium) contents of 

a driver assembly in Table 3-3 and the elemental content of the 316 SS of a driver assembly in 

Table 3-4. The ORAUT reactor model specifies that:

...a representative burnup of 80,000 MWd/ MTHM was selected, which equates to 

2,633.4 MWd for the fuel assembly defined above. The composition was therefore 

assumed to be irradiated for 487.7 days at a constant power of 5.4 MW in 

increments of 50 days (ORAUT 2014b). 

SC&A verified that the material contents listed in Tables 3-3 and 3-4 correspond to the values in 

Tables 1 and 2, respectively, of DOE 2003. In addition, DOE 2003 (Table 3 and p. A-6) gives 

the average power for a standard fuel assembly as 5.4 MW, which corresponds to the assumption 

in the Reactor Modeling Report.

DOE 2003 (p. A-6) notes that the highest burnup was 152,230 MWd/MTHM, with most in the 

range of 70,000–90,000. Therefore, SC&A believes that the Reactor Modeling Report’s choice 

of 80,000 MWd/MTHM appears representative. In addition, ORIGEN2 already contains crosssection and fission product yield libraries created specifically for the FFTF.

Hanford N Reactor

The features of the Hanford N Reactor, which operated from 1963 to 1987, are summarized in 

DOE 2003 (Template 7, p. A-45), which, as noted in that document, was largely taken from 

Bergsman (1994); the former characterizes and inventories the spent fuel stored at the Hanford 

site. 

The ... N-Reactor is a graphite-moderated, [slightly enriched] pressurized watercooled reactor ... initially designed for plutonium production for national defense 

[and, later] modified to produce steam to be used by the Washington Public 

Power Supply System to generate electricity. N-Reactor was the only dual-
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

14 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

purpose reactor in the United States. The core of the N-Reactor was a 1800-ton 

graphite block, 33 feet (10 meters) high by 33 feet (10 meters) wide by 39 feet 

(12 meters) long (DOE 2003, p. A-46). 

Bergsman notes that the N Reactor primarily used two different types of fuel elements, the Mark 

IV and the Mark IA, both clad with Zircaloy-2. The Mark IV elements had a pre-operations 

enrichment of 0.947% U-235 in both of its concentric tubes, while the Mark IA elements had a 

pre-operations enrichment of 1.25% U-235 in the outer tube and 0.947% in the inner tube. The 

Mark IV elements, though, contained a greater uranium weight than the Mark IA elements—

50.0 lb (22.7 kg) vs. 35.9 lb (16.3 kg) (Bergsman 1994, p. 3). DOE 2003 based its radioactive 

inventory determination on the Mark IV elements, as does the Reactor Modeling Report. The 

Reactor Modeling Report lists the uranium composition per metric ton of fuel in Section 3.2.3. 

The values can be obtained from DOE 2003 by dividing by 11.6 MTU (DOE 2003, p. A-53). 

Table 3-5 lists the elemental composition of the Zircaloy 2 cladding and trace elements. 

ORAUT selected a representative burnup of 1,188 MWd/MTHM (Schwarz 1997), which is 

characteristic of the outer fuel elements (and a higher burnup than experienced by the inner 

elements) and, in its reactor model, irradiates the fuel for 100 days at 11.88 MW in increments of 

20 days. The Reactor Modeling Report notes that “a cross section and fission product yield 

library specifically for the N Reactor exists, but it was not readily available” (ORAUT 2014b). 

Instead, ORAUT ran two sets of ORIGEN2 cases, using data from a slightly enriched CANDU 

reactor and a typical, more highly enriched, U.S. PWR. 

SC&A believes that the Reactor Modeling Report appears to have taken all its data for the 

N Reactor from authoritative sources and reasonably simulated the actual isotopic composition 

after irradiation by considering two different data sets.

