NUREG/CR-7251 

ORNL/TM-2018/782

Margins for Uncertainty in 

the Predicted Spent Fuel 

Isotopic Inventories for

BWR Burnup Credit

Office of Nuclear Regulatory Research
AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically access 

NUREG-series publications and other NRC records at 

NRC’s Library at www.nrc.gov/reading-rm.html. Publicly 

released records include, to name a few, NUREG-series

publications; Federal Register notices; applicant, 

licensee, and vendor documents and correspondence; 

NRC correspondence and internal memoranda; bulletins 

and information notices; inspection and investigative 

reports; licensee event reports; and Commission papers 

and their attachments.

NRC publications in the NUREG series, NRC regulations, 

and Title 10, “Energy,” in the Code of Federal Regulations

may also be purchased from one of these two sources.

1. The Superintendent of Documents

U.S. Government Publishing Office

Washington, DC 20402-0001

Internet: bookstore.gpo.gov

Telephone: (202) 512-1800

Fax: (202) 512-2104

2. The National Technical Information Service

5301 Shawnee Road

Alexandria, VA 22312-0002

www.ntis.gov

1-800-553-6847 or, locally, (703) 605-6000

A single copy of each NRC draft report for comment is

available free, to the extent of supply, upon written

request as follows:

Address: U.S. Nuclear Regulatory Commission

Office of Administration

Multimedia, Graphics, and Storage &

Distribution Branch 

Washington, DC 20555-0001 

E-mail: distribution.resource@nrc.gov

Facsimile: (301) 415-2289

Some publications in the NUREG series that are posted 

at NRC’s Web site address www.nrc.gov/reading-rm/

doc-collections/nuregs are updated periodically and may 

differ from the last printed version. Although references to 

material found on a Web site bear the date the material 

was accessed, the material available on the date cited 

may subsequently be removed from the site.

Non-NRC Reference Material

Documents available from public and special technical 

libraries include all open literature items, such as books, 

journal articles, transactions, Federal Register notices, 

Federal and State legislation, and congressional reports. 

Such documents as theses, dissertations, foreign reports 

and translations, and non-NRC conference proceedings 

may be purchased from their sponsoring organization.

Copies of industry codes and standards used in a

substantive manner in the NRC regulatory process are 

maintained at—

The NRC Technical Library

Two White Flint North

11545 Rockville Pike

Rockville, MD 20852-2738

These standards are available in the library for reference 

use by the public. Codes and standards are usually 

copyrighted and may be purchased from the originating 

organization or, if they are American National Standards, 

from—

American National Standards Institute

11 West 42nd Street

New York, NY 10036-8002

www.ansi.org

(212) 642-4900

Legally binding regulatory requirements are stated only in 

laws; NRC regulations; licenses, including technical 

specifications; or orders, not in NUREG-series publications. 

The views expressed in contractor prepared publications in 

this series are not necessarily those of the NRC.

The NUREG series comprises (1) technical and 

administrative reports and books prepared by the staff 

(NUREG–XXXX) or agency contractors (NUREG/CR–XXXX), 

(2) proceedings of conferences (NUREG/CP–XXXX),

(3) reports resulting from international agreements

(NUREG/IA–XXXX),(4) brochures (NUREG/BR–XXXX), and

(5) compilations of legal decisions and orders of the

Commission and Atomic and Safety Licensing Boards and

of Directors’ decisions under Section 2.206 of NRC’s

regulations (NUREG–0750).

DISCLAIMER: This report was prepared as an account 

of work sponsored by an agency of the U.S. Government. 

Neither the U.S. Government nor any agency thereof, nor 

any employee, makes any warranty, expressed or implied, 

or assumes any legal liability or responsibility for any third 

party’s use, or the results of such use, of any information, 

apparatus, product, or process disclosed in this publication, 

or represents that its use by such third party would not 

infringe privately owned rights.
NUREG/CR-7251 

ORNL/TM-2018/782

Margins for Uncertainty in 

the Predicted Spent Fuel

Isotopic Inventories for

BWR Burnup Credit

Manuscript Completed: June 2018 

Date Published: December 2018 

Prepared by: 

I. Gauld

U. Mertyurek

Oak Ridge National Laboratory

Bethel Valley Road

Oak Ridge, TN 37831

Mourad Aissa, NRC Project Manager 

Office of Nuclear Regulatory Research

iii

ABSTRACT

US Nuclear Regulatory Commission staff guidance on burnup credit for spent fuel storage and 

transportation is currently limited to spent fuel assemblies from pressurized water reactors. This 

report describes research to develop a technical basis to expand burnup credit to boiling water 

reactors (BWRs). One of the largest components of uncertainty in burnup credit analyses is the 

predicted isotopic inventories of spent fuel as applied to the criticality safety analysis application 

model. The analysis of BWR fuel inventories is challenging due to the complexity of BWR 

assembly designs, the lack of publicly available radiochemical assay measurements, and limited 

access to documentation on fuel design and operating conditions. This study has compiled and 

evaluated experimental data on measured nuclide concentrations in commercial spent fuel for 

more than 75 fuel samples that cover a wide range of modern assembly designs and operating 

conditions. These data were applied to predict the net effect of isotopic uncertainties on the 

effective neutron multiplication factor for a representative spent nuclear fuel storage system. The 

experimental data, uncertainty analysis methodology, and results for a dry storage cask 

application system are described. The uncertainty analysis methodology presented in this report is

independent of the depletion analysis code and the application model and can be easily adopted 

to estimate margins of uncertainty for other codes, nuclear data libraries, and application models.

The results are only applicable to BWR burnup credit beyond peak reactivity where any initial 

gadolinium present in the fuel has been fully depleted.

v

TABLE OF CONTENTS

ABSTRACT..........................................................................................................................................iii

LIST OF FIGURES..............................................................................................................................vii

LIST OF TABLES................................................................................................................................. ix

EXECUTIVE SUMMARY .....................................................................................................................xi

ABBREVIATIONS AND ACRONYMS..............................................................................................xiii

1 INTRODUCTION........................................................................................................................... 1-1

2 BURNUP CREDIT ANALYSIS METHODOLOGY ...................................................................... 2-1

3 NUCLIDES IMPORTANT TO BURNUP CREDIT........................................................................ 3-1

4 CODE AND MODELLING DESCRIPTIONS................................................................................ 4-1

4.1 Depletion Analyses................................................................................................................. 4-1

4.2 Criticality Analyses ................................................................................................................. 4-2

4.3 Nuclear Data Libraries............................................................................................................ 4-3

5 EXPERIMENTAL ASSAY DATA ................................................................................................. 5-1

5.1 Dodewaard (6 × 6) ................................................................................................................. 5-2

5.2 Forsmark 3 SVEA-100 (10 × 10) ........................................................................................... 5-5

5.3 Forsmark 3 GE14 (10 × 10)................................................................................................... 5-8

5.4 Fukushima Daini 1 (9 × 9-9)................................................................................................. 5-11

5.5 Fukushima Daini 2 (8 × 8-2)................................................................................................. 5-14

5.6 Fukushima Daini 2 GE9 (8 × 8-4) ........................................................................................ 5-17

5.7 Leibstadt SVEA-96 (10 × 10)............................................................................................... 5-19

5.8 Limerick 1 GE11 (9 × 9-7).................................................................................................... 5-22

5.9 Quad Cities GE11 (9 × 9)..................................................................................................... 5-25

6 CRITICALITY APPLICATION MODELS ..................................................................................... 6-1

6.1 GE14 Assembly...................................................................................................................... 6-1

6.2 GBC-68 Cask ......................................................................................................................... 6-2

7 UNCERTAINTY ANALYSIS......................................................................................................... 7-1

7.1 Methodology........................................................................................................................... 7-1

7.1.1 Measurements ............................................................................................................ 7-1

7.1.2 Measurement dates .................................................................................................... 7-2

7.1.3 Partial Isotopic Datasets ............................................................................................. 7-3

7.2 Isotopic Bias and Uncertainty................................................................................................. 7-4

7.3 Criticality Calculations ............................................................................................................ 7-9

7.4 Experimental Correlations.................................................................................................... 7-16

7.5 Isotopic Model Uncertainties................................................................................................ 7-16

7.5.1 Measurements .......................................................................................................... 7-16

7.5.2 Void ........................................................................................................................... 7-16

7.5.3 Fuel Temperature...................................................................................................... 7-17

7.5.4 Sample Burnup ......................................................................................................... 7-17

7.5.5 Summary of Model Uncertainties ............................................................................. 7-18

7.6 Trending Analysis................................................................................................................. 7-18

7.6.1 Trends with Fuel Burnup........................................................................................... 7-19
vi

7.6.2 Trends with Moderator Void...................................................................................... 7-21

7.6.3 Trends with Fuel Reactivity....................................................................................... 7-21

7.7 Margins for Isotopic Uncertainty........................................................................................... 7-24

8 SUMMARY AND CONCLUSIONS............................................................................................... 8-1

9 REFERENCES.............................................................................................................................. 9-1

APPENDIX A MEASURED NUCLIDE CONCENTRATIONS IN SPENT FUEL SAMPLES ........A-1

APPENDIX B COMPARISON OF POLARIS-CALCULATED NUCLIDE COMPOSITIONS 

WITH MEASUREMENTS........................................................................................B-1

APPENDIX C POLARIS MODEL INPUT FILES ............................................................................C-1
vii

LIST OF FIGURES

Figure 1 Overview of the Burnup Credit Validation Process [6]................................................ 2-2

Figure 2 Polaris Lattice Physics Calculation Flow [26] ............................................................ 4-2

Figure 3 Polaris Model of Dodewaard 6 × 6 Assembly............................................................. 5-4

Figure 4 Polaris Model of Forsmark Unit 3 SVEA-100 Assembly............................................. 5-7

Figure 5 Polaris Model of Forsmark Unit 3 GE14 10 × 10 Assembly ..................................... 5-10

Figure 6 Polaris Model of the Fukushima Daini-1 9 × 9-9 Assemblies ................................... 5-14

Figure 7 Polaris Model of the Fukushima Daini-2 8 × 8-2 Assemblies ................................... 5-16

Figure 8 Polaris Model of the Fukushima Daini-2 8 × 8-4 Assemblies ................................... 5-19

Figure 9 Polaris Model of Leibstadt SVEA-96 10 × 10 Assembly........................................... 5-21

Figure 10 Polaris Model of Limerick-1 GE11 9 × 9 Assembly .................................................. 5-24

Figure 11 (a) DOM and (b) VAN Lattice of the GE14 Assembly (Shown in a Dry Cask

Storage Cell)..............................................................................................................6-1

Figure 12 Radial View of the KENO V.a GBC-68 Cask Model (Elevation of Vanished 

Lattice).......................................................................................................................6-3

Figure 13 Uncertainty Analysis Methodology for Nuclide Compositions 

(Adapted from [19])....................................................................................................7-3

Figure 14 Box Plot of the Major Actinide Isotopes..................................................................... 7-6

Figure 15 Box Plot of the Minor Actinides and Fission Products (Mo, Tc, Ru, Ag, and 

Rh).............................................................................................................................7-7

Figure 16 Box Plot of the Fission Products (Nd, Cs, Sm, Eu, and Gd)...................................... 7-8

Figure 17 keff Bias for Actinide-Only and Actinide-Plus-Fission Products Results................... 7-15

Figure 18 keff Bias for Actinide-Only (top) and Actinide-Plus-Fission Products (bottom) 

Results as a Function of fuel Burnup Showing Mean (red) and 95% LTL Band 

(blue) .......................................................................................................................7-20

Figure 19 keff Bias for Actinide-Only (bottom) and Actinide-Plus-Fission Products (top) 

Results as a Function of Moderator Void Showing Mean (red) and 95% LTL 

Band (blue)..............................................................................................................7-22

Figure 20 keff Bias for Actinide-Only (bottom) and Actinide-Plus-Fission Products (top) 

Results as a Function of Application keff Showing Mean (red) and 95% LTL 

Band (blue)...............................................................................................................7-23

ix

LIST OF TABLES

Table 1 Major Actinides Considered in Burnup Credit Criticality Analyses............................... 3-1

Table 2 Actinides and Fission Products Important to Burnup Credit Criticality Analyses ......... 3-1

Table 3 Summary of BWR Spent Fuel Samples ...................................................................... 5-1

Table 4 Summary of Dodewaard 6 × 6 Assembly Fuel Sample Measurements...................... 5-3

Table 5 Dodewaard 6 × 6 Assembly Parameters..................................................................... 5-5

Table 6 Summary of Forsmark Unit 3 SVEA-100 10 × 10 Assembly Fuel Sample 

Measurements.............................................................................................................5-6

