Draft Report 

NATIONAL INSTITUTE FOR 

OCCUPATIONAL SAFETY AND HEALTH 

ADVISORY BOARD ON RADIATION AND WORKER HEALTH 

Review of ORAUT-OTIB-0054, 

Fission and Activation Product Assignment for Internal Dose-Related 

Gross Beta and Gross Gamma Analyses, Revision 1 

Contract No. 200-2009-28555

SCA-TR-PR2013-0084, Revision 0 

Prepared by 

Stephen Ostrow 

John Mauro 

Joyce Lipzstein 

Ron Buchanan 

SC&A, Inc. 

1608 Spring Hill Road, Suite 400 

Vienna, VA 22182 

November 2013 

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. 
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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 No.

SCA-TR-PR2013-0084

Effective Date:

Draft – November 5, 2013

Revision No:

0 – Draft

Review of ORAUT-OTIB-0054,

Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross 

Gamma Analyses, Revision 1

Supersedes:

N/A

Task Manager:

_____________________ Date: ____________

Stephen L. Ostrow, PhD

Reviewers:

 John Stiver

 John Mauro

 Stephen Ostrow

 Joyce Lipzstein

Project Manager:

_____________________ Date: ____________

John Stiver, CHP
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TABLE OF CONTENTS 

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

Executive Summary .........................................................................................................................6

Review Methodology ...........................................................................................................7

Summary of Findings ...........................................................................................................8

1.0 SC&A Evaluation of ORAUT-OTIB-0054, Revision 1 ....................................................10

1.1 Reactor Modeling...................................................................................................10

1.2 Intake Fractions ......................................................................................................16

1.3 Instructions to Dose Reconstructor ........................................................................21

1.4 Example Problems .................................................................................................22

1.4.1 Example 1: Gross Beta Bioassay Data (with Major Chemical 

Processing) from Waste Management Worker ..........................................22

1.4.2 Example 2: Gross Beta Bioassay Data (with Minor Chemical 

Processing) from Worker Involved in Spent Fuel Storage Operations......23

1.5 Worksheet ..............................................................................................................23

1.6 Assessment of Overall Technical Basis and Usability of OTIB ............................25

2.0 Procedure Checklist ...........................................................................................................26

3.0 References ..........................................................................................................................28

LIST OF TABLES 

Table 1. Reactor Categories and Representative Reactors .............................................................7

Table 2. Review Findings ...............................................................................................................8

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

Table 4. ORIGEN-S Irradiation Parameters for the Nine Representative Reactor Cases ............12

Table 5. Fission and Activation Product Decay Times for General Steps in the Fuel Cycle .......16

Table 6. ATR 1 Radionuclide Intake Rates Derived from Sr-90 Intake Rate of 76.0 pCi/d ........24

Table 7. Results of OTIB-0054 Workbook 1.01 for Example 1: Radionuclide Intake 

Ratios Derived from Sr-90 Intake ...............................................................................24
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ABBREVIATIONS AND ACRONYMS 

Al aluminum 

ATR Advanced Test Reactor 

d day 

DCF dose conversion factor 

DOE U.S. Department of Energy 

FFTF Fast Flux Test Facility 

g gram 

GA General Atomics 

HEPA high efficiency particulate air 

IMBA Integrated Modules for Bioassay Analysis 

INEL Idaho National Engineering Laboratory 

INL Idaho National Laboratory 

IREP Interactive RadioEpidemiological Program 

IRF intake retention fraction 

kg kilogram 

kW kilowatt 

MDA minimum detectable activity 

MDL minimum detectable level 

MTHM metric ton of heavy metal (p. 13) 

MTU metric ton unit 

MW megawatt 

MWd megawatt-day 

n/cm2-s neutron per square centimeter per second (p. 13) 

NIF normalized intake fraction 

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 

pCi picocurie 

PIC pocket ion chamber 
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POC probability of causation 

PWR Pressurized Water Reactor 

RF release fraction 

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

SRS Savannah River Site 

SS stainless steel 

Sv/Bq Sievert per Becquerel 

T1/2 half-life 

TLD thermoluminescent dosimeter 

TRIGA Training Research and Isotope Production, General Atomics 

y year 
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EXECUTIVE SUMMARY 

This report presents a technical review of Fission and Activation Product Assignment for 

Internal Dose-Related Gross Beta and Gross Gamma Analyses, ORAUT-OTIB-0054, Rev. 1, 

June 13, 2013 (ORAUT 2013, hereafter referred to as “the OTIB” or “OTIB-0054”). SC&A had 

previously reviewed Rev. 0 of the OTIB (ORAUT 2007) in 2008 (SC&A 2008) and its findings 

were subsequently discussed in several communications between SC&A and NIOSH and at 

Subcommittee on Procedures Review meetings. However, since the technical basis of the Oak 

Ridge Associated Universities Team (ORAUT) Technical Information Bulletin (OTIB) was 

substantially revised in Rev. 1 of the OTIB, the Subcommittee at its July 18, 2013, meeting 

authorized SC&A to perform a full review of the latest revision. This review is limited to those 

aspects of the OTIB judged by SC&A to be potentially significant to dose reconstruction, and is 

not a line-by-line check of data sources and calculational results. 

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, OTIB-0054 

provides guidance and a standard approach to the dose reconstructor on how to assign 

radionuclide-specific intakes to exposed workers, as follows: 

Interpretation of the results of these gross analyses of air samples requires 

knowledge of the activity ratios of the various fission and activation products in 

the air breathed by the worker and the beta or gamma yield of each product as 

measured by the detector. Interpretation of gross urinalysis results requires 

knowledge of the ratios of the products in urine at some time after the start of 

chronic intake and whether those ratios changed due to sample processing 

activities (ORAUT 2013, Section 4.0). 

The OTIB, however, states that its guidance should only be used in the absence of, or to 

supplement, site- and personnel-specific data that may be available for a particular dose 

reconstruction. The OTIB contains several major parts: 

 A lengthy and detailed explanation of how the representative reactors are modeled and 

fission and activation product activities determined, resulting in the selection of four 

reactors and four decay times (Section 5) 

 Determination of intake fractions and identification of dosimetrically significant 

radionuclides, along with reduction of the size of the radionuclide dataset (Section 6) 

 Determination of activity and intake fraction values to assign intakes from gross gamma 

or beta assay data (Section 7) 

 Instructions to the dose reconstructor on the use of the OTIB (Section 8) 

 Examples of how to apply the OTIB (Attachment H) 

The OTIB is applicable to a broad range of reactor and fuel types, operations, and situations, 

with several important exclusions and caveats as well as some possible extensions. The 
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methodology does not apply to alpha-emitting radionuclides in irradiated fuel and does not 

include activation products produced outside the fuel and its cladding. It also does not apply to 

operations involving decay times of less than 10 days and to cases where radionuclides have 

been deliberately extracted and concentrated for other purposes. The OTIB notes, however, that 

the guidance can also be extended to those cases where in-vivo or fecal bioassay data are 

available, or where data are available for exposures from some, but not all, relevant 

radionuclides. 