Hanford Single-Pass Reactors

Nine light water-cooled, graphite-moderated reactors to produce plutonium for nuclear weapons 

were built along the Columbia River on the Hanford reservation over a 20-year period beginning 

in 1943. They were designated 100-B, C, D, DR, F, H, KW, KE, and N. The first eight were 

“single pass,” in that the cooling water for the core was taken from the river and not recirculated, 

while the last, the N Reactor, recirculated the primary coolant water in a closed loop. The 

cylindrical metallic uranium fuel “slugs” were clad with aluminum. The single-pass reactors 

were shut down between 1964 and 1971 (Gerber 1995). 

The Reactor Modeling Report presents data on the single-pass reactors in Section 3.2.4, 

including typical fuel (uranium and aluminum) dimensions and densities for the fuel slugs used 

during the Manhattan Project era. Scaling up the amount of uranium and aluminum in a single 

slug to 1 metric ton of metal results in the aluminum composition in Table 3-6 and the uranium 

loading just above it: U-234 - 55.3 g, U-235 - 7,110 g, and U-238 - 992,834.7 g. 

The Reactor Modeling Report adopts a representative burnup of 220 MWd/MTHM for those 

slugs, and the ORIGEN2 runs irradiated the fuel for 200 days at a constant power of 1.1 MW. 

Since ORIGEN2 cross-section and fission product yield libraries specific to the single-pass 

reactors were not available, ORAUT ran cases using a natural uranium oxide fuel CANDU 

reactor library and a U.S. 3.2 wt% enrichment PWR library. The Reactor Modeling Report 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

15 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

provides several references (Kelly 1945; Gawthrop and Jones 1945; and Heeb 1993) as data 

sources.

SC&A believes that ORAUT has adequately documented its data sources for the Hanford SinglePass Reactor modeling. 

Pressurized Water Reactor (PWR)

Section 3.2.5 of the Reactor Modeling Report describes features of the standard burnup U.S. 

PWR from an ORNL report (Ludwig and Reiner 1989). The Reactor Modeling Report states 

that this ORIGEN2 model “was included to see how the results from the other models compared 

with those for a system with such well-known behavior” (ORAUT 2014b). The UO2 PWR fuel 

is enriched to 3.2% and clad with Zircaloy 4. Table 3-7 and an unlabeled table give the isotopic 

composition of the fuel and the elemental composition of the fuel and cladding normalized to 

1 metric ton of uranium. SC&A checked that the values correspond to those of Ludwig and 

Reiner 1989. The fuel loading is U-234 - 290.0 g, U-235 - 32,000 g, and U-238 - 967,710.0 g. 

A total fuel burnup of 33,000 MWd/MTHM and an irradiation period of 880 days at a constant 

power of 37.5 MW in increments of 60 days are assumed. The ORIGEN2 code package contains 

standard burnup PWR cross-section and fission product datasets. SC&A finds the NIOSH 

procedure reasonable and data sources adequately referenced. 

TRIGA Reactors

Section 3.2.6 of the Reactor Modeling Report considers Al-clad and SS-clad TRIGA reactors. 

The low-power, water-cooled, pool-type, uranium zirconium hydride (UZrH)-fueled reactors are 

manufactured by General Atomics (GA) primarily for research and educational purposes, and are 

considered inherently safe due to their very large prompt negative temperature coefficient of 

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

raises the water coolant temperature). This feature also allows the TRIGA to be pulsed to a high 

power level, followed by a quick automatic self-shutdown. GA has supplied over 65 reactors 

throughout the world over the past 45 years, ranging in (thermal) power from 20 kW to 16 MW 

and containing 60–100 fuel elements (GA 2015). The models progressed from Mark I through 

Mark III, with a number of specially modified units also produced. The early models used 

highly enriched uranium (70%), but the later models greatly reduced the enrichment (to 20%) in 

response to U.S. government proliferation concerns, and the earlier models were converted to the 

lower fuel enrichment as well.