Table 7 Forsmark 3 SVEA-100 Assembly Parameters ............................................................ 5-8

Table 8 Summary of Forsmark Unit 3 GE14 10 × 10 Assembly Fuel Sample 

Measurements.............................................................................................................5-9

Table 9 Forsmark 3 GE14 10 × 10 Assembly Parameters [41].............................................. 5-11

Table 10 Summary of Fukushima Daini-1 9 × 9-9 Assembly Fuel Sample Measurements .... 5-12

Table 11 Fukushima Daini-1 9 × 9-9 Assembly Parameters ................................................... 5-13

Table 12 Summary of Fukushima Daini-1 8 × 8-2 Assembly Fuel Sample Measurements .... 5-15

Table 13 Fukushima Daini-2 8 × 8-2 and 8 × 8-4 Assembly Parameters................................ 5-17

Table 14 Summary of Fukushima Daini-2 8 × 8-4 Assembly Fuel Sample Measurements .... 5-18

Table 15 Summary of Leibstadt SVEA-96 10 × 10 Assembly Fuel Sample 

Measurements..........................................................................................................5-20

Table 16 Leibstadt SVEA-96 10 × 10 Assembly Parameters.................................................. 5-22

Table 17 Summary of Limerick GE11 9 × 9 Assembly Measurements ................................... 5-23

Table 18 Limerick-1 GE11 9 × 9 Assembly Parameters.......................................................... 5-25

Table 19 Statistical Analysis of Predicted Isotopic Concentrations (C/M-1) (%) ....................... 7-5

Table 20 Criticality Results for the GBC-68 Cask Model for Major Actinide-Only 

Compositions............................................................................................................7-11

Table 21 Criticality Results for the GBC-68 Cask Criticality Analysis Model for Actinide 

and Fission Product Spent Fuel Compositions......................................................... 7-13

Table 22 Summary of Uncertainties ........................................................................................ 7-18

Table 23 Summary of Bias and Uncertainties for Trending Parameters ................................. 7-24

EXECUTIVE SUMMARY

Applicants for certificates of compliance for spent nuclear fuel transportation and dry storage 

systems perform analyses to demonstrate that these systems are adequately subcritical per the 

requirements of Title 10 of the Code of Federal Regulations (10 CFR) Parts 71 and 72. For spent 

fuel from pressurized water reactors, these analyses may credit the reduction in assembly 

reactivity caused by depletion of fissile nuclides and buildup of neutron-absorbing nuclides during 

power operation. This credit for reactivity reduction during depletion is commonly referred to as 

burnup credit. US Nuclear Regulatory Commission (NRC) staff review burnup credit analyses 

according to the guidance in the Division of Spent Fuel Storage and Transportation Interim Staff 

Guidance (ISG) 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel 

in Transportation and Storage Casks. However, currently no equivalent guidance exists for boiling 

water reactor (BWR) spent fuel.

The technical basis for BWR burnup credit, beyond the burnup region of peak reactivity, is 

presently being evaluated by the NRC. A key element of the burnup credit analysis methodology

is the validation of the depletion codes used to calculate the nuclide compositions of the spent fuel 

that are used in the criticality safety calculations. This step is usually performed by comparing 

calculated nuclide concentrations with measured values obtained from destructive assay of spent 

fuel samples. These comparisons can be used to develop margins for uncertainty in the criticality 

calculation that are associated with the calculated nuclide compositions.

Previous studies to validate nuclide predictions for BWR fuels have been limited by a lack of welldocumented destructive assay data and measurements for modern BWR assembly designs. 

Spent fuel measurements previously considered for BWR isotopic validation included early 6 × 6 

and 7 × 7 BWR assemblies with low enrichments and designs that lacked the heterogeneity of 

modern BWR assemblies. Moreover, the void conditions for these older assemblies were not 

reported. The restricted availability of public sources of BWR data is due in part to the proprietary 

nature of the newer assembly designs, enrichment configurations, and operating conditions in the 

reactor.

The present analysis uses an expanded experimental database of destructive radiochemical 

assay measurements that include detailed design and operating history information. Most of these 

data were made publicly available by the Japan Atomic Energy Agency (JAEA) and the Japan 

Nuclear Regulation Authority (NRA). These data and reference reports have been documented in 

a recently released database of spent fuel compositions developed through international 

cooperation activities of the Organisation for Economic Co-operation and Development (OECD) 

Nuclear Energy Agency (NEA). These data cover several 8 × 8 and 9 × 9 fuel assembly designs. 

In addition to the data contributed by Japan, this work applies measurements for a modern 

General Electric (GE) GE14 10 × 10 fuel assembly made under a proprietary experimental 

program coordinated by the Spanish fuel manufacturer ENUSA Industrias Avanzadas, S.A. and 

the Spanish Nuclear Safety Council, Consejo de Seguridad Nuclear (CSN). Data were obtained 

for a SVEA-96 10 × 10 assembly from the proprietary MALIBU experimental program coordinated 

by the Belgian Nuclear Research Center (SCK•CEN), and additional data for a GE11 9 × 9 

assembly design were obtained from measurements made under the US Department of Energy 

Office of Civilian Radioactive Waste Management (OCRWM) Yucca Mountain project. These data 

provide an improved experimental basis for the evaluation of BWR nuclide compositions

uncertainties by including modern heterogeneous assembly designs, expanded isotopic 

measurements, and more complete reactor operating history information.

xi
These experimental data are applied in this report to develop margins for uncertainty in burnup 

credit criticality calculations associated with the calculation of nuclide compositions for BWR spent 

fuel. A total of 77 spent fuel measurements are used in this analysis that cover a wide range of 

assembly designs including 8 × 8-2, 8 × 8-4, 9 × 9-7 9 × 9-9, GE11 9 × 9, GE14 10 × 10, SVEA96 and SVEA-100 10 × 10 designs. The void fractions extend to 74% with a burnup range from 7 

to 68 GWd/MTU.

The uncertainty analysis methodology applied in the present research is based on the direct 

application of measured nuclide concentrations to a criticality application model for each 

measured spent fuel sample. Separately, nuclide concentrations are calculated for each sample 

and applied to the same application model. Margins for uncertainty are then developed by a 

statistical analysis of the distributions of the difference in measured and calculated keff results for 

all samples. This validation approach is largely independent of the depletion and criticality 

computational methods and safety analysis application model.

This analysis approach is demonstrated in this report using SCALE 6.2.2. with ENDF/B-VII.1 

cross section data. Specifically, depletion calculations were performed using the Polaris code and 

criticality calculations were performed using KENO V.a. Margins for uncertainty associated with 

the predicted nuclide compositions are developed for a GBC-68 dry storage cask model. 

Therefore, the results presented here are specific to this code system and application model but 

are expected to be similar for other comparable dry storage and transportation cask designs, 

when using same computer codes and cross section data.

A margin for uncertainty is developed using the 95% one-sided lower tolerance limit (LTL) for the 

population of data, and trending analysis is performed for sample burnup, average void, and keff of 

the application model. For the Polaris calculations and the GBC-68 application model used in the 

present study, the keff bias for actinide-only calculations is determined to be 253 pcm with a 

maximum margin for isotopic uncertainty of 2,170 pcm for the range of fuel samples included in 

the current analysis. This margin is observed to be largely independent of fuel burnup or void 

fraction based on the analysis of available experimental data. For actinide-plus-fission product 

calculations, the keff bias is 161 pcm and the maximum margin for isotopic uncertainty is 

2,390 pcm.

These results are similar to those seen in previous studies for PWR burnup credit, where average 

biases for a dry storage cask of 320 to 720 pcm and uncertainties of 1,430 to 2,050 pcm were 

reported for fuel with a burnup less than 50 GWd/MTU. The larger uncertainties seen in the 

present study are likely attributed to the increased complexity of the BWR fuel and larger 

uncertainties in the reactor operating data used in the models. 

xii
xiii

ABBREVIATIONS AND ACRONYMS

ARIANE Actinides Research In A Nuclear Element

BWR boiling water reactor

C/M calculated-to-measurement (ratio)

CSN Consejo de Seguridad Nuclear (Spanish Nuclear Safety Council)

DOE US Department of Energy

DOM full lattice 

ENUSA Spanish fuel manufacturer ENUSA Industrias Avanzadas, S.A

ESSM embedded self-shielding method

GBC generic burnup credit (cask design)

GE General Electric

HPLC high performance liquid chromatography

ICPMS inductively coupled plasma mass spectrometry

IDA isotopic dilution analysis

IDMS isotope dilution mass spectrometry

JAEA Japan Atomic Energy Agency

JAERI Japanese Atomic Energy Research Institute

JNES Japan Nuclear Energy Safety Organization

JPDR Japanese Power Demonstration Reactor

LCE laboratory critical experiement

LTL lower tolerance limit

LWR light water reactor

MALIBU MOX and UOX LWR Fuels Irradiated to High Burnup

MG multi-group

MOC method of characteristics

MOX mixed oxide fuel

MS mass spectrometry

MTU metric ton uranium (106 grams)

NEA Nuclear Energy Agency

NRA Nuclear Regulation Authority of Japan

NRC US Nuclear Regulatory Commission

OCRWM Office of Civilian Radioactive Waste Management

OECD Organization for Economic Cooperation and Development

ORNL Oak Ridge National Laboratory

PNNL Pacific Northwest National Laboratory

PSI Paul Scherrer Institute

PWR pressurized water reactor

RCA radiochemical assay

SCALE Nuclear analysis modeling and simulation code suite

SCK•CEN Studiecentrum voor Kernenergie - Centre d'étude de l'Energie Nucléaire

SFCOMPO Spent Fuel Isotopic Composition Database

TEPCO Tokyo Electric Power Company

TIMS thermal ionization mass spectrometry
xiv

VAN vanished lattice

WPNCS Working Party on Nuclear Criticality Safety

YMP Yucca Mountain Project
1-1

1 INTRODUCTION

Interim Staff Guidance 8 [1] on the implementation of burnup credit for storage and transportation 

systems (ISG-8 rev. 3) issued in 2012 by the US Nuclear Regulatory Commission (NRC) applies 

only to pressurized water reactor (PWR) fuel assemblies. There has been less incentive to take 

credit for fuel burnup in criticality safety analyses for boiling water reactor (BWR) assembly 

designs in dry storage and transportation casks because the benefit of burnup credit in increasing 

cask loading capacity is less significant and the assembly designs and reactor operations of 

BWRs are more complex compared to PWR assemblies. However, as the enrichments of BWR 

assemblies have increased from a median of about 2.8 wt% 235U in 1990 to more than 4.0 wt% in 

2013 [2], application of burnup credit in criticality safety analyses can provide increased 

enrichment capacity and/or reduced neutron absorber concentration, and it can also demonstrate 

subcriticality of reconfigured spent fuel in transportation packages.

Consequently, criticality safety bases for BWR spent fuel transportation/storage casks using 

burnup credit are currently being developed by the nuclear industry.

Burnup credit requires, as part of the analysis methodology, that the concentrations of the spent 

fuel nuclides being credited in the criticality safety analysis be calculated using isotopic depletion 

codes. This step of the analysis procedure introduces uncertainties that must be accurately 

quantified using measurement data in order to develop conservative margins in the analysis that 

account for this uncertainty. The uncertainty due to predicted nuclide concentrations represents 

an important component of the total uncertainty in criticality calculations using burnup credit.

The purpose of this report is to present a general methodology to assess the uncertainties 

associated with calculated nuclide compositions and develop margins for the associated 

aggregate uncertainty in the neutron multiplication factor (keff) for criticality calculations. Other 

components of keff uncertainty in BWR burnup credit criticality safety calculations are addressed in 

separate reports [3].

Previous studies to validate nuclide predictions for BWR fuels [4]–[7] have been limited by a lack 

of well-documented measurements of spent fuel compositions and representative assembly

designs and operating information. The restricted availability of public sources of BWR spent fuel 

assay data for validation for modern assembly designs and enrichments is due in part to the 

commercial proprietary nature of the newer assembly designs, enrichment configurations, and 

operating conditions in the reactor. Spent fuel measurements previously considered for BWR 

isotopic validation included early 6 × 6 and 7 × 7 BWR assemblies with low enrichments and 

designs that lacked the heterogeneity of modern BWR assemblies. Moreover, the axial void 

conditions for these older assemblies were not reported. Measurements of an 8 × 8 BWR 

assembly from the Fukushima Daini-2 reactor were reported by the Japan Atomic Energy Agency 

(JAEA) with void information included [8]; these data were also used in the earlier studies.