NIOSH intends for the OTIB to be broadly applicable by encompassing a number of different 

reactor types and operations that were commonly encountered in the government nuclear 

program. The following table summarizes the five reactor categories considered along with the 

actual reactors chosen to represent those categories. 

Table 1. Reactor Categories and Representative Reactors 

Reactor Category Representative Reactor

Plutonium Production Reactors Hanford N-Reactor

Hanford Single-Pass Reactors

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

Advanced Test Reactors (high enrichment) Advanced Test Reactor (ATR)

Research Reactors TRIGA Reactor (Al-clad fuel)

TRIGA Reactor (SS-clad fuel)

Generic Reactor Pressurized Water Reactor (PWR)

REVIEW METHODOLOGY 

SC&A reviewed the OTIB in a systematic manner, following the OTIB’s organization sectionby-section, including the provided examples and associated workbook, and following the 

guidance of SC&A’s review procedure, A Protocol for the Review of Procedures and Methods 

Employed by NIOSH for Dose Reconstruction (SC&A 2004). Section 1 of this review report 

summarizes the OTIB sections and presents SC&A’s findings and other observations, Section 2 

presents a checklist taken from SC&A 2004, and Section 3 contains reference citations. In a few 

cases, as noted in the report, SC&A independently checked some of the assumptions and tabular 

values appearing in the OTIB. The checklist has sections concerned with the following issues: 

 The degree to which the procedure supports a process that is expeditious and timely for 

dose reconstruction 

 Whether the procedure provides adequate guidance to be efficient in instances where a 

more detailed approach to dose reconstruction would not affect the outcome 

 The extent to which the procedure accounts for all potential exposures, and to ensure that 

resultant doses are complete and based on adequate data 

 Whether the procedure provides a consistent approach to dose reconstruction, regardless 

of claimant’s exposures by time and employment locations

 Whether the procedure is fair and gives the benefit of the doubt to the claimant 

 Whether the procedure adequately accounts for the uncertainty of dose estimates 
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 Whether the procedure strikes a balance between technical precision and process 

efficiency 

In addition, the report assesses whether the overall process presented in the OTIB is reasonable 

from a technical point of view, is claimant favorable, and provides clear guidance to the dose 

reconstructor attempting to use it. 

SUMMARY OF FINDINGS 

SC&A’s detailed review of the OTIB is presented in Section 1. In the interest of clarity, SC&A 

reviewed the OTIB “linearly,” following the OTIB’s development section-by-section. The 

review produced 10 Findings, listed in Table 2. In addition, Section 1 also identifies suggested 

areas where the OTIB can be improved, as well as comments supportive of the OTIB. Further, 

SC&A included two “Notes” (Section 1.2) pointing out apparent typos in the OTIB. 

Table 2. Review Findings

Finding No. 

and Report 

Section

Finding Summary

1

(Sect. 1.1)

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.

2

(Sect. 1.1)

The OTIB does not provide sufficient information to allow evaluation of its downselect from the 

initial seven to the final four representative reactors chosen. 

3

(Sect. 1.1)

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.

4

(Sect. 1.1)

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. 

5

(Sect. 1.2)

In selecting release fractions for exposures to airborne radionuclides associated with reactor 

operations, the OTIB starts with the fuel inventory. However, it might have been more 

appropriate to use the mix of radionuclides in the gas gap or primary coolant as the starting point 

for assigning the isotopic composition in urine samples. Also, if a worker was involved in 

handling waste streams, such as ion exchange resins or HEPA or charcoal filters, using the 

isotopic mix in fuel as the starting point might not be appropriate and might lead to non-claimantfavorable results. These issues should be addressed in the OTIB.
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Table 2. Review Findings

Finding No. 

and Report 

Section

Finding Summary

6

(Sect. 1.2)

The use of effective dose conversion factors (i.e., DCFs that relate to effective whole-body dose) 

is appropriate for screening purposes if the objective of the OTIB was to reconstruct whole-body 

doses, but not necessarily claimant-favorable for organ doses. For example, a radionuclide that 

does not contribute significantly to the whole-body dose could be an important contributor to an 

organ dose and might be eliminated. For certain reactor scenarios, some radionuclides that were 

not present in Table E contribute to the intake in a significant way and can deliver important 

contributions to organ doses. The OTIB would benefit from some discussion of this matter.

7

(Sect. 1.2)

Intakes and organ doses should be calculated using the same set of radionuclides as used to derive 

the contributions to the total beta excretion rate results.

8

(Sect. 1.2)

The OTIB explains that it was recognized that some of the methods used to determine the 

beta/gamma concentration of fission and activation products in urine would miss certain 

radionuclides, such as radioiodines. The OTIB claims that this is not a problem because the 

isotopic assignments of intake are based on the predicted relative concentrations of 17 

radionuclides in air. Hence, these radionuclides are not missed when deriving doses. This seems 

reasonable, except if a large fraction of the activity is lost during the analysis of the urine samples. 

This, of course, would result in an underestimate of the actual gross beta/gamma composition of 

the urine, which would underestimate the radionuclide intake. The OTIB states that this issue was 

taken into consideration, and it was found to be important at the Savannah River Site (SRS). For 

this reason, a separate protocol is used in the SRS site profile. It is not apparent, however, how 

the dose reconstructor deals with situations where the airborne mix of radionuclides that might be 

associated with reactor operations or maintenance bears little resemblance to the mix of 

radionuclides in the fuel. 

9

(Sect. 1.5)

The current OTIB workbook (Workbook 1.01) needs to be revised to match the current version of 

OTIB-0054 (Rev. 1), and then re-evaluated.

10

(Sect. 1.6)

Our primary concern with this OTIB is that although NIOSH developed a protocol that simplifies 

and likely overestimates the radionuclide intakes for individual workers, the protocol seems to be 

somewhat arbitrary when applied to a particular worker. In the process of developing the 

protocol, indicator radionuclides are used to derive intake values of the dosimetrically significant 

radionuclides that do not necessarily relate to the real intakes and excretion rates for any 

given worker. NIOSH does not show the degree of realism or conservatism built into the dose 

reconstruction for a given worker due to the scenario that was assumed by all other assumptions 

taken to derive the values in the tables presented in Section 7. The methods described in the 

OTIB will provide intakes and doses not necessarily correlated with the real ones. The 

differences between the intakes provided through the use of the document and the real ones are 

unknown and depend heavily on the scenario (periods of fuel irradiation and decay), the reactor 

type, and detection methods. It could be argued that as long as the protocol is scientifically valid 

and claimant favorable for a given worker, the approach is consistent with the letter and intent of 

the rule. However, we can envision a situation where two workers are assigned a dose using the 

OTIB, where in one case, the protocol is extremely claimant favorable and the worker is 

compensated, and in the other case, the protocol is less claimant favorable and the claim is denied. 