The Reactor Modeling Report relies on information on the TRIGA from the INL report Sterbentz 

(1997) and the Westinghouse Hanford report Schmittroth and Lessor (1996) for constructing its 

ORIGEN2 model for one fuel rod, distinguishing between Al-clad and SS-clad cases. The 

Reactor Modeling Report notes that:

The compositions and burnup for the ORIGEN2 modeling for TRIGA reactors 

were selected to represent TRIGA reactors in the DOE complex. These reactors 

were primarily used for radiography applications. It was not intended to 

represent the entire gamut of TRIGA reactor varieties at universities and other 

research-oriented facilities. (ORAUT 2014b)
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

16 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

The Al-clad case considers a single fuel element containing 180 g of uranium, and assumes 20% 

enriched U-235 at 8 wt% of the UZrH, with a Zr:H ratio of 1.0. Table 3-8 presents the 

composition of the Al cladding, and the information before the table lists the uranium isotopic 

composition: U-234 - 0.299 g, U-235 - 35.508 g, U-236 - 0.177 g, and U-238 - 144.014 g. The 

constituent and impurity compositions for the ZrH matrix are presented in Table 3-9, taken from 

DOE 2003. The Reactor Modeling Report states, “A midrange burnup of 21,116 MWd/MTHM 

was used, which equates to 5.9 MWd for the fuel element” (ORAUT 2014b). The ORIGEN2 

run assumed irradiation for 5.384 years at a constant power of 3.0 kW per fuel element in 

increments of 0.25 years. 

The SS-clad case for a single fuel element specified 20% enrichment of U-235 at 8.5 wt% of the 

UZrH matrix, with a Zr:H atom ratio of 1:7. Each fuel element contained 195 g of uranium and 

819.4 g of 304 SS. The uranium isotopic composition is U-234 - 0.324 g, U-235 - 38.467 g,

U-236 - 0.192 g, and U-238 - 156.015 g. Table 3-10 lists the 304 SS composition, and 

Table 3-11 lists the constituent and impurity composition of the ZrH matrix. A midrange burnup 

of 19,462 MWd/MTHM was assumed, which is equivalent to 3.8 MWd for one fuel element. 

The element was irradiated in ORIGEN2 for 3.469 years at a constant power of 3.0 kW in 

increments of 0.25 years. 

ORAUT could not locate a TRIGA-specific ORIGEN2 model, so for both Al-clad and SS-clad 

cases, ORAUT used available ATR and standard burnup PWR cross-section and fission product 

yield libraries. 

SC&A believes that the Reactor Modeling Report adequately references the data sources and 

assumptions underlying the ORIGEN2 runs. 

SC&A’s Assessment for Finding 1: After review of the data sources and input 

parameters presented in the ORAUT Reactor Modeling Report (ORAUT-RPRT0067, Rev. 0, August 26, 2014) as well as an examination of the literature, SC&A 

is satisfied that the report adequately specifies and references the pertinent input 

parameters and assumptions associated with the ORIGEN2 runs and finds them 

appropriate. SC&A, therefore, recommends that this finding be Closed. 

2.3 FINDING 2

The BRS entry associated with this finding is shown in Table 1 and the background for that 

finding is found in Section 1.1 of SC&A 2013. To repeat the finding here for convenience: 

“The OTIB does not provide sufficient information to allow evaluation of its down-select from 

the initial seven to the final four representative reactors chosen.”

Section 3.3 of the Reactor Modeling Report discusses the process that ORAUT used, subsequent 

to its 11 ORIGEN2 runs on the initial set of reactor cases, to reduce that set to the final 4 chosen 

for further analysis. That section, copied in its entirety, explains:

The results from the 11 ORIGEN2 runs were compared on the basis of relative 

activity (to 137Cs) at 10 days of decay. Fission and activation product inventories 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

17 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

from the 11 runs were compared separately so potential dependence on the choice 

of cross-section libraries could be evaluated independently for these different 

modes of production. Each ORIGEN2 run gave activity (in curies) for 879 fission 

product nuclides and 688 activation product nuclides. These include some stable 

species, and many nuclides appear in both categories (i.e., they are produced by 

both fission and activation).