The present analysis uses an expanded experimental database obtained using destructive 

radiochemical assay measurements and more detailed operating history information. Extensive 

measurements from experiments performed in Japan were recently made publicly available by the 

Japan Nuclear Regulation Authority (NRA) through international cooperation activities of the 

Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency 

(NEA) Working Party on Nuclear Criticality Safety (WPNCS) [9]. These data and experimental 

references are documented in an OECD/NEA database of spent fuel compositions [10].
1-2

In addition to the public data contributed by Japan, this work applies measurements for a modern 

General Electric (GE) GE14 10 × 10 fuel assembly made under a proprietary experimental 

program coordinated by the Spanish fuel manufacturer ENUSA Industrias Avanzadas, S.A. and 

the Spanish Nuclear Safety Council, Consejo de Seguridad Nuclear (CSN) [11]. Data were 

obtained for a SVEA-96 10 × 10 assembly from the proprietary MALIBU experimental program 

coordinated by the Belgian Nuclear Research Center (SCK•CEN) [12]. Additional data for a GE11 

9 × 9 assembly design were obtained from measurements made under the US Department of 

Energy Office of Civilian Radioactive Waste Management (OCRWM) Yucca Mountain project [13]. 

These data provide an improved experimental basis for the evaluation of BWR isotopic 

uncertainties by including modern heterogeneous assembly designs, expanded isotopic 

measurements, and more complete reactor operating history information.

The validation approach in this report is independent of the depletion and criticality computational 

methods and safety analysis application model. The range of application includes fuel burnup 

beyond peak reactivity that is associated with the use of fuel containing gadolinium oxide (Gd2O3)

or other integral neutron absorbers. Analyses are included that take credit in the criticality safety 

calculations for the major actinides in spent fuel with and without the addition of minor actinides 

and principal fission products. The uncertainty in keff due to biases and uncertainties in calculated 

nuclide concentrations is presented based on analyses performed with the Polaris lattice physics 

code [14] in SCALE 6.2.2 [15] and Evaluated Nuclear Data File/B Version VII.1 (ENDF/B-VII.1) 

nuclear cross section and decay data [16]. Criticality calculations were performed using the KENO 

V.a Monte Carlo neutron transport code and the 252-energy group ENDF/B-VII.1 cross section

library available in SCALE 6.2.2. 

This report presents an overview of the burnup credit analysis methodology in Section 2 that

identifies the main components of the depletion and criticality validation methodology. The 

isotopes that are considered in burnup credit are discussed in Section 3. Section 4 describes the 

computational methods and nuclear data used in the depletion calculations applied in this report. 

The experimental isotopic assay data used to validate the computational codes are presented in 

Section 5. Representative application models for BWR spent fuel assemblies and a dry storage 

cask are described in Section 6. Section 7 presents an analysis of keff bias and uncertainty results 

based on SCALE 6.2.2 for the application model. Conclusions are provided in Section 8.

Appendices A, B and C provide measured nuclide concentrations used in this report, comparisons 

of measurements to Polaris calculated concentrations, and sample Polaris input files for different 

lattice types.
2-1

2 BURNUP CREDIT ANALYSIS METHODOLOGY

Criticality safety analyses are performed to demonstrate that a proposed fuel storage or transport 

configuration meets the applicable requirements of Title 10, Code of Federal Regulations (CFR), 

Parts 71 and 72 [17]. A general overview of the process for implementing burnup credit for 

criticality safety analyses is outlined in Figure 1. The process includes development of safety 

analysis models considering a range of parameters important to criticality safety and isotopic and 

criticality validation analyses to demonstrate that the proposed configuration will meet the 

maximum keff limits specified in the applicable requirements and guidance.

This report addresses only the uncertainty component associated with the predicted isotopic 

inventory used in burnup credit calculations beyond peak reactivity.

The methods used to calculate the burnup credit nuclide compositions are usually validated 

through comparisons to measured radiochemical assay (RCA) data. The validation steps 

addressed in this report cover items (5) through (7) in Figure 1. The criticality validation addresses 

items (8) through (15) in Figure 1. In this report, the consensus terminology from ANSI/ANS-8.27 

[18] is used.

The criterion for establishing subcriticality with credit for the reactivity decrease due to fuel burnup 

requires the calculated multiplication factor kp plus allowances for biases and uncertainties shall 

be equal to or less than an established, allowable neutron multiplication factor; that is,

kp + ∆kp + βi + ∆ki + β + ∆kβ + Δkx + ∆km ≤ klimit, (1)

where

kp is the calculated multiplication factor of the model for the system being evaluated;

∆kp is an allowance for

• statistical or convergence uncertainties, or both, in the determination of kp,

• material and fabrication tolerances, and

• uncertainties due to geometric or material representation limitations of the models 

used in the determination of kp;

βi is the bias in kp due to depletion code bias in the calculated nuclide concentrations;

∆ki is the bias uncertainty in kp due to depletion code bias uncertainty in the calculated nuclide 

concentrations; 

β is the bias that results from using a particular calculation method and nuclear cross section 

data to calculate keff values for benchmark criticality experiments;
2-2

RCA: Radiochemical Assay

LCE: Laboratory Critical Experiment

Figure 1 Overview of the Burnup Credit Validation Process [6]
2-3

∆kβ is the criticality bias uncertainty, which includes 

• statistical or convergence uncertainties, or both, in the computation of β,

• uncertainties in the benchmark criticality experiments,

• uncertainty in the bias resulting from application of the linear least-squares fitting

technique to the critical experiment results, and

• tolerance interval multiplier to yield a single-sided 95% probability and 95% confidence

level;

Δkx is a supplement to β and ∆kβ that may be included to provide an allowance for the bias 

and uncertainty from nuclide cross section data that might not be adequately accounted 

for in the benchmark criticality experiments used for calculating β; 

∆km is a margin for unknown uncertainties deemed to be adequate to ensure subcriticality of 

the physical system being modeled (this term is typically referred to as an administrative 

margin); and

klimit is the upper limit on the keff value for which the system is considered safe. 

The validation approach using measured nuclide concentrations requires that fuel samples are 

available for representative assembly designs and operating conditions and that measured 

nuclide concentrations are available for the actinide and/or fission products being credited in the 

criticality safety application model.

Several approaches for propagating isotopic composition uncertainty to keff uncertainty have been 

developed and demonstrated [20], each involving different levels of analysis complexity and 

having different degrees of conservatism. The approach used in the present study analysis has 

three steps. In the first step, the calculated isotope concentrations for each measured fuel sample 

are used in the criticality application model to determine the keff. This step is performed without 

any adjustment of the model parameters for conservatism and with no accounting for bias or 

uncertainty in the calculated concentrations. In the second step, the measured isotopic 

concentrations for each sample are applied in the criticality model to obtain the keff. The difference 

between the keff values obtained using calculated nuclide compositions in the model, kc, and 

measured nuclide composition in the model, km, is a direct measure of the keff bias, βi, due to the 

depletion calculation code for that sample.

By analyzing a significant number of spent fuel samples with variable enrichments, burnup, and 

void conditions, the behavior of the keff bias can be statistically estimated for the range of fuel 

parameters and the keff bias, βi, and uncertainty in the bias, ∆ki, can be determined for the 

application system. This last step is referred to as the propagation of isotopic composition 

uncertainties to keff and is identified as step (6) in Figure 1. 

The significant benefit of this approach is that the bias and uncertainty in each individual nuclide 

do not need to be determined. Since nuclide concentrations as calculated from depletion codes

and measured data are applied directly in criticality calculations, correlations in the nuclide bias 

associated with both the measurements and the calculations for each sample are implicitly 

addressed by this approach. This method also avoids the complexity of Monte Carlo sampling of 

individual nuclide uncertainties that has been applied previously [6]. A limitation of the present 
2-4

approach is that criticality calculations are restricted to the fuel compositions for each sample that 

have been measured. Therefore, this approach does not address cases of variability of fuel 

compositions in the models, such as axial burnup distributions variations due to zoned assembly 

enrichments.

This methodology is similar to an approach applied previously under the Department of Energy 

OCRWM in support of post-closure criticality safety assessments for the Yucca Mountain Project

[19]. However, these previous studies applied bounding input parameters in the depletion analysis

in order to provide conservative estimates of the nuclide compositions. The current work performs 

the depletion calculations for each sample using the best available design and operating 

parameters without conservatism. In addition, the previous work did not account for burnup credit 

nuclides that were not measured in some spent fuel samples, resulting in a different set of 

nuclides being credited in different samples. The current work applies estimates of the 

concentrations for isotopes that were not measured based on results for samples that did 

measure those isotopes. Additional uncertainty introduced by this procedure is estimated.

A limitation of the approach of using isotopic destructive RCA measurements to validate the 

depletion calculations is that the calculations require, as input, data that can have large 

uncertainties. In particular, unlike PWR assemblies where the properties of the water moderator 

are relatively well known, the moderator void fractions for BWR assemblies are not well known. 

Void fraction is not measured directly but is obtained from core follow codes that predict the 

neutronic and thermal-hydraulic behavior of the reactor. These codes provide data on a spatial 

average (nodal) level sufficient to model core behavior and fuel management. However, when 

applied to the analysis of individual fuel rods, larger uncertainties may be introduced. Average 

nodal data do not account for spatial variations of void fraction within an assembly and average 

void fractions may not be representative of the moderator environment in the proximity of the 

measured fuel samples. Consequently, the modeling uncertainties for BWR fuel are expected to 

be larger than for PWR fuel.

Other uncertainties in the methodology include the burnup of the measured fuel sample and the 

accuracy of the assay measurements themselves. The sample burnup is estimated using 

measured burnup indicator nuclides that can have large measurement uncertainties. Additionally, 

there can be inconsistencies between different burnup indicators. The impact of these modeling 

uncertainties is discussed further in the report.

In practice, criticality calculations are usually performed using spent fuel nuclide compositions 

calculated using conservative depletion (operational) parameters to ensure conservative 

estimates of the nuclide inventory. The purpose of this report is to develop additional margins for 

the uncertainty associated with the code calculations themselves by using experiments with well

documented operating parameters.
3-1

3 NUCLIDES IMPORTANT TO BURNUP CREDIT

The burnup credit analysis methodology presented in this report considers spent fuel 

compositions consisting of 12 actinide and 16 fission product nuclides selected on the basis of 

their importance to fuel reactivity (i.e., nuclides with the largest macroscopic neutron fission cross 

sections and neutron absorption cross sections) and on the basis of the chemical volatility of the 

isotope that may cause it to be released from the fuel. These nuclides are frequently categorized

as the nine major actinides (Table 1) and 28 major actinides, minor actinides, and major fission 

product nuclides (Table 2). The burnup credit nuclides have been identified in previous studies as 

having the most significant effects on keff for burnup credit analyses related to dry cask storage 

and transportation of PWR and BWR fuel assemblies [21].

These 28 nuclides are commonly considered in burnup credit criticality safety analyses that base 

validation of calculated nuclide concentrations on comparisons to available assay data. Of the 16 

fission products listed in Table 2, 149Sm, 151Sm, 143Nd, 103Rh, 133Cs, and 155Gd have the largest 

reactivity impact. Credit for fission product nuclides has been limited in the past by the relatively 

small number of measurements available for BWR fuel. The burnup credit nuclides are stable or 

very long lived with the exception of 151Sm (T1/2 = 90 years), 238Pu (T1/2 = 87.7 years), 241Pu (T1/2 = 

14.4 years), and 241Am (T1/2 = 432.7 years).

The reactivity of commercial spent fuel increases immediately after discharge due to the reduction 

in 135Xe and other short-lived fission products. Starting at approximately 3 days after discharge,

the reactivity decreases largely due to the decay of the fissile nuclide 241Pu leading to 241Am 

(neutron absorber) and to the formation of 155Gd (a strong large neutron absorber) from the decay 

of 155Eu (T1/2 = 4.75 years). After approximately 100 years, the reactivity begins to increase due 

primarily to the decay of 241Am and 240Pu (T1/2 = 6,560 years).

Table 1 Major Actinides Considered in Burnup Credit Criticality Analyses

234U 235U 238U 238Pu 239Pu 240Pu 241Pu 242Pu 241Am

Table 2 Actinides and Fission Products Important to Burnup Credit Criticality Analyses

234U 235U 236

U 238U 237Np 238Pu 239Pu

240Pu 241Pu 242Pu 241Am 243Am 95Mo 99Tc

101Ru 103Rh 109Ag 133Cs 143Nd 145Nd 147Sm

149Sm 150Sm 151Sm 152Sm 151Eu 153Eu 155Gd

4-1

4 CODE AND MODELLING DESCRIPTIONS

A variety of computer codes and nuclear data libraries can be used to perform burnup credit 

criticality safety analyses. Fuel depletion calculations are typically performed by industry using 

two-dimensional (2D) lattice physics models of an assembly to calculate fuel compositions. Widely 

used codes in the United States include SCALE [15], CASMO [22], and PARAGON [23]. Spent

fuel compositions obtained with these codes are then used in criticality codes such as KENO in 

SCALE or MCNP [24] to determine the keff for the application system.