The OTIB would benefit from a discussion of this particular concern.
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1.0 SC&A EVALUATION OF ORAUT-OTIB-0054, REVISION 1 

OTIB-0054 is quite complex and consists of a number of sections, first describing the process of 

developing the tables used by a dose reconstructor, then providing guidance to the dose 

reconstructor on how to estimate doses for a variety of different situations. Hence, we found it 

beneficial to explain our understanding of the OTIB as part of the process of making comments 

on it. For convenience and clarity, we generally followed the OTIB document section 

framework, but grouped our discussion into three primary sections on reactor modeling, 

determination of intake fractions, and application instructions to the dose reconstructor. In 

addition to reviewing the body of the OTIB, SC&A also reviewed the examples provided in an 

appendix and the associated worksheet provided on the NIOSH computer system. 

1.1 REACTOR MODELING 

The goal of the modeling portion of the OTIB 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 for this OTIB is the radionuclide mix in spent fuel for a 

number of different reactor types, each with different operating conditions (specific power, 

irradiation time, and burnup), calculated at several different decay times following removal from 

the reactor. It is assumed that an actual claimant case condition will fit somewhere within the 

parameter space defined by the OTIB. Section 5 of the OTIB begins 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 Representative Reactor

Plutonium Production Reactors

Hanford N-Reactor

Hanford Single-Pass Reactors

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

Advanced Test Reactors (high enrichment) Advanced Test Reactor (ATR)

Research Reactors

TRIGA Reactor (Al-clad fuel)

TRIGA Reactor (SS-clad fuel)

Generic Reactor Pressurized Water Reactor (PWR)

Source: ORAUT-OTIB-0054, Rev. 1, Table 5-1 

The OTIB uses the ORIGEN2 Version 2.1 isotope generation and depletion code (Croff 1980) to 

calculate isotopic inventories for the 7 representative reactors of Table 3 in 11 runs (4 reactors 

had 2 runs each, using 2 different sets of cross section and fission product yield libraries), 

decayed for 10 days, 40 days, 60 days, 90 days, 180 days, 1 year, and 3 years after irradiation 

had ceased. 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. 
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The OTIB describes how the reactors were modeled in ORIGEN2. Material compositions were 

given based on a single fuel element or per metric ton of fuel and burnup, and “nominal values 

for burnup and specific power were selected as appropriate given the composition basis.” 

Finding 1: 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. 

The OTIB notes that, “No decay (shutdown) intervals or other power history considerations were 

included. Activity ratios for short-lived radionuclides, are, therefore, generally overstated.” 

SC&A concurs with this statement. 

ORIGEN2 is a widely-used, 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. 

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 2 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 (as stated in Note 1 of the OTIB, that 

choice of decay time minimizes loss of short-lived radionuclides). 

The OTIB states, “Based on comparison of the fission and activation product relative activity 

data for the 11 ORIGEN2 runs, 4 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, and the TRIGA reactor.”

Finding 2: The OTIB does not provide sufficient information to allow evaluation 

of its downselect from the initial seven to the final four representative reactors 

chosen. 

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 is a more modern version of ORIGEN2 and is part of the SCALE code system 

(ORNL 2011) for nuclear safety analysis and design developed and maintained by ORNL. Other 

modules in the SCALE system were used to create case-specific cross section libraries and 

preprocess data for ORIGEN-S. Multiple sets of irradiation parameters were defined for each of 

the four representative reactors, resulting in nine ORIGEN cases; parameters (specific power, 

irradiation time and burnup) for the nine resulting cases are summarized in Table 5-2 of the 

OTIB. Bases for the nine cases are also listed in the table, illustrating how the selections attempt 

to span the potential ranges of the parameters for each reactor. 
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Finding 3: 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. 

Notwithstanding the preceding finding, SC&A attempted to verify the parameters used in the 

ORIGEN-S runs as best it could using the information provided in the OTIB as well as in several 

outside sources. Table 5-2 of the OTIB is reproduced here as Table 4 for convenience. 

Table 4. ORIGEN-S Irradiation Parameters for the Nine 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” 

Source: ORAUT-OTIB-0054, Rev. 1, Table 5-2 

SC&A made a quick check by multiplying the specific power by the irradiation time for each 

case, which should 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) = 
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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.

1,794 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. 

The four representative reactors will now be examined individually. 

Advanced Test Reactor (ATR) 

The ATR, which is a pressurized, light-water moderated, beryllium-reflected research reactor, 

has operated at Idaho National Laboratory (INL) for over 45 years (it is still in operation). The 

reactor uses highly enriched uranium fuel (93.15% nominal enrichment) arranged in curved fuel 

element plates, and is run at very high burnup levels to permit accelerated testing of materials for 

radiation damage. An INL website states the following: 

The Advanced Test Reactor (ATR) is located at the ATR Complex on the Idaho 

National Laboratory site and has been operating continuously since 1967. The 

primary mission of this versatile facility was initially to serve the U.S. Navy in the 

development and refinement of nuclear propulsion systems, however, in recent 

years the ATR has been used for a wider variety of government and privately 

sponsored research. DOE’s designation of the ATR as a National Scientific User 

Facility enables DOE to facilitate research that would not have been feasible for 

many researchers and allows for fuller utilization of the facility. 

The ATR design exploits a unique serpentine core configuration that offers a 

large number of test positions. It has large test volumes in high-flux areas, and is 

designed to permit simulation of long neutron radiation exposures in a short 

period of time. The maximum thermal power rating is 250 MW, with a maximum 

unperturbed thermal neutron flux of 1.0 × 1015 n/cm2

-s. [INL 2013]

SC&A consulted Source Term Estimates for DOE Spent Nuclear Fuels (DOE 2003) for technical 

information about the ATR. The document 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. DOE 2003 explicitly estimates the 

ATR spent fuel source terms as Template 12. The DOE report’s upper-end burnup of 

314,683 MWd/MTHM (megawatt-day per metric ton of heavy metal) corresponds to the burnup 

used in the OTIB. The OTIB modeled two power level and irradiation time cases at maximum 

burnup: Case ATR 1 simulates a nominal power level with a long irradiation time; and Case 

ATR 2 simulates a maximum power level with a short irradiation time. The third case, ATR 3, 

simulates a nominal power level with an intermediate irradiation time producing a lower fuel 

burnup. 