Based on comparison of the fission and activation product relative activity data 

for the 11 ORIGEN2 runs, four reactors were selected as having isotopic 

inventories and ratios that would be representative of the variety of reactors and 

fuel combinations under consideration:

 The ATR

 The FFTF

 The N Reactor

 The TRIGA reactor (with SS-clad fuel)

Section 3.3 of the Reactor Modeling Report, reproduced above, appears identical to Section 5.2 

of the OTIB (ORAUT 2014a), with only the addition of the parenthetical text in the last bullet of 

the former (which answers the question of which TRIGA reactor was chosen). While the section 

of the Reactor Modeling Report explains the procedure that ORAUT used, which appears 

reasonable, it does not provide sufficient detail, such as comparative data in tables, to allow 

assessment of the down-selection from the initial seven to the final four representative reactors. 

The selected reactors should both capture the full range of possible isotopic mixtures to which 

workers might have been exposed, as well as represent the most commonly encountered types of 

reactors. 

SC&A Assessment for Finding 2: Section 3.3 of the Reactor Modeling Report 

discusses the down-selection process. However, it is virtually identical to 

Section 5.2 of the OTIB (Rev. 1 or 2) on which SC&A originally made the 

finding. While the explanation of the procedure followed to reduce the number of 

reactors considered from the initial seven to the final four representative ones 

appears reasonable, the section does not provide sufficient detail, such as 

comparative data in tables, to allow assessment of the final choices; i.e., whether 

they capture the full range of isotopic mixtures that might have been encountered 

by workers, and whether they represent the most commonly encountered types of 

reactors. SC&A, therefore, recommends that this finding remain In Progress until 

such time that NIOSH provides additional information that can be evaluated. 

2.4 FINDING 3

The BRS entry associated with this finding is shown in Table 1 and the background for that 

finding is found in Section 1.1 of SC&A 2013. To repeat the finding here for convenience: 

While Rev. 0 of the OTIB (Section 5.2) provides extensive discussions of the 

ORIGEN2 runs for each reactor, Rev. 1 does not for the ORIGEN-S runs. For 

each of the nine representative reactor cases, the OTIB (Table 5-2) specifies the 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

18 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

specific power, irradiation time, and burnup, and includes a basis (e.g., 

“maximum burnup at nominal power” for ATR 1), but does not say how the

values were selected or cite any reference; Rev. 0 made extensive use, for 

example, of the DOE report, Source Term Estimates for DOE Spent Nuclear 

Fuels, DOE/SNF/REP-078, Rev. 0, March 2003 [DOE 2003]. SC&A cannot fully 

evaluate the appropriateness of the values chosen for each case without such 

information.

Section 5 of the Reactor Modeling Report provides extensive information on the ORIGEN-S 

cases for each of the representative reactors: the ATR, the FFTF, the Hanford N Reactor, and the 

TRIGA Mark I (SS cladding). In addition, Section 6 provides detailed information on the 

TRITON depletion modeling portion (required to generate case-specific cross-section sets) of the 

overall analysis for each reactor.

The OTIB (ORAUT 2014a) summarizes the important modeling parameters of the Reactor 

Modeling Report in Table 5-2, which is reproduced here in Table 5.

Table 5. ORIGEN-S Irradiation Parameters for the Representative Reactor Cases

Case Parameters Basis

ATR 1 Specific power = 2,379.1 MW/MTU

Irradiation time = 132.27 days

Burnup = 314,684 MWd/MTU

Maximum burnup at nominal power

ATR 2 Specific power = 8,651.2 MW/MTU

Irradiation time = 36.4 days

Burnup = 314,904 MWd/MTU

Maximum burnup at maximum assembly power

ATR 3 Specific power = 2,379.1 MW/MTU

Irradiation time = 56 days

Burnup = 133,230 MWd/MTU

Nominal burnup at nominal power

FFTF 1 Specific power = 163.8 MW/MTHM

Irradiation time = 929.4 days

Burnup = 152,230 MWd/MTHM

Maximum burnup at nominal power

FFTF 2 Specific power = 163.8 MW/MTHM

Irradiation time = 488.3 days

Burnup = 79,984 MWd/MTHM

Nominal burnup at nominal power

N-Reactor 1 Specific power = 10.4 MW/MTU

Irradiation time = 114.2 days

Burnup = 1,188 MWd/MTU

Production of weapons-grade plutonium (nominal 

6% Pu-240 content) at nominal power

N-Reactor 2 Specific power = 10.4 MW/MTU

Irradiation time = 285.6 days

Burnup = 2,970 MWd/MTU

Production of fuel-grade plutonium (nominal 

12% Pu-240 content) at nominal power

TRIGA 1 Specific power = 15.57 MW/MTU (a)