All depletion and criticality calculations in the present study were performed using codes and 

nuclear data libraries in SCALE 6.2.2, which is available publicly through the Radiation Safety 

Information Computational Center (RSICC) as package CCC-834. SCALE 6.2.2 is the first version 

that contains BWR modeling capabilities in the Polaris code for a large variety of assembly 

designs. 

4.1 Depletion Analyses

Polaris is a new module introduced in SCALE 6.2 that provides 2D lattice physics analysis with 

pin-by-pin depletion capability for production calculations of light water reactor (LWR) fuel 

assembly designs. A detailed description of the methods and calculational approach of Polaris is 

provided by Jesse et al. [14]. Polaris provides a multigroup (MG) neutron transport capability 

coupled with the ORIGEN module to solve the time-dependent transmutation equations and 

isotopic evolution of materials during irradiation and decay.

Polaris was developed as a more efficient transport and depletion code for LWR analyses, 

compared to the general-purpose TRITON depletion capability in SCALE that uses onedimensional (XSDRN), 2D (NEWT), or three-dimensional (KENO) neutron transport solutions. For 

the neutron transport calculation, Polaris employs the method of characteristics (MOC) which 

solves the characteristic transport equation over a set of equally spaced particle tracks across the 

lattice geometry. Polaris also provides an easy-to-use input format allowing users to set up lattice 

models with a minimal amount of input compared to TRITON requirements.

An efficient embedded self-shielding method (ESSM) is used in Polaris for resonance selfshielding of all fuel rods in an assembly [25]. ESSM is similar to the subgroup method where selfshielding effects due to neighboring fuel pins, guide tubes, water rods, and assembly structures 

are accounted for in the calculation. ESSM neglects resonance interference between resonanceabsorbing nuclides in the same material. Cross section self-shielding is performed automatically to 

account for changes in the moderator void fraction and other operating conditions during the 

depletion analysis. In previous depletion studies for BWR fuel performed using TRITON, Dancoff 

factors used for resonance cross section corrections had to be calculated externally, usually with 

the MCDANCOFF code in SCALE, or an equivalent code, and applied manually as input to the 

model. When the Dancoff factors changed during irradiation due to variations in the moderator 

void and burnup, updating the factors required halting the calculation, saving the intermediate 

nuclide concentrations, inputting new Dancoff factors, and restarting the case. This procedure is 

performed internally in Polaris.

Within Polaris, the ORIGEN code is used to calculate time-dependent concentrations, activities, 

and radiation source terms for a large number of isotopes simultaneously generated or depleted 

by neutron transmutation, fission, and radioactive decay. 
4-2

Polaris has been validated for reactor physics lattice calculations [26]. Comparisons of Polaris and 

TRITON results generally show consistent accuracies. The present study represents the first 

application of Polaris for extensive BWR isotopic validation.

The Polaris calculation flow is shown in Figure 2.

Figure 2 Polaris Lattice Physics Calculation Flow [26]

4.2 Criticality Analyses

Criticality calculations were performed using the SCALE KENO V.a three-dimensional (3D) Monte 

Carlo neutron transport code with MG cross sections. The KENO V.a calculations are accessed

through the criticality safety analysis sequence (CSAS). This criticality sequence was used to 

perform automated problem-dependent cross section processing followed by KENO V.a

calculations to solve the keff eigenvalue problem.

Criticality calculations were performed using the two sets of nuclides considered for fuel modeling 

in burnup credit: (1) major actinides only and (2) major and minor actinides and major fission 

products, as listed in Tables 1 and 2. Note that although not explicitly listed as a burnup credit 

isotope, oxygen is included in the fuel compositions, as it is a major constituent of the fuel matrix. 

XS Processing

(ESSM)

MG Transport 

(MOC) 

Depletion

(ORIGEN)

Time step loop

• Input (.inp)

• Nuclear data 

libraries

PolarisBranch loop
4-3

4.3 Nuclear Data Libraries

Neutron transport calculations in Polaris are performed using the 56-group ENDF/B-VII.1 cross 

section library for all analyses in this report.

Following each transport calculation performed by Polaris, these cross sections are collapsed and 

applied directly to the ORIGEN calculation to determine reaction rates and the nuclide 

transmutation inventories. ENDF/B-VII.1 [16] provides the cross sections for 388 individual 

isotopes. Cross sections for 386 isotopes not available in ENDF/B-VII.1 are taken from a MG 

library based on the JEFF-3.1/A, a special-purpose activation library [27], and are collapsed using 

the same procedures. Due to their negligible self-shielding and impact on transport calculations, 

nuclides obtained from the JEFF-3.1/A cross sections are not processed through the ESSM 

module and are applied as unshielded (infinitely dilute) cross sections.

All decay data used by ORIGEN are adopted from ENDF/B-VII.1. Independent fission product 

yields are developed from England and Rider [28] as included in ENDF/B-VII.0. The independent 

yields used by ORIGEN [29] have been adjusted to account for changes in the decay data and

provide greater consistency in the cumulative fission yields in the England and Rider evaluation.

KENO V.a criticality calculations of the application model were performed using the 252-group 

ENDF/B-VII.1 neutron transport cross section library in SCALE. 

5-1

5 EXPERIMENTAL ASSAY DATA

Measured BWR nuclide compositions are obtained from destructive RCA experiments of spent 

fuel rods selected from assemblies irradiated in eight different reactors operated in five countries. 

These assemblies include 6 × 6, 8 × 8, 9 × 9, and 10 × 10 lattice designs. A total of 77 measured 

samples were evaluated. More than 80% of the samples are from Fukushima Daini Units 1 and 2 

in Japan. The measured data used in the present study are summarized in Table 3.

Table 3 Summary of BWR Spent Fuel Samples

Reactor and Unit Country Assembly 

Design

Number of 

Samples

Enrichments

(wt % 235U)

Burnup

(GWd/MTU)

Dodewaard Belgium 6 × 6 1 4.94 55

Forsmark 3 Sweden 10 × 10 (SVEA-96) 1 3.97 61

Forsmark 3a Sweden 10 × 10 (GE14) 8 3.95 38–50

Fukushima Daini 1 Japan 9 × 9-9 13 2.1, 4.9, 3.0 (Gd) 35–68

Fukushima Daini 2 Japan 8 × 8-4 25 3.4, 4.5, 3.4 (Gd) 9–59

Fukushima Daini 2 Japan 8 × 8-2 18 3.9, 3.4 (Gd) 7–44

Leibstadt 3b Switzerland 10 × 10 (SVEA-96) 3 3.9 56–63

Limerick 1c United 

States

9 × 9 (GE11) 8 3.95, 3.6 (Gd) 37–65

a Spanish Nuclear Safety Council (CSN) (Proprietary data)

b MALIBU International Program (Proprietary data) c US DOE Yucca Mountain Project (Proprietary data)

Measurements of the major actinide isotopes were available for all samples. Minor actinide and 

fission product measurements are available for many of the samples.

Several experimental datasets analyzed in previous studies [4]–[7] were not used in the current 

study due to insufficient documentation on the reactor operating conditions, most notably the local 

void fractions for the samples. The datasets that were not considered here included

measurements from the Cooper reactor [30], Gundremmingen reactor [31], and the Japanese 

Power Demonstration Reaction (JPDR) [32] and [33]. Previous studies used semiempirical 

correlations of assembly power and axial void distributions from other assemblies to estimate the 

local void for measured assemblies. In the present study, only experimental datasets with 

reported axial void fractions were considered.

Many public sources of isotopic assay data are currently compiled as part of the OECD/NEA 

Spent Nuclear Fuel Isotopic database SFCOMPO 2.0 [10], maintained and distributed by the NEA 

Data Bank. The experimental data in SFCOMPO has been compiled from international 

contributions through the Expert Group on Assay Data for Spent Nuclear Fuel, under the auspices 

of the OECD/NEA WPNCS. All primary experimental reports on each dataset are maintained and 

made available as part of the database.

SFCOMPO 2.0 includes isotopic measurements for spent fuel samples from the following BWR 

reactors and assembly designs:

• Cooper (US) 7 × 7

• Dodewaard (Belgium) 6 × 6
5-2

• Forsmark 3 (Sweden) 10 × 10

• Fukushima Daiichi 3 (Japan) 8 × 8

• Fukushima Daini 1 (Japan) 9 × 9

• Fukushima Daini 2 (Japan) 8 × 8

• Garigliano (Italy) 8 × 8 and 9 × 9

• Gundremmingen (Germany) 6 × 6

• Japanese Power Demonstration Reaction (Japan) 6 × 6

• Monticello (US) 8 × 8

• Quad Cities (US) 8 × 8

• Tsuruga (Japan) 7 × 7

Only the measurements from Dodewaard, Forsmark 3, Fukushima Daini 1, and Fukushima Daini

2 reactors were used in this study since they include relatively complete design and operating 

history data.

Additional data used in this report were obtained from commercial proprietary programs that 

measured fuel samples from the Forsmark 3, Leibstadt 3, and Limerick 1 reactors, but information 

on these datasets is restricted. Descriptive data included in this report are limited to information 

available from public sources. Additional proprietary information required for modeling and 

simulation of these fuel assemblies is only available through a nondisclosure agreement.

Each experimental dataset used in the present study is described in the following sections. This 

information includes a general description of the measurements and the assembly design data. 

Because of the complexity of BWR design and operating history data, the information presented 

here is not intended to provide a complete description necessary for modeling purposes; rather,

only a summary of the general assembly characteristics is given. Detailed information is available 

in the primary experimental reports cited in this report.

The burnup of the measured samples is a critical parameter in the modeling and simulation of the 

nuclide compositions. Burnup is not measured directly but is usually estimated from the measured 

concentration of one or more fission product burnup indicators. In this work 148Nd was used as the 

primary indicator of the sample burnup. Depletion calculations were performed by adjusting the 

specific power of the sample to obtain agreement with the measured 148Nd inventory within 1% or 

better. Cases where 148Nd was not used for burnup determination are clearly indicated in the 

report.

5.1 Dodewaard (6 × 6)

Dodewaard was a BWR nuclear power plant that operated in the Netherlands until 1997. 

Destructive RCA measurements of fuel samples were performed as part of the ARIANE (Actinide 

Research in a Nuclear Element) international project [34]. ARIANE was an experimental program 

coordinated by Belgonucleaire with participants from institutions and companies in Belgium, 

Germany, Japan, the Netherlands, Switzerland, United Kingdom, and the United States. The 

objective of the ARIANE program was to obtain an accurate radiochemical post-irradiation 

examination of mixed oxide (MOX) and UO2 fuel samples in PWR and BWR operating conditions.

Experimental data from ARIANE were released publicly to the OECD/NEA Expert Group on 

Assay Data for Spent Nuclear Fuel, under the WPNCS. Measurement data and experimental 

reports are available through the SFCOMPO database.
5-3

The Dodewaard UO2 sample, designated DU1, had an initial 235U enrichment of 4.941% and was 

irradiated for five cycles to about 55 GWd/MTU in fuel assembly Y013. This assembly was

discharged in January 7, 1993. The assembly was manufactured by the Japan Nuclear Fuels Co., 

Ltd. (JNF), as a lead test assembly of an early BWR 6 × 6 lattice design containing one water rod 

and five gadolinium oxide (Gd2O3) rods. The DU1 sample was cut from a segmented rod located 

in position B2 of the assembly lattice. Sample DU1 was located in the third fuel segment from the 

bottom (four total segments) of segmented rod B2 at a height of 1,111 mm from the bottom end of 

the active fuel. The reported average void fraction was 50%. The basic fuel sample characteristics 

are listed in Table 4.

Table 4 Summary of Dodewaard 6 × 6 Assembly Fuel Sample Measurements

Assembly

ID

Rod

ID

Sample

ID Fuel type

Axial 

height 

(mm)

Avg. Void 

(%)

Enrichment

(wt % 235U)

Gd content

(wt % Gd2O3)

Burnup

(GWd/MTU)

Y013 B2 DU1 UO2 1,111 50 4.941 0 55.5

The other fuel rods in assembly Y013 were standard UO2 rods (full length) with variable 

enrichments (3.2, 2.6, and 1.8 wt %) except for two other experimental rods located in positions 

D5 and E4 that contained MOX with 6.43 wt% plutonium content. The MOX rods were positioned 

well away from the measured DU1 sample. Two gadolinium rods with 2.7 wt% Gd2O3 content in 

fuel and 3.2% enriched in 235U were adjacent to the measured rod. The Y013 assembly layout is 

shown in the Polaris model in Figure 3.