DOE 2003 (p. A-104) lists 10 MW as the maximum power level per fuel element and the total 

uranium (U-235, U-238, U-234, U-236) loading in a fuel element as 1166.89 g. The latter is 

equivalent to 0.001167 MTU. The specific power can then by found by dividing the power level 

by the fuel loading: 

10 MW/0.001167 MTU = 8,569 MW/MTU. 
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The OTIB ATR 2 case (maximum burnup at maximum power) of 8,651 MW/MTU is reasonably 

close to this. However, it is not evident what assumptions the OTIB used to derive the ATR 1 

case (maximum burnup at nominal power) and the ATR 3 case (nominal burnup at nominal 

power) values. 

Fast Flux Test Facility (FFTF) 

The FFTF was a 400 MWth, liquid sodium-cooled, FFTF, located at the DOE Hanford site that 

operated from April 1982 to April 1992 (DOE 2007). 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 OTIB considers two cases: FFTF 1, with a maximum burnup of 152,230 MWd/MTHM at 

nominal power; and FFTF 2, with a nominal burnup of 79,984 MWd/MTHM at nominal power. 

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. Hence, the OTIB choice of burnups appears to have adequately 

bracketed the FFTF operating conditions. 

DOE 2003 (Table 3 and p. A-6) gives the average power for a standard fuel assembly as 

5.4 MW. Dividing the average power by the specific power of 163.8 MW/MTHM assumed by 

the OTIB, yields the heavy metal fuel loading in a fuel assembly of 0.03297 MTHM/assembly 

(or, 32,970 g heavy metal/assembly). By comparison, DOE 2003 (Table 1) lists the total heavy 

metal composition (Pu-239, Pu-240, Pu-241, Pu-242, Am-241, U-235, and U-238) of a FFTF 

fuel assembly as 32,918 g, which supports the OTIB value used in its reactor modeling. 

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, were largely taken from 

Bergsman 1994; the latter 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-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 have a preoperations enrichment of 0.947% U-235 in both of its concentric tubes, while the Mark IA 

elements have 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, contain 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 
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radioactive inventory determination on the Mark IV elements. The total fuel loading is 

11.6 MTU (DOE 2003, p. A-53). 

The OTIB selects a representative burnup of 1,188 MWd/MTHM for the first case, N-Reactor 1, 

and a greater value, 2,970 MWd/MTHM, for the second case, N-Reactor 2. The former value is 

given in Schwarz 1997 for an outer Mark IV fuel element, and it results in a discharge Pu-240 

content of 6%. As far as SC&A could check, the OTIB appears to have taken its data for the 

N-Reactor from authoritative sources. 

TRIGA 

The low-power, water-cooled, pool-type 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 raise 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 clad either with aluminum or stainless steel (General Atomics). 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. Pertinent information for 

constructing an ORIGEN model can be found in the INL report, Sterbentz (1997), and the 

Westinghouse Hanford report, Schmittroth and Lessor (1996). An aluminum-clad fuel element 

contains 180 g of uranium, and a stainless steel-clad fuel element contains 195 g of uranium 

(Sterbentz 1997). 

Finding 4: 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. 

In the next step, the OTIB reduces decay times under consideration from the initial seven to four 

representative ones: 10 days, 40 days, 180 days, and 1 year. OTIB Table 5-3, reproduced here 

as Table 5, provides guidance to the dose reconstructor for choosing an appropriate decay time 

for specified worker activities. 
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Table 5. Fission and Activation Product Decay Times for General Steps in the Fuel Cycle 

Activity Decay Time

Reactor operations, spent fuel storage, fuel examination; activity from fuel leakage/failure 

or segmentation

10 days

Dissolution of fuels from early production reactors (e.g., Hanford production reactors 

1940s–1950s)

40 days

Fuel dissolution; general or later years 180 days

Waste management 1 year

Source: ORAUT-OTIB-0054, Rev. 1, Table 5-3 

Hence, after a number of steps, the OTIB has transformed the original set of cases to four 

different types of reactors, each with different operating conditions, and four different decay 

times. These are the reference reactors that will be used by dose reconstructors as the starting 

point for reconstructing the doses to a given claimant with a given type of cancer.

1.2 INTAKE FRACTIONS 

For convenience and clarity in the following discussion, SC&A informally labels different 

portions of the OTIB as “Step 1,” “Step 2,” etc. 

Step 1 

As discussed above, the starting point for this OTIB (described in OTIB Section 5) is the 

derivation of the radionuclide mixture in spent fuel for four different representative reactors, 

each with different operating conditions (specific power, irradiation time, and burnup), and four 

different decay times (10 days, 40 days, 180 days, and 1 year). The dose reconstructors will use 

these reference reactors as a basis for reconstructing the doses to a given claimant with a given 

type of cancer. 

Given the isotopic mixtures of Section 5 of the OTIB, Section 6 explains how the list was 

reduced through a multi-step process. First, radionuclides that could not contribute significantly 

to the internal dose, such as very short-lived radionuclides and noble gases, were removed from 

the list. Then, the activities of the remaining radionuclides were multiplied by the following 

assumed release fractions (RFs): 1.0 for tritium; 0.5 for volatiles, such as radioiodines; and 0.01 

for semi-volatiles, like Cs-137. For particulates and solids, the OTIB uses an RF of 0.01, which 

is cited in the report as being conservative, since solids are likely to have release fractions of 

0.001 or less. The OTIB cites DOE 1997, Hazard Categorization and Accident Analysis 

Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, as the 

basis for these release fractions. Appendix B of the OTIB presents the assumed RFs for the 

radionuclides under consideration. The outcome of this step is an expression of the activity of 

each radionuclide that is likely to be airborne. This process is performed for each of the nine 

categories of fuel and four different decay periods. 

SC&A believes that these RFs, although crude, are reasonable and will likely result in a 

claimant-favorable reconstruction of the isotopic mix of radionuclides in urine, assuming the 

worker is involved in fuel dissolution processes. However, as discussed below, it is not apparent 
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that using the core inventory as the starting point for deriving doses is always scientifically 

appropriate and claimant-favorable. 

Note 1: There appears to be a typo in the RF for iodines in Attachment B of the 

OTIB. Appendix B lists an RF of 0.05, but we believe 0.5 is the value that the 

OTIB intends to use. 

Finding 5: In selecting release fractions for exposures to airborne radionuclides 

associated with reactor operations, the OTIB starts with the fuel inventory. 

However, it might have been more appropriate to use the mix of radionuclides in 

the gas gap or primary coolant as the starting point for assigning the isotopic 

composition in urine samples. Also, if a worker was involved in handling waste 

streams, such as ion exchange resins or HEPA or charcoal filters, using the 

isotopic mix in fuel as the starting point might not be appropriate and might lead 

to non-claimant-favorable results. These issues should be addressed in the OTIB. 