Irradiation time = 730.1 days

Burnup = 11,368 MWd/MTU

Maximum burnup at nominal power

TRIGA 2 Specific power = 15.57 MW/MTU

Irradiation time = 115.2 days

Burnup = 1,994 MWd/MTU

Nominal burnup at nominal power

(a) OTIB Table 5-2 typo misstates “MWd” rather than the correct “MW,” which is used here. 

Source: ORAUT-OTIB-0054, Rev. 2, Table 5-2 (ORAUT 2014a)

SC&A made a quick check by comparing the tabulated values to those in the Reactor Modeling 

Report, and by multiplying the specific power by the irradiation time for each case, which should 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

19 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

then produce the last entry, the burnup. The “arithmetic” is essentially correct, except for the 

TRIGA 2 case, where (15.57 MW/MTU) × (115.2 days) = 1,794 MWd/MTU. The table lists 

1,994 MWd/MTU, however, which might just be a typo, since it is off in only one digit; this is 

not seen as an important difference in any event. 

SC&A Assessment for Finding 3: SC&A believes that the Reactor Modeling 

Report contains sufficient information on the parameters selected for the 

ORIGEN-S runs for each of the representative reactors to inform an assessment 

that the values chosen are appropriate. SC&A, therefore, recommends that this 

finding be Closed. 

2.5 FINDING 4 

The BRS entry associated with this finding is shown in Table 1 and the background for that 

finding is found in Section 1.1 of SC&A 2013. To repeat the finding here for convenience: 

SC&A notes that Table 5-1 of the OTIB lists both aluminum and stainless steelclad TRIGA reactors as belonging to the initial set of seven reactors. However, 

Table 5-2, which lists the four reactors chosen as references, as well as the 

accompanying text, do not indicate which cladding was assumed for the TRIGA 

reactor. The OTIB also does not indicate what fuel enrichment was chosen, give 

a source for the specific power or the chosen burnups, or provide justification for 

its assumptions.

Section 3.2.6 of the Reactor Modeling Report provides information and references pertaining to 

the TRIGA reactors, and, in particular, the SS-clad one, which was chosen as the TRIGA 

reference reactor. The report states, “The SS-clad TRIGA fuel that was modeled was 

20%-enriched 235U at 8.5 wt% of the UZrH matrix.” The uranium and cladding compositions 

per fuel rod are given in the tables in that section. Section 5.4 of the Reactor Modeling Report, 

as discussed above, gives the specific powers and burnups assumed for the TRIGA reactor cases, 

and Section 7.4 contains the composition of the UZrH fuel and the 304 SS cladding for the entire 

core. 

SC&A Assessment for Finding 4: The Reactor Modeling Report contains the 

information on the TRIGA reactor cases found lacking in the OTIB, which is 

consistent with the literature. SC&A, therefore, recommends that this finding be 

Closed. 
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

20 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

3.0 REFERENCES

Bergsman, K.H., 1994. Hanford Spent Fuel Inventory Baseline, WHC-SD-SNF-TI-001, Rev. 0, 

July 15, 1994.

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

Generation and Depletion Code, ORNL-5621, Oak Ridge National Laboratory, July 1980. 

[SRDB Ref. ID: 23569]

DOE 2003. Source Term Estimates for DOE Spent Nuclear Fuels, U.S. Department of Energy, 

DOE/SNF/REP-078, Rev. 0, March 2003.

DOE 2014. 400 Area/Fast Flux Test Facility. http://www.hanford.gov/page.cfm/400AreaFFTF. 