Assembly Y013 was irradiated for five cycles and achieved an operation-estimated assembly 

average burnup of 35.24 GWd/MTU. The large difference between the assembly and sample 

burnup is due to the high initial 235U content of the sample rod, while the other UO2 fuel rods in the 

assembly had a maximum enrichment of 3.2%.

Assembly Y013 is not highly representative of modern designs, and it contained a segmented test 

rod from which sample DU1 was obtained. However, detailed design and operating history 

information was available from the operator at the sample axial location and extensive nuclide 

measurements were reported. Applicability of the DU1 sample for validation has been 

independently evaluated [35], and it was concluded that these data are suitable for validating 

isotopic depletion codes.

Independent measurements of the DU1 sample were performed at laboratories of the Belgian 

Nuclear Research Center, Studiecentrum voor Kernenergie (SCK•CEN) in 1996, and also at the 

Paul Scherrer Institute (PSI) in Switzerland in 1999 [34]. These measurements were conducted as 

part of the same experimental program to provide a cross check of the nuclide concentrations and 

verification of the estimated measurement uncertainties. Extensive isotopic data obtained using 

destructive RCA measurements are reported for 53 nuclides. The measurements at SCK•CEN 

were made using high precision Thermal Ionization Mass Spectrometry (TIMS) coupled to the 

Isotopic Dilution Analysis (IDA) technique. The measurements by PSI were performed using a 

multi-collector Inductively Coupled Plasma Mass Spectrometer (ICPMS) also with ID analysis. 

Measurement data are available for all 28 burnup credit isotopes listed in Table 2. In this report, 

analyses were performed using both sets of measurements to provide an estimate of the impact 

of isotope measurement uncertainties.
5-4

Figure 3 Polaris Model of Dodewaard 6 × 6 Assembly

Detailed core follow data for the measured sample are included in the ARIANE report [34]. Timedependent void fraction, burnup, center and surface fuel temperatures are provided for all five 

cycles. These time-dependent operating data were applied in the Polaris model. An effective fuel 

temperature was calculated from the fuel center and surface temperatures using Rowlands’s 

formulation [36]. The main fuel and assembly characteristics are summarized in Table 5. 

The sample burnup was determined by matching the 148Nd concentration predicted by Polaris with 

the measurement data, estimated by the laboratory to have an accuracy of better than 1% 

(95% confidence).
5-5

Table 5 Dodewaard 6 × 6 Assembly Parameters

Reactor and Assembly Data

Reactor Dodewaard

Operating pressure (bar) 75.5

Assembly type 6 × 6 GE/RDM

Fuel assembly pitch (cm) 12.7

Number of fuel rods 35

Number of gadolinia rods 5

Number of water rods 1

Fuel rod pitch (cm) 1.793

Channel bypass gap (cm) 0.872 / 0.476

Channel thickness (cm) 0.17

Channel corner radius a (cm) 0.9652

Channel material Zirc-2

Channel temperature b (K) 564

Active fuel length (cm) 179.3

Fuel Rod Data

Pellet radius c (cm) 0.5740 / 0.5176

Pellet material UO2

Pellet density d (% theoretical density) 95.46

Fuel temperature (K) 1052

Clad inner radius (cm) c 0.5855 / 0.5275

Clad outer radius (cm) c 0.6745 / 0.6135

Clad material Zirc-2

Clad temperature b (K) 564

Water Rod Data

Inner radius e (cm) 0.5855

Outer radius e (cm) 0.6745

Water rod material Zirc-2

Water rod temperature b (K) 564 a Assumed to be the same as the 8 × 8 assembly [37] b Assumed to be at core average coolant temperature c Standard and segmented rod dimensions d Segmented UO2 rod e Assumed to be identical to fuel rod cladding dimensions

5.2 Forsmark 3 SVEA-100 (10 × 10)

Measurements of fuel samples from SVEA-100 10 × 10 fuel assembly 14595, irradiated in the 

Forsmark Unit 3 reactor located in Sweden, were performed at the Studsvik Nuclear Laboratory. A

10 mm sample (identified as sample FFBU) from the central part of the UO2 rod located at position 

F6 of assembly 14595 was dissolved at Studsvik. Aliquots of the fuel solution were shipped to two

other laboratories in 1996, Harwell in the UK and Dimitrovgrad in Russia, for independent 

radiochemical determination of the isotopic composition and burnup analysis. Another sample 

was analyzed by Studsvik. These measurements and the experimental report were published in 

2008 by Zwicky [38] and are available through the SFCOMPO database.

The Studsvik sample used in this study, identified as F3F6, was adjacent to FFBU and was 2 mm 

in length. It was dissolved and analyzed at Studsvik in 2003, and an aliquot of the same fuel 

solution was reanalyzed with new equipment in 2006. Results of these measurements and 

detailed assembly design and operating history information are published. The F3F6
5-6

measurements were used in this study. The Harwell and Dimitrovgrad measurements (FFBU) 

were used by Studsvik as cross checks of measurement accuracy.

Assembly 14595 was irradiated for six cycles and achieved a reactor-operator estimated burnup 

of 58 MWd/kgU. The assembly was discharged on June 6, 1993. The measured rod F6 had an 

enriched zone height of 3450 mm and natural uranium blanket zones that were 150 mm high at 

the top and bottom of the rod, resulting in a total active fuel height of 3750 mm. Sample F3F6 was 

obtained at an axial position 2004 mm from the bottom of the fuel rod and experienced an 

average void fraction of 58%. The fuel sample characteristics are listed in Table 6. The F3F6 

sample burnup was estimated by Studsvik based on the measurements performed in 2006 using 

weighted burnup values based on measurements of neodymium isotopes, 235U, and 239Pu.

Table 6 Summary of Forsmark Unit 3 SVEA-100 10 × 10 Assembly Fuel Sample

 Measurements

Assembly

ID

Rod

ID

Sample 

ID

Fuel 

type

Axial 

height 

(mm)

Avg. 

Void 

(%)

Enrichment

(wt % 235U)

Gd content

(wt % Gd2O3)

Burnup

(GWd/MTU)

14595 F6 F3F6 UO2 2004 58 3.97 0 55.8

The layout of the Forsmark-3 assembly 14595 is shown in Figure 4 with the location of the 

measured rod F6 at the inner corner of the assembly subchannel and the subchannel structure 

(water cross) shown. The assembly used 10 different enrichments in the fuel rods, and five rods 

had a Gd2O3 content of 3.15 wt%. The general design specifications for the SVEA-100 assembly 

14595 are listed in Table 7. Detailed time-dependent void fractions, fuel temperature, and specific 

power for the measured sample are provided in the reference report [38].
5-7

Figure 4 Polaris Model of Forsmark Unit 3 SVEA-100 Assembly

Measurements used in this study were part of the December 2006 analysis campaign of the F3F6 

sample at Studsvik and were made using a High Performance Liquid Chromatography (HPLC) 

system for the chemical separations. Mass spectrometry was performed using an ICPMS. 

Measurements are reported for isotopes of uranium, plutonium, cerium, and neodymium. IDA was 

used with spike solutions of 233U, 242Pu, 140Ce, and 148Nd.
5-8

Table 7 Forsmark 3 SVEA-100 Assembly Parameters

Reactor and Assembly Data

Reactor Forsmark Unit 3

Operating pressure (bar) 70.0

Assembly type 10 × 10 SVEA-100

Fuel assembly pitch (cm) 15.475

Number of fuel rods 100

Number of gadolinia rods 5

Number of water rods 0

Fuel rod pitch (cm) 1.24

Channel bypass gap (cm) 0.6775 / 0.7775

Channel thickness (cm) 0.14

Channel corner radius (cm) 1.01

Channel material Zirc-2

Channel temperature a (K) 560

Active fuel length (cm) 375

Fuel Rod Data

Pellet radius (cm) 0.481

Pellet material UO2

Pellet density (% theoretical density) 96.9

Clad inner radius (cm) 0.5275 

Fuel temperature (K) 792

Clad outer radius (cm) 0.6135 

Clad material Zirc-2

Clad temperature a (K) 560

Water Rod Data

Inner radius (cm) N/A

Outer radius (cm) N/A

Water rod material N/A

Water rod temperature (K) N/A a Assumed to be at core average coolant temperature

5.3 Forsmark 3 GE14 (10 × 10)

Under a proprietary Spanish experimental program [11] coordinated by the Spanish fuel vendor 

ENUSA, isotopic measurements were made on a modern GE14 10 × 10 assembly from the 

Forsmark Unit 3 reactor operated in Sweden. The program involved the Spanish safety council for 

nuclear activities, CSN, and the organization responsible for waste management in Spain, 

ENRESA, with support provided by the US NRC and Oak Ridge National Laboratory (ORNL). The 

program provided extensive isotopic composition data for benchmarking codes and computational 

models used in reactor safety studies as well as for interim storage, transportation, and final 

disposal of spent nuclear fuel.

Fuel samples of rod J8 from GE14 assembly GN592 were measured at the Studsvik Nuclear 

Laboratory. A total of eight fuel samples from fuel rod J8 were measured over the entire rod length

to provide data on burnup and void variations. Samples 1 and 2, and samples 3 and 7, were 

selected from adjacent axial positions of the rod to verify measurement repeatability and 

uncertainty. The measurements provided isotopic data at six unique axial positions.

Measured fuel rod J8 was located at the periphery of the assembly, on a side away from the 

control blades. All samples from rod J8 were from the enriched zone of the rod with an initial 
5-9

enrichment of 3.95 wt % 235U. The fuel assembly was irradiated for five consecutive operating 

cycles (cycles 16–20) and was discharged on May 28, 2005. The measured fuel rod attained an 

estimated rod average burnup of 41 GWd/MTU and peak burnup of about 56 GWd/MTU.

Table 8 provides a summary of sample identification names, the elevation of each sample, and 

the estimated void at the sample locations. Sample elevations are measured from the lower end 

plug of the fuel rod. The distance from the lower end plug to the start of the active fuel region is 

about 40 mm.

Table 8 Summary of Forsmark Unit 3 GE14 10 × 10 Assembly Fuel Sample

 Measurements

Assembly

ID

Rod

ID

Sample

ID

Fuel 

type

Axial

height

(mm)

Avg. Void

(%)

Enrichment

(wt % 235U)

Gd content

(wt % Gd2O3)

Burnup

(GWd/MTU)

GN592 J8

ENUSA-1 UO2 1,847 51 3.95 0 50.4

ENUSA-2 UO2 1,858 51 3.95 0 50.7

ENUSA-3 UO2 718 13 3.95 0 49.0

ENUSA-4 UO2 2,508 61 3.95 0 51.1

ENUSA-5 UO2 3,282 67 3.95 0 43.6

ENUSA-6 UO2 403 2.2 3.95 0 43.5

ENUSA-7 UO2 707 13 3.95 0 49.0

ENUSA-8 UO2 3,389 67 3.95 0 38.3

The layout of assembly GN592 is shown in Figure 5. This assembly has 92 fuel rods including

12 part-length rods; nine of the rods contained Gd2O3 in fuel. Seven different uranium enrichments

are used in the assembly design. The enriched zone of the rods is 323.4 cm in length, with natural 

uranium axial blankets of 29.4 cm on the top and 15.2 cm on the bottom, resulting in a total active 

rod length of 368 cm. The general design specifications for the GE14 assembly GN592 are listed 

in Table 9.
5-10

Figure 5 Polaris Model of Forsmark Unit 3 GE14 10 × 10 Assembly

Detailed time-dependent reactor operating data, including void fraction, fuel temperature, and 

power for the measured samples, are documented in reference reports prepared by Vattenfall in 

Sweden [39].