Step 2 

It is our understanding that the mix of airborne radionuclides obtained in Step 1 was then 

expressed in terms of fractions of the total activity, where the sum of the assigned fractions for 

each set of airborne activities was set equal to 1.0. Section 6.2 of the OTIB provides a 

description of the methods used to further reduce the large number of radionuclides present in 

irradiated fuel to a manageable number (without compromising the dose reconstruction) by a 

two-step process. The first cut was to multiply the release fraction of each radionuclide by the 

50-year effective dose commitment coefficient (e.g., Sv/Bq inhaled) taking into consideration the 

different solubility types of each radionuclide. We assume that, for each reactor type and decay 

time, this step in the process results in the product of a release fraction with a DCF, which in turn 

is expressed in terms of a metric referred to as a dosimetrically significant source term. Each of 

these metrics was again converted to a fraction, and is presented in Attachment D. A total of 36 

isotopes are included in the tables of this attachment. This list was further reduced by 

eliminating those radionuclides that contributed less than 0.01 to what is referred to as the 

normalized intake fractions (NIFs). Attachment E presents the revised list of NIFs, which now 

has been reduced to 17 radionuclides. 

Finding 6: The use of effective dose conversion factors (i.e., DCFs that relate to 

effective whole-body dose) are appropriate for screening purposes if the objective 

of the OTIB were to reconstruct whole-body doses, but not necessarily claimant 

favorable for organ doses. For example, a radionuclide that does not contribute 

significantly to the whole-body dose could be an important contributor to an 

organ dose and might be eliminated. For certain reactor scenarios, some 

radionuclides that were not present in Table E contribute to the intake in a 

significant way and can deliver important contributions to organ doses. The 

OTIB would benefit from some discussion of this matter.
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Step 3 

It is our understanding that, for each set of normalized intake fractions using Cs-137 and Sr-90 as 

the normalizing (indicator) radionuclides, the relative concentrations of each radionuclide in 

urine assuming a chronic intake for 2 years, followed by a 24-hour urine sample, was calculated 

and given in Attachment G. However, these are theoretical radionuclide concentrations. In 

reality, processing of urine samples requires some time, where some ingrowth and decay might 

occur, and a certain amount of some radionuclides might be lost in the preparation of the sample 

for counting. 

The OTIB makes adjustments to the relative radionuclide concentrations in urine to account for 

this ingrowth and decay. The NIFs from Table D-1 are multiplied by the corresponding intake 

retention fractions (IRFs) of Table F-1 (obtained from the Integrated Modules for Bioassay 

Analysis (IMBA) program for the least possible solubility type for all radionuclides, except for 

Sr-90 for which Type F was used), and then adjusted according to the physical decay and 

ingrowth during the delay time between sample collection and counting time. The maximum 

delay time considered was 30 days. The half-lives (T1/2) for decay corrections are given in 

Table F-2. 

Note 2: The T1/2 for Cs-136 in Table F-3 is listed as 13.16 y instead of the correct 

13.16 d. 

Tables G-1, G-2, and G-3 give the relative radionuclide concentrations in urine at the time of 

gross beta or gross gamma counting. Tables G-4 and G-5 list the contributions from Table D-1 

radionuclides to gross beta and gross gamma activity, respectively, for air concentration or 

surface concentration. 

Step 4 

For each reactor scenario considered, the contribution of Sr-90 to the total beta activity of a urine 

sample collected after 2 years of chronic inhalation intakes is calculated using the values in 

Tables G-1 and G-2 and then divided by the total beta contributions of all the other radionuclides 

of that scenario to yield a fractional value. In a similar fashion, the relative gamma activity of 

Cs-137 in the urine sample collected after 2 years of chronic inhalation intakes is calculated 

using the values of Table G-3. 

Table 7-1a defines the contribution of Sr-90 to gross beta urinalysis for minimally processed 

samples, Table 7-1b defines the contribution of Cs-137 to gross gamma urinalysis for minimally 

processed samples, and Table 7-2 shows the contribution of Sr-90 to the gross beta urinalysis 

following major chemical processing. The activity fractions in Tables 7-1a and 7-2 are used as 

appropriate to determine Sr-90 contribution to gross beta analysis results in all urine samples 

taken from a worker. In the same way, Table 7-1b is used to calculate the Cs-137 activity 

contribution to the gross gamma result in all urine samples taken from a worker. 

SC&A Calculation: SC&A performed a sample calculation to enhance its understanding of the 

procedure and to check the use of the tables to determine the beta activity fraction. The problem 

assumes a worker is exposed chronically for 2 years through inhalation to Sr-90 from case 
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ATR 1 fuel with 10-day decay [ATR 1 (10d)]. A 24-hour urine sample is taken on the last day 

of exposure and is minimally processed before it is counted 30 days after sample collection. 

Note that for a delay of 30 days between sample collection and counting, the adjustment factor 

for Sr-90 is 1.0. 

The excretion of Sr-90 in urine at the end of those 2 years can be expressed as: 

 u(Sr-90, 2y) = 0.231 × I(Sr-90) (1) 

where, 

0.231 is the Intake Retention Function (IRF) for Sr-90 from Table F-1, and 

I(Sr-90) is the Sr-90 intake. 

For the ATR 1 (10d) case: 

I(Sr-90) = 4.01 × 10-4 × I(total) (2)

where, 

4.01 × 10-4 is the Sr-90 Normalized Intake Fraction (NIF) from Table D-1, and

I(total) is the total daily beta activity intake for a worker exposed in the ATR 1 (10d) 

scenario 

Then, substituting Eq. (2) in to Eq. (1), 

u(Sr-90, 2y) = 0.231 × 4.01 × 10-4 × I(total) = 9.24 × 10-5 × I(total) (3) 

Note that the value of 9.24 x 10-5 calculated in Eq. (3) corresponds to the “contribution to gross 

beta activity for an unprocessed urinalysis” value for Sr-90 in the ATR 1 (10 d) scenario 

appearing in Table G-1. 

For each radionuclide in Table G-1, the corresponding urine excretion of a particular 

radionuclide after 2 years chronic exposure is:

u(radionuclide, 2y) = I(total) × value for that radionuclide in the first column of 

Table G-1 (i.e., the contribution to gross beta activity for that radionuclide in the 

ATR 1 (10d) scenario). (4) 

The total beta activity in urine after 2-year chronic intake is equal to the sum of the contributions 

of all radionuclides in Table G-1, which is 4.32 × 10-3 for the ATR 1 (10d) scenario (adding 

values in first column of Table G-1). Then, 

u(total beta, 2y)= 4.32 × 10-3 × I(total) (5) 

Solving for I(total) in Eq. (5) gives 

I(total) = u(total beta, 2y)/(4.32 × 10-3) (6) 
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Substituting the expression for I(total) from Eq. (6) into Eq. (3) gives 

u(Sr-90, 2y) = 9.24 × 10-5 × u(total beta, 2y)/(4.32 × 10-3) (7) 

Doing the divsion in Eq. (7) yields 

 u(Sr-90, 2y) = 2.14 × 10-2× u(total beta, 2y) (8)

If a worker has a set of gross beta readings in urine, all taken after 2 years of chronic exposures, 

the contribution from the indicator Sr-90 to the gross beta activity in the urine samples is 

calculated using the activity fractions in Table 7-1a or Table 7-2 (depending on the degree of 

sample processing) to determine the indicator radionuclide activity. IMBA is then applied to the 

set of Sr-90 activities in urine, u(Sr-90, 2y), and the best fit intake of Sr-90 is calculated. Intakes 

for the other radionuclides are then assigned using the appropriate values of Tables 7-3a through 

7-3i. 