(updated 9/3/14)

Gawthrop and Jones 1945. Gawthrop, A., and Jones, K.G., Metal Fabrication – Submerged 

Canning Experience to July 1, 1945, H-119, Hanford Engineer Works, Richland, Washington, 

August 8, 1945. 

GA 2015. TRIGA Reactor, General Atomics, http://triga.ga.com/45years.html. (accessed 

1/30/2015)

Gerber 1995. Gerber, M.S., Hanford's Historic Reactors: Constant Change in the Early Years, 

Washington Historical Society, Columbia Magazine, Spring 1995: Vol. 9, No. 1. 

Heeb 1993. Heeb, C.M., Iodine-131 Releases from the Hanford Site, 1944 through 1947,

Volume 1 – Text, PNWD-2033 HEDR Vol. 1, Battelle, Pacific Northwest Laboratory, Richland, 

Washington, March 1993. 

INL 2008. Advanced Test Reactor National Scientific User Facility User’s Guide, INL/EXT-08-

14709. 

INL 2015. http://atrnsuf.inl.gov/default.aspx?Page=Reactors&cat=1&Display=5&id=219

(accessed 1/29/15). 

Kelly 1945. Kelly, J.H., Metal Fabrication – Machining Experience to July 1, 1945, H-113, 

Hanford Engineering Works, Richland, Washington, July 30, 1945. 

Ludwig, S., and Reiner J.P., 1989. Standard- and Extended-Burnup PWR and BWR Reactor 

Models for the ORIGEN2 Computer Code, ORNL/TM-11018, December 1989. 

ORAUT 2007. Fission and Activation Product Assignment for Internal Dose-Related Gross 

Beta and Gross Gamma Analyses, ORAUT-OTIB-0054, Rev. 0 PC-1, November 19, 2007.

ORAUT 2013. Fission and Activation Product Assignment for Internal Dose-Related Gross 

Beta and Gross Gamma Analyses, ORAUT-OTIB-0054, Rev. 1, June 13, 2013.
Effective Date:

February 6, 2015

Revision No.:

0 – Draft

Document No.: Response to Comments on Findings 1–4

of SCA-TR-PR2014-0084 Review

Page No.

21 of 21

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution.

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82.

ORAUT 2014a. Fission and Activation Product Assignment for Internal Dose-Related Gross 

Beta and Gross Gamma Analyses, ORAUT-OTIB-0054, Rev. 2, March 6, 2014.

ORAUT 2014b. Supporting Calculations for ORAUT-OTIB-0054 and ORAUT-RPRT-0047, 

ORAUT-RPRT-0067, Rev. 0, August 26, 2014. 

ORNL 2015. http://scale.ornl.gov/. (accessed 2/6/15)

SC&A 2008. Review of ORAUT-OTIB-0054, “Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross Gamma Analyses,” Rev. 0, PC-1, SCA-TRTASK3-0006, Rev. 0, March 17, 2008. 

SC&A 2013. Review of ORAUT-OTIB-0054, “Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross Gamma Analyses,” Revision 1, SCA-TR-PR2013-

0084, Rev. 0, November 5, 2013.

SC&A 2014. SC&A Review of NIOSH Responses to SC&A Comments on ORAUT-OTIB-0054, 

Rev. 1: “Fission and Activation Product Assignment for Internal Dose-Related Gross Beta and 

Gross Gamma Analyses,” SCA-TR-PR-2014-0084, April 10, 2014.

Schmittroth, F.A. and Lessor, D.E., 1996. ORIGEN2 Calculations Supporting TRIGA Irradiated 

Fuel Data Package, WHC-SD-SNF-ANAL-010, Rev. 0, Westinghouse Hanford Company, 

September 1996.

Schwartz, R.A., 1997. Modifications to ORIGEN2 for Generating N-Reactor Source Terms, 

PNNL-11555, Pacific Northwest National Laboratory, April 1997. 

Sterbentz, J.W., 1997. Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate 

Parametric data for TRIGA Spent Nuclear Fuels, INEL-96/0482, Idaho National Laboratory, 

March 1997.