Destructive and nondestructive assay measurements of fuel rod J8 were performed at the 

Studsvik Nuclear [40] Laboratory in 2009. HPLC was used for chemical separations for the 

analysis of individual isotopes of more than 10 different elements. Mass spectrometry was 

performed using ICPMS with ID analysis for most isotopes. Nuclides without any isobaric overlap 

were measured by ICPMS with external calibration analysis. Gamma-emitting nuclides were 

analyzed by gamma spectrometry. The measurements include more than 60 isotopes and include 

most of the burnup credit isotopes listed in Table 2.
5-11

Table 9 Forsmark 3 GE14 10 × 10 Assembly Parameters [41]

Reactor and Assembly Data

Reactor Forsmark Unit 3

Operating pressure (bar) 70.0

Assembly type 10 × 10 GE14

Fuel assembly pitch a (cm) [15.4]

Number of fuel rods 92

Number of gadolinia rods 9

Number of water rods 2

Fuel rod pitch (cm) 1.295

Channel bypass gap a (cm) [0.688 /1.051]

Channel thickness a (cm) [0.165 / 0.254]

Channel corner radius a (cm) [0.87376]

Channel material Zirc-2

Channel temperature b (K) 560

Active fuel length (cm) 368

Fuel Rod Data

Pellet radius (cm) 0.438

Pellet material UO2

Pellet density (g/cm3) 10.50

Fuel temperature (K) 792

Clad inner radius (cm) 0.447

Clad outer radius (cm) 0.513

Clad material Zirc-2

Clad temperature b (K) 560

Water Rod Data

Inner radius a (cm) [1.1685]

Outer radius a (cm) [1.2445]

Water rod material Zirc-2

Water rod temperature b (K) 560 a Proprietary value b Assumed to be at core average coolant temperature

5.4 Fukushima Daini 1 (9 × 9-9)

As part of a validation study of burnup calculations of BWR cores conducted by the Japan Nuclear 

Energy Safety (JNES) organization, now the Japan Regulation Authority (NRA), physics and 

depletion analyses were performed using post-irradiation measurements of burnup and isotopic 

inventories of eight samples taken from two 9 × 9-9 BWR lead test fuel assemblies irradiated in 

the Fukushima Daini Unit 1 reactor (2F1). Assembly 2F1ZN2 was discharged on May 9, 2000,

after three cycles of irradiation. Assembly 2F1ZN3 was irradiated for five cycles and discharged 

on January 7, 2003. This assembly design is similar to the ATRIUM-9 design.

Measurements for isotopes of uranium, plutonium, and neodymium for eight samples selected 

from five different fuel rods of the two assemblies were reported by Yamamoto [42]. An additional 

five samples from the same rods were later reported by Suzuki [43] that included measurements 

of additional fission products. Supplementary design and operating information necessary to 

model the 9 × 9-9 assemblies was provided by Yamamoto [44] through the OECD/NEA

coordinated activity on spent fuel assay data; these data and reports are currently included in the 
5-12

SFCOMPO database. The additional data included the fuel rod enrichment layout, timedependent void fractions, and accumulated burnup for the assemblies at the axial locations 

(nodes) of all measured samples.

The measurements included both UO2 and UO2-Gd2O3 type fuel rods with initial enrichments of 

2.1, 3.0, and 4.9 wt% 235U. The C2 fuel rods contained Gd2O3 with a content of 5 wt% in the fuel. 

Sample burnup values ranged from 29 GWd/MTU (UO2-Gd2O3) to 68.4 GWd/MTU (UO2). The 

sample burnup was estimated using the measured 148Nd concentration. Measurements were 

obtained for the bottom, middle, and top elevations of the rods. A summary of the measured 

sample characteristics is given in Table 10. The axial elevations of each sample are relative to the 

bottom of the active region of the fuel rod.

Table 10 Summary of Fukushima Daini-1 9 × 9-9 Assembly Fuel Sample Measurements

Assembly

ID

Rod

ID

Sample

ID Fuel type

Axial 

height 

(mm)

Avg.

Void 

(%)

Enrichment

(wt % 235U)

Gd content

(wt % Gd2O3)

Burnup

(GWd/MTU)

2F1ZN2

C2 GDB UO2-Gd2O3 757 18 3.0 5.0 35.6

GDT UO2-Gd2O3 2,922 74 3.0 5.0 29.0

C3 UB UO2 788 18 4.9 0 46.5

UT UO2 2,922 74 4.9 0 38.9

2F1ZN3

A9

UB UO2 788 18 2.1 0 61.2

UM UO2 1,654 38 2.1 0 68.0

UT UO2 2,844 60 2.1 0 55.7

C2

GDB UO2-Gd2O3 804 18 3.0 5.0 55.6

GDM UO2-Gd2O3 1,654 38 3.0 5.0 57.7

GDT UO2-Gd2O3 2,875 60 3.0 5.0 46.8

C3

UB UO2 788 11 4.9 0 68.3

UM UO2 1,639 38 4.9 0 68.4

UT UO2 2,844 60 4.9 0 58.0

The configuration of assemblies 2F1ZN2 and 2F1ZN3 is shown in Figure 6 with the measured rod 

locations C2, C3, and A9 highlighted. The measured assemblies use five different 235U 

enrichments and contain 12 Gd2O3 fuel rods, as indicated by the different colored rods in the 

figure. The general design specifications of the 9 × 9-9 assemblies are listed in Table 11.

Post-irradiation examination and destructive radiochemical analysis of the fuel samples were 

performed at the JAEA and resulted in high-accuracy isotopic inventory data for the irradiated 

samples. Measurements for isotopes of uranium (234U, 235U, 236U, 238U), plutonium (238Pu, 239Pu, 

240Pu, 241Pu, 242Pu) and neodymium (142Nd, 143Nd, 144Nd, 145Nd, 146Nd, 148Nd, 150Nd) were 

performed using TIMS with ID analysis. Isotopes of samarium, europium, and gadolinium were 

measured using ID and ICPMS. The metallic fission product isotopes 95Mo, 99Tc, 101Ru, 103

Rh, 

109Ag, and 133Cs were measured by ICPMS using the external calibration method. Measurements 

of assemblies 2F1ZN2 and 2F1ZN3 were made in 2002 and 2005, respectively.
5-13

Table 11 Fukushima Daini-1 9 × 9-9 Assembly Parameters

Reactor and Assembly Data

Reactor Fukushima Daini Unit 1

Operating pressure (bar) 70.7

Assembly type 9 × 9-9

Fuel assembly pitch (cm) 15.2

Number of fuel rods 72

Number of gadolinia rods 12

Number of water rods 1 (square channel)

Fuel rod pitch (cm) 1.45

Channel bypass gap (cm) 0.65

Channel thickness (cm) 0.25

Channel corner radius a (cm) 0.9652

Channel material b Zirc-2

Channel temperature c (K) 559

Active fuel length (cm) 371

Fuel Rod Data

Pellet radius (cm) 0.47

Pellet material UO2 / Gd2O3-UO2

Pellet density (g/cm3) 10.63 / 10.38

Fuel temperature (K) 900

Clad inner radius (cm) 0.49

Clad outer radius (cm) 0.55 

Clad material Zirc2

Clad temperature c (K) 559

Water Rod Data

Inner width d (cm) 3.685

Outer width d (cm) 3.85

Water rod material Zirc-2

Water rod temperature b (K) 559 a Assumed to be same as 8 × 8 assembly [37] b Assumed value based on earlier designs c Assumed to be at core average coolant temperature d Assumed to be same as the ATRIUM-9 fuel design [45]
5-14

Figure 6 Polaris Model of the Fukushima Daini-1 9 × 9-9 Assemblies

5.5 Fukushima Daini 2 (8 × 8-2)

Under a burnup credit research project at the Japan Atomic Energy Research Institute (JAERI) 

and supported by the Science and Technology Agency of Japan in cooperation with the utilities, 

experiments were performed on spent fuel assemblies to obtain criticality data, and destructive 

and nondestructive measurements were made to determine the nuclide compositions of the fuel 

[46]. Analyses of these data by Japan have been reported by Nakahara [8] and Yamamoto [47].

The measurements and the reference reports have been compiled as part of the SFCOMPO 

database [48].

Measurements are reported for two fuel rods from lattice positions B2 and C2 of an 8 × 8 

assembly identified as 2F2DN23 irradiated for three cycles in Unit 2 the Fukushima Daini Power 

Station 2 (2F2) reactor, operated by Tokyo Electric Power Company (TEPCO). Assembly 

2F2DN23 was discharged on November 14, 1992. The measured fuel rods from assembly lattice 

positions B2 and C2 were identified as SF98 and SF99 in the research project. The rod 

designated SF98 was a UO2 fuel with an enriched zone of 3.90 wt% 235U with natural uranium 

blankets having a nominal length of 155 mm at the top and bottom of the rod. Rod SF99 was a 

UO2-Gd2O3 rod with two axial enrichment zones and two natural uranium blankets. The lower 

2,937 mm section of the enriched zone had 3.40 wt% 235U and 4.5 wt% Gd2O3 enrichment 

whereas the upper section had 3.40 wt% 235U and 3.0 wt% Gd2O3 enrichment. 
5-15

Isotope measurements were reported for 18 different samples obtained from different axial 

positions of the two rods. Three samples were selected from the natural uranium blanket regions 

of the rods. The sample characteristics are listed in Table 12. The sample axial locations in the 

fuel rods were measured from the bottom of the active fuel length. The burnup values were 

determined using the measured 148Nd content in the fuel samples.

The configuration of assembly 2F2DN23 is shown in Figure 7. The 8 × 8-2 assembly is similar to 

the GE7 design. There are six different enrichments (excluding the natural uranium regions) used 

in the assembly, as shown by the different colored rods in the figure. The experimental reports 

include assembly design information and provide time-dependent specific power for each sample 

location and the average void fraction as a function of axial location in the assembly. Timedependent void data was not available for this assembly. The average void fractions are those 

provided by TEPCO and are standard values as written in the Application for Permission for the 

Installation of a Nuclear Reactor [8]. The impact of using average void compared to detailed void 

data was studied and found to be minor, as evaluated in Section 7 of this report. The general 

assembly design information is summarized in Table 13.

Destructive assay measurements of the fuel samples from rod B2 and C3 were performed at 

JAEA laboratories in May 1998 and July 1999, respectively. Fuel samples with a thickness of 

0.5 mm were dissolved and measured using chemical separations and mass spectrometry with 

alpha counting. Isotopic compositions and absolute contents of uranium, plutonium, neodymium, 

and samarium isotopes were measured by isotopic dilution mass spectrometry (IDMS) using 

spikes of 150Nd, 233U, and 242Pu. Concentrations of 234U, 237Np, 238Pu, 241Am, 243Am, 242Cm, and 

244Cm were measured as ratios to uranium by alpha spectrometry.

Table 12 Summary of Fukushima Daini-1 8 × 8-2 Assembly Fuel Sample Measurements

Assembly

ID

Rod

ID

Sample 

ID Fuel type

Axial 

height 

(mm)a

Avg. 

Void 

(%)

Enrichment

(wt % 235U)

Gd content

(wt % Gd2O3)

Burnup

(GWd/MTU)

2F2DN23

B2

SF98-1 UO2 39 0 0.71 0 4.2

SF98-2 UO2 167 0 3.91 0 26.5

SF98-3 UO2 423 3 3.91 0 36.9

SF98-4 UO2 692 11 3.91 0 42.4

SF98-5 UO2 1,214 32 3.91 0 44.0

SF98-6 UO2 2,050 54.5 3.91 0 39.9

SF98-7 UO2 2,757 68 3.91 0 39.4

SF98-8 UO2 3,397 73 3.91 0 27.2

C2

SF99-1 UO2 134 0 0.71 0 7.5

SF99-2 UO2-Gd2O3 286 1.4 3.4 4.5 22.6

SF99-3 UO2-Gd2O3 502 5.8 3.4 4.5 32.4

SF99-4 UO2-Gd2O3 686 10.8 3.4 4.5 35.4

SF99-5 UO2-Gd2O3 1,189 27.7 3.4 4.5 37.4

SF99-6 UO2-Gd2O3 2,061 54.7 3.4 4.5 32.4

SF99-7 UO2-Gd2O3 2,744 66.5 3.4 4.5 32.1

SF99-8 UO2-Gd2O3 3,388 71.7 3.4 3.0 21.8

SF99-9 UO2-Gd2O3 3,540 72.9 3.4 3.0 16.7

SF99-

10 UO2 3,676 74.3 0.71 0 7.2

a Measured from the bottom of the active fuel length.
5-16

Figure 7 Polaris Model of the Fukushima Daini-2 8 × 8-2 Assemblies
5-17

Table 13 Fukushima Daini-2 8 × 8-2 and 8 × 8-4 Assembly Parameters

Reactor and Assembly Data

Reactor Fukushima Daini Unit 2

Operating pressure (bar) 70.7

Fuel assembly pitch (cm) 15.23

Assembly type 8 × 8-2 8 × 8-4

Number of fuel rods 62 60

Number of gadolinia rods 8 8

Number of water rods 2 1

Fuel rod pitch (cm) 1.625 1.625

Channel bypass gap (cm) 0.65 0.65

Channel thickness (cm) a 0.2032 0.2032

Channel corner radius a (cm) 0.9652 0.9652

Channel material b Zirc-2 Zirc-2 

Channel temperature c (K) 559 559

Active fuel length (cm) 371 371

Fuel Rod Data

Pellet radius c (cm) 0.515 0.519

Pellet material UO2 / Gd2O3-UO2 UO2 / Gd2O3-UO2

Pellet density d (g/cm3) 10.48 10.61

Fuel temperature (K) 900 900

Clad inner radius (cm) 0.527 0.529

Clad outer radius (cm) 0.613 0.615

Clad material Zirc-2 Zirc-2

Clad temperature b (K) 559 559

Water Rod Data

Inner radius d (cm) 0.6738 1.623

Outer radius (cm) 0.75 1.7

Water rod material Zirc-2

Water rod temperature cb (K) 559 a Assumed to be the same as for the 8 × 8 assembly [37] b Assumed value based on earlier designs c Assumed to be at core average coolant temperature d From IAEA TECDOC-849 [45]

5.6 Fukushima Daini 2 GE9 (8 × 8-4)

Isotopic measurements of four BWR 8 × 8-4 lead test assemblies, irradiated in Unit 2 of the 

Fukushima Daini Power Station 2 (2F2), were report by the Japan NRA [47]. The assemblies, 

identified as 2F2D1, 2F2D2, 2F2D3, and 2F2D8, were discharged on March 8, 1990, August 24, 

1991, November 14, 1992, and January 30, 1997, after one, two, three, and five cycles of 

irradiation, respectively, providing a wide range of sample burnups. Supplementary detailed 

design information necessary for modeling these assemblies was provided by Yamamoto [48]. 