Tables 7-3a through 7-3i (for the different reactor scenarios) are based on the reduced list of 

dosimetrically important radionuclides, listed in Table E-1. The OTIB claims that it is claimant 

favorable to assign doses to the radionuclides in Table E-1 instead of to the larger set of 

radionuclides in Table D-1. Even if true, it is not correct to use Table D-1 intakes to derive the 

contributions from each nuclide to total urine excretion and then back-calculate the intake rates 

using a different table, i.e., Table E-1, with a reduced number of nuclides. 

Finding 7: Intakes and organ doses should be calculated using the same set of 

radionuclides as used to derive the contributions to the total beta excretion rate 

results. 

The OTIB does not explain why chronic intakes shorter than 2 years were not considered. For 

example, all the example problems in Attachment H assume a worker whose urine samples were 

taken at a minimum of 2 years after the first day of employment. SC&A tested the assumption 

of 90 days chronic exposure instead of 2 years to see if there is a significant change. For the 

scenario ATR 1 (10d) and 90 days chronic exposure, the fraction of Sr-90 in minimally 

processed samples is equal to 0.0254 instead of 0.0214. Taking into consideration all the 

uncertainties and approximations throughout the procedure, SC&A concludes that there is no 

relevant difference and the assumption of a 2-year chronic intake scenario is acceptable. SC&A 

also determined that the Cs-137 gross gamma activity fraction in urine for minimally processed 

samples for 90 days continuous exposure is lower than that for 2 years’ exposure. Thus, the 

scenario of 2-years continuous exposure assumed by NIOSH is claimant favorable for gammas 

as well.

Step 5 

Each set of chronic isotopic intakes determined as outlined above is then used as input to IMBA 

to derive doses to the organ of concern. The intake that results in the highest organ dose is used 

as input to the Interactive RadioEpidemiological Program (IREP) to derive probability of 

causation (POC). 
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Finding 8: The OTIB explains that it was recognized that some of the methods 

used to determine the beta/gamma concentration of fission and activation products 

in urine would miss certain radionuclides, such as radioiodines. The OTIB claims 

that this is not a problem because the isotopic assignments of intake are based on 

the predicted relative concentrations of 17 radionuclides in air. Hence, these 

radionuclides are not missed when deriving doses. This seems reasonable, except 

if a large fraction of the activity is lost during the analysis of the urine samples. 

This, of course, would result in an underestimate of the actual gross beta/gamma 

composition of the urine, which would underestimate the radionuclide intake. 

The OTIB states that this issue was taken into consideration, and it was found to 

be important at the SRS. For this reason, a separate protocol is used in the SRS 

site profile. It is not apparent, however, how the dose reconstructor deals with 

situations where the airborne mix of radionuclides that might be associated with 

reactor operations or maintenance bears little resemblance to the mix of 

radionuclides in the fuel. 

1.3 INSTRUCTIONS TO DOSE RECONSTRUCTOR 

Section 8 of the OTIB summarizes what the document contains and provides instructions 

(beginning in the third paragraph) to the dose reconstructor on how to apply the methodology 

contained in the OTIB. In short, activity fraction data (Tables 7-1a/b, 7-2, or 7-4a/b) are applied 

to determine the indicator radionuclide activities for a given gross beta or gamma measurement. 

Decay times are selected from consideration of which worker category (Table 5-3) is most 

appropriate for the case. After the appropriate number of indicator radionuclide activity levels 

are found, IMBA and activity fraction ratios are used to determine radionuclide intakes. Finally, 

indicator radionuclide doses are entered into IREP to determine the POC for the cancer under 

consideration. 

The OTIB cautions that site- or claimant-specific information should be used if available: 

This document provides activity and intake fraction values for use in interpreting 

gross beta or gross gamma activity measurements related to reactor source terms. 

The values were derived using representative parameters and models; site- and 

claim-specific information, if available, should take precedence over the guidance 

in this document. This includes site-specific models and guidance in Site Profiles

(ORAUT 2013). 

The OTIB provides significant guidance in cases where there is no specific information 

available: 

If a specific decay time or times cannot be established, the dose evaluation will 

need to consider all four possibilities. Similarly, in most instances, all nine of the 

representative reactor cases will need to be considered. If the dose reconstructor 

does not know if significant separations chemistry was performed prior to a gross 

beta urinalysis, it is favorable to the claimant to assume significant chemistry was 

performed... If both gross beta and gross gamma results are available for a given 
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assay then the more sensitive measure [i.e., that with the lowest minimum 

detectable activity (MDA)] should be used. If the dose reconstructor does not 

know if a given assay result was a gross beta or a gross gamma count (and both 

are realistic choices for the site under consideration) then both should be 

evaluated to determine which yields the larger assigned dose (ORAUT 2013). 

SC&A believes that following the above guidance will result in claimant-favorable outcomes. 

Overall, SC&A believes that, except for a few possible issues, such as those raised regarding the 

starting point for the calculation (i.e., the isotopic inventories of radionuclides in fuel), the 

protocol described in OTIB-0054, Rev. 1, seems reasonable (if our understanding of the protocol 

is correct). 

1.4 EXAMPLE PROBLEMS 

Because of the complexity of the protocol, the OTIB provides three example calculations in 

Attachment H to guide the dose reconstructor in applying the OTIB. The following presents our 

review of two of the example calculations in order to ensure that our understanding of the 

protocol is accurate. 

1.4.1 Example 1: Gross Beta Bioassay Data (with Major Chemical Processing) from 

Waste Management Worker 

In this example, the worker was involved in waste management activities and had a series of 

eight gross beta urine sample analyses from 1964 to 1966 that ranged from less than the 

minimum detectable level (MDL) of 20 pCi/day to 57 pCi/day. The introduction to the example 

explains that: 

...in dose reconstructions intakes and doses to the organ of concern have to be 

calculated for all the reactor cases to determine the one that maximizes the organ. 

This is necessary to justify use of the constant (upper bound) distribution. 

However, the user can choose to determine the intakes for one reactor case only 

and use the OTIB-0054 tool [to] calculate the intakes (and organ doses) for the 

other reactor cases without having to use IMBA more than once. 