This information included the enrichments of all fuel rods in the assembly. The measurements, 

design data, and reference reports are captured in the SFCOMPO database. 

All assemblies have an 8 × 8-4 lattice arrangement with a large water rod that occupies the 

position of four lattice positions, similar to the GE9 assembly design. The assemblies used five 

different 235U enrichments and included eight UO2-Gd2O3 rods with Gd2O3 contents of 3.0 and 
5-18

4.5 wt% in fuel. Measurements for each assembly include both UO2 and UO2-Gd2O3 type fuel 

rods. The sample characteristics are given in Table 14. These data cover a wide range of burnup 

and void fraction values.

The configuration of the assembly is shown in Figure 8. All assemblies have the same layout and 

enrichment zoning. The general assembly design parameters for the 8 × 8-4 assembly are similar 

to the 8 × 8-2 design and are listed in Table 13.

Time-dependent void distributions for the 8 × 8-4 assemblies were not reported. However, the 

node average values of the channel void fractions of the 8 × 8-4 fuel assemblies were provided by 

the plant operator for all axial nodes of the assemblies which included the measured fuel samples

[47]. 

Measurements made at laboratories of the Nippon Nuclear Fuel Development (NFD) Company

are reported for isotopes of uranium, plutonium, and curium, as well as for 148Nd, 241Am and 

243Am. The method of IDMS was applied to neodymium and uranium isotopes with spikes of 150Nd 

and natural uranium, respectively. Alpha spectrometry and mass spectrometry were used for the 

plutonium isotopes, and alpha spectrometry was used to measure the americium isotopes. The 

sample burnups estimated by the laboratory were based on the 148Nd method with the inventory 

data of uranium, plutonium. The 148Nd measurement had a reported uncertainty of 6.4%. The 

burnup values used in this study used the measured 148Nd content in each sample.

Table 14 Summary of Fukushima Daini-2 8 × 8-4 Assembly Fuel Sample Measurements

Assembly

ID

Rod

ID

Sample

ID Fuel type

Axial 

height 

(mm)a

Avg. 

Void 

(%)

Enrichment

(wt % 235U)

Gd content

(wt % Gd2O3)

Burnup

(GWd/MTU)

2F2D1

F6 TU101 UO2 3,378 64.0 4.5 0 14.0

TU102 UO2 642 12.9 4.5 0 18.2

B3 TU103 UO2-Gd2O3 3,343 64.0 3.4 4.5 10.0

TU104 UO2-Gd2O3 2,743 60.2 3.4 4.5 9.4

TU105 UO2-Gd2O3 740 17.3 3.4 4.5 12.3

F6 TU106 UO2 2,689 59.8 4.5 0 16.1

2F2D2

F6 TU201 UO2 3,178 63.1 4.5 0 29.1

TU202 UO2 478 7.0 4.5 0 32.9

B3 TU203 UO2-Gd2O3 3,178 63.1 3.4 4.5 24.5

TU204 UO2-Gd2O3 2,592 58.5 3.4 4.5 23.5

TU205 UO2-Gd2O3 578 10.4 3.4 4.5 22.8

2F2D3

H5 TU301 UO2 2,793 60.6 3.4 0 34.6

TU302 UO2 423 5.2 3.4 0 31.4

A4 TU304 UO2 2,856 61.0 3.4 0 37.8

TU306 UO2 447 6.0 3.4 0 32.3

B3 TU308 UO2-Gd2O3 3,242 63.5 3.4 4.5 30.2

TU309 UO2-Gd2O3 2,780 60.5 3.4 4.5 34.8

TU311 UO2-Gd2O3 543 9.1 3.4 4.5 33.5

2F2D8

H5 TU501 UO2 3,202 63.2 3.4 0 53.2

TU502 UO2 2,453 58.0 3.4 0 58.9

TU503 UO2 803 20.6 3.4 0 55.6

A4 TU505 UO2 2,229 54.9 3.4 0 59.1

TU506 UO2 850 23.0 3.4 0 57.5

B3 TU510 UO2-Gd2O3 2,952 62.2 3.4 4.5 53.1

TU511 UO2-Gd2O3 670 14.0 3.4 4.5 48.1 a Measured from the bottom of the active fuel length.
5-19

Figure 8 Polaris Model of the Fukushima Daini-2 8 × 8-4 Assemblies

5.7 Leibstadt SVEA-96 (10 × 10)

Measurements of isotopic concentrations from the MALIBU (MOX AND UOX LWR Fuels 

Irradiated to High Burnup) international experimental program [12] were analyzed in this study.

MALIBU is a commercial proprietary program managed by the Belgian Nuclear Research Center, 

Studiecentrum voor Kernenergie–Centre d’Études de l’Énergie Nucleaire (SCK•CEN). This

proprietary program included participants from Belgium, Sweden, Japan, France, Germany, 

Switzerland, and the United States. Measurements were performed at several radiochemical 

laboratories as a measurement cross check and to assess and reduce uncertainties.

Under an extension phase of the program, isotopic measurements were made on BWR fuel 

samples from a SVEA-96 Optima 10×10 assembly from the Kernkraftwerk Leibstadt reactor in 

Switzerland [49]. The measured assembly, identified as AIA003, was irradiated for seven cycles 

and was discharged in March 28, 2005.

Three samples were taken at different axial positions of rod H6 of assembly AIA003 to assess 

different void conditions. All samples had an initial enrichment of 3.90 wt% 235

U. Characteristics of 

the measured samples are given in Table 15. The burnup values for samples KLU1 and KLU3 

were determined using the 148Nd concentration; this burnup was in good agreement with burnup
5-20

estimates based on other neodymium isotopes and 137Cs. The KLU2 sample used 145+146Nd and 

137Cs measurements to estimate the sample burnup, which was about 8% different from the 

burnup obtained using 148Nd.

The SVEA-96 assembly design has a highly heterogeneous 10 × 10 lattice array with four 

subassembly regions similar to the SVEA-100 design. The fuel rod arrays are not exactly regular, 

with the outer fuel rods shifted slightly inwards from the edge of the assembly, and one fuel rod 

from the inner corner of each sub-assembly removed to accommodate a central square water 

channel. The assembly includes eight part-length rods located nearest the water channel. The fuel 

rods have seven different 235U enrichments, and 12 of these rods contain Gd2O3 with a content of 

4 wt% in the fuel. The fuel rods have two different diameters. The assembly design is shown in 

Figure 9 for the configuration of the dominant lattice (below the level of the part length rods) with 

different colors identifying different types of fuel rods in the assembly. The general assembly 

design parameters for the SVEA-96 assembly are listed in Table 16.

Detailed operating data including time-dependent specific power, void conditions, and fuel 

temperatures were provided by the Vattenfall Nuclear Fuel and Kernkraftwerk Leibstadt [50].

All samples were measured at Studsvik Nuclear Laboratory in Sweden during 2010. The sample 

at the lowest elevation, KLU1, was selected as a cross check sample and was also analyzed at 

the laboratories of SCK•CEN in Belgium and the PSI in Switzerland. Radiochemical analysis 

techniques were used to analyze more than 50 actinides and fission products. Mass spectrometry 

measurements at SCK•CEN were performed using TIMS in 2009 and 2010, and PSI used HPLC 

and ICPMS techniques and measurements were performed in 2010.

Table 15 Summary of Leibstadt SVEA-96 10 × 10 Assembly Fuel Sample Measurements

Assembly

ID

Rod

ID

Sample 

ID

Fuel 

type

Axial 

height 

(mm)a

Avg.

Void 

(%)

Enrichment

(wt % 235U)

Gd content

(wt % Gd2O3)

Burnup

(GWd/MTU)

AIA003 H6 KLU1 UO2 588 8.4 3.90 0 60.5

KLU2 UO2 1,922 51 3.90 0 65.0

KLU3 UO2 3,302 70 3.90 0 58.4 a Measured from the bottom of the active fuel length.
5-21

Figure 9 Polaris Model of Leibstadt SVEA-96 10 × 10 Assembly
5-22

Table 16 Leibstadt SVEA-96 10 × 10 Assembly Parameters

Reactor and Assembly Data

Reactor Leibstadt

Operating pressure (bar) 74

Assembly type 10 × 10 

SVEA-96 Optima

Fuel assembly pitch (cm) 15.24

Number of fuel rods 96

Number of gadolinia rods 12

Number of water rods 1 (square channel)

Fuel rod pitch (cm) a [1.27]

Channel bypass gap (cm) d 0.69

Channel thickness (cm) a [0.14]

Channel corner radius (cm) a [1.13]

Channel material Zirc-2

Channel temperature (K) b 562

Active fuel length (cm) 381

Fuel Rod Data

Pellet radius (cm) d 0.4095

Pellet material UO2

Pellet density (g/cm3) 10.52

Fuel temperature (K) c 800

Clad inner radius (cm) d 0.418

Clad outer radius (cm) d 0.481

Clad material Zirc-2

Clad temperature (K) b 562

Water Rod Data

Inner width (cm) a [2.74]

Outer width (cm) a [2.90]

Water rod material Zirc-2

Water rod temperature (K) b 562 a Proprietary value b Assumed to be at core average coolant temperature c Average value d Reference [51]

5.8 Limerick 1 GE11 (9 × 9-7)

Isotopic composition measurements of a spent fuel assembly from the Limerick Unit 1 reactor 

were measured in laboratories at GE Vallecitos Nuclear Center. These measurements have been 

analyzed in previous validation studies performed under the Yucca Mountain Project (YMP) in 

2004 under the Office of Civilian Radioactive Waste Management [13]. Measurements were 

performed for eight samples selected from a high-burnup assembly identified as YJ1433 [52]. The 

reported measurement data include nuclide concentrations for 32 actinides and fission products.

Assembly YJ1433 is a GE11 9 × 9 design with two large water rods. There are five different 235U 

enrichments for the UO2 rods, eight part-length rods, and nine rods containing Gd2O3 at 5 wt% in 

the fuel. The assembly configuration is shown in Figure 10. The assembly was irradiated for three 

cycles and was discharged after cycle 7 on May 27, 1998. The general design specifications for 

the GE11 assembly YJ1433 are listed in Table 18.
5-23

Three different fuel rods were measured, including a full length UO2 rod from lattice location D9, a 

UO2-Gd2O3 rod from location D8, and a part-length UO2 rod from location H5. The characteristics 

of the measured samples are listed in Table 17.

The burnups assigned to these samples are based on values determined by GE Nuclear Energy

[52]. These burnup values were determined using uranium, plutonium, and neodymium isotope 

ratios. However, for some samples, large deviations, of up to 7%, are observed between 

measured and calculated 148Nd content, a common burnup indicator. Adjusting the calculations to 

match the measured 148Nd resulted in inconsistencies in other nuclides, and this inconsistency in 

sample burnup has not been resolved. The impact of uncertainties in the estimated sample 

burnup values is assessed later in this report.

Table 17 Summary of Limerick GE11 9 × 9 Assembly Measurements

Assembly

ID

Rod

ID

Sample

ID

Fuel

type

Axial

height 

(mm)

Avg.