In accordance with Table 5-3 of the OTIB, 1-year old fuel is assumed for the waste management 

function. Also, it was determined that the mix applicable to ATR 1 is appropriate, and, since the 

bioassay data are gross beta, Sr-90 Type F is used as the indicator radionuclide. Table 7-2 

(major chemical processing of the urine) indicates that the activity fraction of Sr-90 for 1-year 

old ATR 1 fuel is 0.471 of the gross beta results. Based on these assumptions, and using IMBA, 

the chronic Sr-90 intake rate that yields the derived urine excretion rates reported in Table H-1 of 

the OTIB is 76 pCi/d. This value is used as the normalization factor to derive the intake 

fractions of the other radionuclides (see Table H-1 of the OTIB). 

This exercise was performed for each of the other eight reactor types (by scaling from the ATR 1 

results) to derive the maximum dose to the organ of concern. This approach seems to be 

claimant favorable. 
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1.4.2 Example 2: Gross Beta Bioassay Data (with Minor Chemical Processing) from 

Worker Involved in Spent Fuel Storage Operations 

In this example, the worker was involved in spent fuel storage operations. The bioassay data are 

the same as used in Example 1, but a 10-day decay time is assumed from the listings in 

Table 5-3. The gross beta fraction for Sr-90 in urine is 0.0214 for 10-day old fuel for the ATR 1, 

as listed in Table 7-1a (minor chemical processing of the urine). We understand this to mean 

that the chronic intake of Type F Sr-90 for 2 years results in a relative concentration of Sr-90 in 

urine of 0.0214. Hence, 0.0214 of the gross beta activity excreted in urine in Table H-3 is Sr-90, 

or about 0.0214 times gross beta activity observed in urine or about 0.6 pCi/d. Table H-4 

presents the chronic intake rates for all radionuclides by normalizing against Sr-90 (i.e., 

assuming the relative intake of Sr-90 intake is one and the other radionuclides have an intake 

that is proportional to that of Sr-90, as listed in Table 7-3a of the OTIB). 

The OTIB states that, using IMBA, the intake rate of Sr-90 that results in a urinary excretion rate 

of 0.6 pCi/day is 3.51 pCi/d. As a check, SC&A calculated an intake rate of 3.52 pCi/d using the 

Sr-90 excretion rates, which is in agreement with the OTIB value. 

1.5 WORKSHEET 

As part of reviewing the OTIB, SC&A also evaluated the associate workbook (Workbook 1.01), 

which is used by the dose reconstructors to automate the dose reconstruction process. It should 

be noted that while Rev. 1 of the OTIB was released on June 13, 2013, Version 1.01 of the 

Workbook was released on October 9, 2007 (after Rev. 0 and before Rev. 0, PC-1 of the OTIB). 

SC&A analyzed the three examples provided in the OTIB to verify the methodology and 

compatibility of the examples with the text of OTIB-0054. SC&A then ran the intake 

calculations for these three examples using the OTIB-0054 Workbook 1.01; a sample input 

screen is shown in Figure 1. 

[This figure has been redacted as it contains information protected by the Privacy Act.]

Figure 1. Example of OTIB-0054 Workbook 1.01 Data Input Screen 

After SC&A ran the intake calculations for these three examples using the workbook, it was 

found that some of the radionuclides (Eu-155, Fe-55, etc.) and intake values provided by the 

workbook were different than those listed in the tables in the examples, i.e., Table H-2, H-4, and 

H-5. The resulting values provided by the workbook analyses did not match those listed in the 

corresponding tables for the three examples, nor did the ratios of the radionuclides match. 

For Example 1, the Ru-106/Sr-90 value derived from Table H-2 is 0.917, but the workbook 

returns a corresponding ratio of 3.80. Additionally, the workbook does not list Co-60, but the 

OTIB table did, and the OTIB table does not list Fe-55 while the workbook does. Tables 6 and 7 

help to illustrate this issue. 
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Table 6. ATR 1 Radionuclide Intake Rates 

Derived from Sr-90 Intake Rate of 76.0 pCi/d

Radionuclide 1-yr Ratios

Co-60 1.41E-02

Sr-89 4.02E+01

Sr-90 7.60E+01

Y-90 7.60E+01

Y-91 9.35E+01

Zr-95 1.44E+02

Nb-95 3.10E+02

Ru-103 6.63E+00

Ru-106 6.97E+01

I-131 4.81E-09

Cs-134 4.20E+01

Cs-137 7.68E+01

Ce-141 3.45E+00

Ce-144 9.58E+02

Pr-143 7.90E-05

Pm-147 1.84E+02

Eu-154 1.69E+00

Total Daily Intake 2.08E+03

Source: ORAUT-OTIB-0054, Rev. 1, Table H-2 

Table 7. Results of OTIB-0054 Workbook 1.01 

for Example 1: Radionuclide Intake Ratios 

Derived from Sr-90 Intake

Radionuclide Ratio to Sr-90

Ce-141 2.10

Ce-144 7.80

Cs-134 0.31

Cs-137 1.30

Eu-155 0.08

Fe-55 0.81

Nb-95 3.00

Pm-147 3.10

Ru-106 3.80

Sr-89 0.31

Sr-90 1.00

Y-91 0.77

Zr-95 1.40

Investigating further, SC&A found that columns X, Y, Z, and AA of the workbook list the 

radionuclides and their ratios based on Sr-90 or Cs-137 for a decay period of 10 days, 40 days, 

180 days, and 1 year. These values and ratios match the results the OTIB-0054 workbook 

provided for each example. 

Considering the dates of the issuance of the OTIB-0054 Workbook 1.01 (2007) and the current 

OTIB-0054, R1 (2013), and the Publication Record Description on page 2 of the OTIB, it is 

apparent that the current version of the OTIB is a complete rework of the methodology and 
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reactor data used. Therefore, the current version of the OTIB and the 2007 version of the 

workbook do not provide the same intake values, ratio values, or radionuclide lists. However, 

SC&A’s hand calculations (using the data tables in the OTIB) for the three examples did provide 

the same resulting values as stated in the examples (see Section 1.4). 

Finding 9: The current OTIB workbook (Workbook 1.01) needs to be revised to 

match the current version of OTIB-0054 (Rev. 1), and then reevaluated.

1.6 ASSESSMENT OF OVERALL TECHNICAL BASIS AND USABILITY OF OTIB

Finding 10: Our primary concern with this OTIB is that although NIOSH 

developed a protocol that simplifies and likely overestimates the radionuclide 

intakes for individual workers, the protocol seems to be somewhat arbitrary when 

applied to a particular worker. In the process of developing the protocol, 

indicator radionuclides are used to derive intake values of the dosimetrically 

significant radionuclides that do not necessarily relate to the real intakes and 

excretion rates for any given worker. NIOSH does not show the degree of 

realism or conservatism built into the dose reconstruction for a given worker due 

to the scenario that was assumed and by all other assumptions taken to derive the 

values in the tables presented in Section 7. The methods described in the OTIB 

will provide intakes and doses not necessarily correlated with the real ones. The 

differences between the intakes provided through the use of the document and the 

real ones are unknown and depend heavily on the scenario (periods of fuel 

irradiation and decay), the reactor type, and detection methods. It could be argued 

that as long as the protocol is scientifically valid and claimant favorable for a 

given worker, the approach is consistent with the letter and intent of the rule. 

However, we can envision a situation where two workers are assigned a dose 

using the OTIB, where in one case, the protocol is extremely claimant favorable

and the worker is compensated, and in the other case, the protocol is less claimant 

favorable and the claim is denied. The OTIB would benefit from a discussion of 

this particular concern.
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2.0 PROCEDURE CHECKLIST

In addition to the technical review of the OTIB described elsewhere, SC&A also reviewed it in 

accordance with its procedure, A Protocol for the Review of Procedures and Methods Employed 

by NIOSH for Dose Reconstruction (SC&A 2004). Table 8 is taken from that procedure. Since 

the table has general applicability, not all of its items are germane to the review of the OTIB.

Table 8. Procedure Review Outline/Checklist

No. Description of Objective Rating

1-5* Comments

1.0 Determine the degree to which procedures support a process that is expeditious and timely 

for dose reconstruction

1.1 Is the procedure written in a style that is clear and unambiguous? 4

1.2 Is the procedure written in a manner that presents the data in a 

logical sequence?

5

1.3 Is the procedure complete in terms of required data (i.e., does not 

reference other sources that are needed for additional data)?

5

1.4 Is the procedure consistent with all other procedures that are part 

of the hierarchy of procedures employed by NIOSH for dose 

reconstruction?

N/A This review did 

not examine other 

procedures.

1.5 Is the procedure sufficiently prescriptive in order to minimize the 

need for subjective decisions and data interpretation?

4

2.0 Determine whether the procedure provides adequate guidance to be efficient in instances 

where a more detailed approach to dose reconstruction would not affect the outcome

2.1 Does the procedure provide adequate guidance for identifying a 

potentially high probability of causation as part of an initial dose 

evaluation of a claim?

N/A Not the intention 

of the procedure.

2.2 Conversely, for claims with suspected cumulative low doses, does 

the procedure provide clear guidance in defining worst-case 

assumptions?

N/A Not the intention 

of the procedure.

3.0 Assess the extent to which procedures account for all potential exposures and ensure that 

resultant doses are complete and based on adequate data

3.1 Assess quality of data collected via interviews: N/A

3.1.1 Is scope of information sufficiently comprehensive? N/A

3.1.2 Is the interview process sufficiently flexible to permit unforeseen 

lines of inquiry?

N/A

3.1.3 Does the interview process demonstrate objectivity, and is it free 

of bias?

N/A

3.1.4 Is the interview process sensitive to the claimant? N/A

3.1.5 Does the interview process protect information as required under 

the Privacy Act?

N/A

3.2 Adequacy and use of site-specific data pertaining to: N/A

3.2.1 Personal dosimeters (e.g., film, TLD, PICs) N/A

3.2.2 In-vivo/in-vitro bioassays N/A

3.2.3 Missing dosimetry data N/A

3.2.4 Unmonitored periods of exposure N/A
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Table 8. Procedure Review Outline/Checklist

No. Description of Objective Rating

1-5* Comments

4.0 Assess procedure for providing a consistent approach to dose reconstruction regardless of 

claimant’s exposures by time and employment locations

4.1 Does the procedure support a prescriptive approach to dose 

reconstruction?

5

4.2 Does the procedure adhere to the hierarchical process as defined 

in 42 CFR 82.2?

4

5.0 Evaluate procedure with regard to fairness and giving the benefit of the doubt to the 

claimant

5.1 Is the procedure claimant favorable in instances of missing data? 4

5.2 Is the procedure claimant favorable in instances of unknown 

parameters affecting dose estimates?

4

5.3 Is the procedure claimant favorable in instances where claimant 

was not monitored?

4

6.0 Evaluate procedure for its ability to adequately account for the uncertainty of dose estimates

6.1 Does the procedure provide adequate guidance for selecting the 

types of probability distributions (i.e., normal, lognormal?)

N/A

6.2 Does the procedure give appropriate guidance in the use of 

random sampling in developing a final distribution?

N/A

7.0 Assess procedures for striking a balance between technical precision and process efficiency

7.1 Does the procedure require levels of detail that can reasonably be 

accounted for by the dose reconstructor?

4

7.2 Does the procedure avoid levels of detail that have only limited 

significance to the final dose estimate and its POC?

4

7.3 Does the procedure employ scientifically valid protocols for 

reconstructing doses?

4

* Rating system of 1 through 5 corresponds to the following: 1=No (Never), 2=Infrequently, 

3=Sometimes, 4=Frequently, 5=Yes (Always), N/A indicates not applicable
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3.0 REFERENCES

42 CFR Part 82.2. Methods for Radiation Dose Reconstruction Under the Energy Employee 

Occupational Illness Compensation Program Act of 2000. 

Bergsman 1994. Bergsman, K.H., Hanford Spent Fuel Inventory Baseline, WHC-SD-SNF-TI001, 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 1997. Hazard Characterization and Accident Analysis Techniques for Compliance with 

DOE Order 5480.23, Nuclear Safety Analysis Reports, U.S. Department of Energy, DOE-STD1027-92 Change Notice No. 1, September 1997. [SRDB Ref ID: 33208]

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

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

DOE 2007. Fast Flux Test Facility, U.S. Department of Energy, DOE Richland Operations 

Office website: http://www.hanford.gov/RL/?page=304&parent=0. 

General Atomics. TRIGA Reactor, General Atomics, http://triga.ga.com/45years.html.

INL 2013. ATR: Capabilities, Idaho National Laboratory, 

http://atrnsuf.inl.gov/Capabilities/Reactors/INLAdvancedTestReactor/tabid/66/Default.aspx

(accessed 9/25/13). 

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.

ORNL 2011. SCALE: A Comprehensive Modeling and Simulation Suite for Nuclear Safety 

Analysis and Design, ORNL/TM-2005/39, Version 6.1, Oak Ridge National Laboratory, June 

2011. Available from Radiation Safety Information Computational Center at Oak Ridge National 

Laboratory as CCC-785.

SC&A 2004. A Protocol for the Review of Procedures and Methods Employed by NIOSH for 

Dose Reconstruction, SCA-PR-Task 3, Rev. 1, Final, April 29, 2004. 

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. 
Effective Date:

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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.

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

Supporting TRIGA Irradiated Fuel Data Package, WHC-SD-SNF-ANAL-010, Rev. 0, 

Westinghouse Hanford Company, September 1996.

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

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

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

Rate Parametric data for TRIGA Spent Nuclear Fuels, Idaho National Laboratory, INEL96/0482, March 1997.