Void

(%)

Enrichment

(wt % 235U)

Gd 

content

(wt % 

Gd2O3)

Burnup

(GWd/MTU)

YJ1433

D8 D8-3D2 UO2-Gd2O3 823 54.8 3.6 5.0 54.4

D8-4G3 UO2-Gd2O3 1,301 68.8 3.6 5.0 37.0

D9

D9-1D2 UO2 308 12.1 3.95 0 62.1

D9-2D2 UO2 623 44.1 3.95 0 65.5

D9-4D4 UO2 823 65.4 3.95 0 64.9

D9-4G1E1 UO2 1,305 69.1 3.95 0 56.5

H5 H5-3A1C UO2 308 54.8 3.95 0 57.9

H5-3A1G UO2 623 57.7 3.95 0 57.8 a Measured from the bottom of the active fuel length.

Measurements performed by the laboratory at GE Vallecitos were made primarily using IDMS

using TIMS. The measured nuclides include isotopes of U, Pu, Nd, Gd, Sm, Eu, Am, Cm, Np, and 

Cs. Alpha spectrometry and mass spectrometry were used to measure 238Pu, the americium 

isotopes, and 237Np.

The Limerick measurements were previously evaluated under the YMP project using depletion 

codes employing both 1D transport models [13] and 2D models [53]. The detailed design 

information for the GE11 assembly and operating history data for assembly YJ1433 are currently 

not public. The calculations reported in this study include restricted information that is only

available under a nondisclosure agreement.

The void fraction data documented with the Limerick data are not based on detailed core 

simulation codes but were instead developed using core average void fractions and semiempirical 

correlations to estimate axial and time-dependent void levels for the measured assembly, 

potentially introducing additional uncertainty.
5-24

Figure 10 Polaris Model of Limerick-1 GE11 9 × 9 Assembly
5-25

Table 18 Limerick-1 GE11 9 × 9 Assembly Parameters

Reactor and Assembly Data

Reactor Limerick Unit 1

Operating pressure (bar) 70

Assembly type GE11 9 × 9

Fuel assembly pitch (cm) 15.24

Number of fuel rods 74

Number of gadolinia rods 9

Number of water rods 2

Fuel rod pitch (cm) a 1.438

Channel bypass gap (cm) b [0.726]

Channel thickness (cm) b [0.1905]

Channel corner radius (cm) b [1.143]

Channel material Zirc-4

Channel temperature (K) c 560

Active fuel length (cm)

Fuel Rod Data

Pellet radius (cm) d 0.471

Pellet material UO2 / Gd2O3-UO2

Pellet density (g/cm3) 10.45

Fuel temperature (K) 1100

Clad inner radius (cm) a 0.4878

Clad outer radius (cm) a 0.559

Clad material Zirc-2

Clad temperature (K) c 560

Water Rod Data

Inner radius (cm) b [2.337]

Outer radius (cm) b [2.489]

Water rod material Zirc-2

Water rod temperature (K) 560 a Reference [54] b Proprietary value c Assumed to be at core average coolant temperature d Reference [55]

5.9 Quad Cities GE11 (9 × 9)

Measurements of fuel rods from lead test assembly LYD449X, a BP8 × 8R design irradiated in 

Quad Cities Unit 1, have been reported by the GE Vallecitos Laboratory [56][57] and by Argonne 

National Laboratory [58]. Extensive isotopic measurements were sponsored by the Office of 

Civilian Radioactive Waste Management to support burnup credit in post-closure criticality 

evaluations for the Yucca Mountain Project. Twelve samples from fuel rod segment numbers 

VW00101, ZB00113, and ZS00102 were measured by Argonne. Six samples from the same rod 

segments were measured by GE Vallecitos.

Detailed assembly design and operating data for the measured assembly however have not been 

made publicly available. Other information indicates that assembly LYD449X was reconstituted 

(rebuilt) with previously irradiated rods [59]. No additional information on the fuel rod 

reconfiguration has been identified. Due to the lack of operating information and assembly 
5-26

reconfiguration details, these measurements were not used in the present study. The Quad Cities 

experiment is cited in this report only for completeness and to indicate that these measurements

were considered for validation and could potentially be used in the future if additional irradiation 

data were made available.
6-1

6 CRITICALITY APPLICATION MODELS

Methods and experimental data described to this point are applicable to any criticality application 

model as long as diversity and availability of the selected measured isotopic compositions cover

the application range.

The application model used in this report to assess the impact of predicted nuclide uncertainties 

on the system keff is based on a generic 68-assembly cask design containing GE14 10 × 10 

assemblies. This section describes the assembly and cask models and the modeling 

assumptions. The methods and experimental data described in this report can be easily extended 

to other application models. 

6.1 GE14 Assembly

The GE14 fuel assembly design was used as the reference design in the criticality analysis model. 

This assembly has a 10 × 10 array of fuel rods and contains two large central water rods that 

displace eight fuel rods from the array. The GE14 assembly uses varying fuel rod enrichments

(Figure 5) and includes fuel rods containing gadolinium. In addition, enrichments can vary axially. 

The assembly uses part-length fuel rods, which terminate at approximately half the total height of 

the fuel assembly, resulting in two different fuel rod lattice configurations at different axial levels in 

the assembly. These two axial regions below and above the top end of the part-length rods are 

referred to as the dominant or full (DOM) lattice and vanished (VAN) lattice. The vanished lattice is 

located axially above the part-length rods, so these part-length rods are in effect removed from 

the lattice. Two-dimensional representations of the two lattice regions in the KENO V.a model are 

shown in Figure 11. 

 

(a) (b)

Figure 11 (a) DOM and (b) VAN Lattice of the GE14 Assembly (Shown in a Dry Cask

 Storage Cell)
6-2

The GE14 fuel assembly is used as the reference design for these studies since it is a common 

assembly type used in US BWRs and includes advanced geometry features seen in modern BWR 

fuel assemblies (e.g., large water rods, part-length rods, relatively high enrichment, and use of

gadolinium-bearing fuel rods). A previous study [60] has shown that the GE14 10 × 10 assembly 

design is more reactive than smaller GE lattices (7 × 7, 8 × 8, and 9 × 9) for most burnups. 

6.2 GBC-68 Cask

The computational benchmark model developed in Reference [41] as a generic burnup credit

(GBC) cask containing 68 BWR assemblies (GBC-68 cask) was used to quantify the impact of 

isotopic bias and uncertainty in the criticality analysis. The cask is modeled using the KENO V.a

Monte Carlo criticality code (Figure 12). This cask model is based on the same model used in 

previous BWR burnup credit studies [61].

The GBC-68 cask model uses GE14 assemblies with full-length and part-length fuel rods. The 

KENO V.a model of the GE14 assemblies explicitly represents each fuel rod, including the fuelcladding gap and cladding. Part-length rods are truncated at the appropriate elevation so that both 

the full lattice (DOM) and the vanished lattice (VAN) are included explicitly in the KENO V.a

model. The assembly channel is approximated in the model with constant thickness and square 

corners.

The GBC-68 cask model assumes that all fuel rods contain the same nuclide compositions both 

axially and radially. Similar assumptions were used previously in this application model [61]. Axial 

variations (i.e., the natural uranium blanket regions) or enrichment zoning of the fuel rods in the 

assembly were not included, but UO2-Gd2O3 fuel rod compositions were explicitly modeled for the 

analysis of axial burnup distributions. Homogenization of the all fuel rods was also used previously 

in the analysis of void fraction and control blade effects [61].

The requirement to homogenize the fuel compositions in the KENO V.a calculations in the present 

analysis is imposed by the isotopic uncertainty analysis methodology. Criticality calculations are 

performed for the GE14 assembly design and GBC-68 application model using measured isotopic 

contents for the 77 spent fuel samples selected and described in Section 4. These samples span

a wide range of enrichments, burnups, and void fractions and include measurements of both UO2

and UO2-Gd2O3 type fuels. However, the compositions are constrained to those of the measured 

samples, and therefore a more spatially detailed application model is not possible. 

The present analysis does not model gadolinium-bearing fuel rods explicitly. Because of this, the 

results are only applicable to BWR burnup credit beyond peak reactivity where any initial 

gadolinium present in the fuel has been fully depleted. In addition, the lowest-burnup-measured 

UO2-Gd2O3 fuel samples used in this study (Fukushima Daini 2 assembly 2F2D1 rod B3) did not 

measure gadolinium isotopes. Any initial 155Gd or 157Gd present in UO2-Gd2O3 fuel rods is 

assumed to be fully depleted in these analyses.
6-3

Figure 12 Radial View of the KENO V.a GBC-68 Cask Model (Elevation of Vanished

 Lattice)

Stainless steel

Water

7-1

7 UNCERTAINTY ANALYSIS

The most widely used approach to burnup credit validation involves validating the two separate 

components of the criticality safety analysis; components related to the prediction of nuclide 

compositions and components associated with the criticality calculation. Validation of the code 

prediction of nuclide compositions is routinely performed using experimental data from destructive 

radiochemical analysis of spent fuel samples and is the subject of this report. Validation of the 

criticality calculation itself is performed using applicable critical experiments and is reported 

separately [3]. 

7.1 Methodology

Several different approaches have been developed and used to assess the effect of uncertainties 

in predicted nuclide compositions on the keff of a criticality application model, each with different 

levels of complexity and conservatism. These methods include, but are not limited to: (1) the 

development of bounding factors [62] applied to adjust the calculated isotopic compositions for 

uncertainty to ensure conservative estimates of keff, (2) Monte Carlo sampling of the statistical 

distributions of the nuclide compositions [63] to obtain the uncertainty distribution of keff, and (3) 

direct application of measured and calculated nuclide compositions in the criticality model [7]. A 

review and comparison of these different approaches has been previously documented [20]. 

In the present study, the direct application of measured nuclide compositions in the criticality 

model is used. The basis for selecting this approach is discussed in more detail in Section 2 of 

this report. KENO V.a calculations are performed using the major actinide-only compositions 

(Table 1) and also actinide and fission product nuclide compositions (Table 2). Fuel compositions 

include the oxygen, which is part of the UO2 fuel matrix. KENO V.a criticality calculations are 

performed using the measured nuclide concentrations for each fuel sample. 

In parallel with these calculations, separate KENO V.a calculations are performed using isotopic 

concentrations obtained directly from the Polaris depletion calculation for the same samples. The

Polaris calculations are performed using assembly design, irradiation history data and 

measurement dates as reported in the experimental references. No adjustment of the irradiation 

parameters is used in the Polaris calculations to obtain conservative isotopic values. Therefore,

the nuclide concentrations calculated using Polaris represent best-estimate values. This 

procedure is illustrated in Figure 13. 

7.1.1 Measurements

The measured isotopic concentrations are those reported by the laboratory as isotopic content or 

isotopic ratios that have been converted to mass units per initial metric ton of heavy metal 

(uranium). Measured concentration units have been converted to units of milligram per gram of 

initial uranium (mg/gUi) to provide a consistent basis for comparison with calculations. Most of the 

experimental datasets included in the SFCOMPO database store measurements in units as 

reported by the laboratory and also in mg/gUi units.
7-2

Measurements for the Limerick-1 samples are reported in units of mg per gram of 238U in the 

measured sample. The reported units are converted to mg/gUi on the basis of the measurement 

data by the following formula:

mmii

UU

UU8 + PPPPUU8 + NNNN + AA + CCUU8 + FFUU8

(1)

where mi is the mass of measured isotope i as reported in mg/g 238U (U8), and the 

denominator is an experimental estimate of the initial uranium content derived from the sum of 

the heavy metal (uranium, plutonium, americium, curium) weights in the measured sample, 

corrected for the weight loss in initial uranium due to fission, F. In this equation the uranium 

mass U/U8 includes the contribution of 238U of 1,000 mg/ g 238U. The reduction in the initial 

heavy metal mass during burnup is estimated from the total number of fissions as determined 

from measured 148Nd by F/U8 = m (238/148)/ YY�, where YY� = 0.0170 is the effective 148Nd fission 

yield applied to UO2 fuel [49], and m is the measured 148Nd mass in mg/g 238U.

The measured isotopic data of the burnup credit isotopes (Table 2) are listed in Appendix A for all 

samples, with the cooling time of the measurements indicated. Measured results for the 

Dodewaard sample DU1 are included for both the SCK•CEN and PSI measurements.

7.1.2 Measurement dates

The cooling times used to compare measured and calculated nuclide concentrations are those 

reported by the laboratory, generally corresponding to the actual date of the measurements. For 

several experiments, measurements of different isotopes in a sample are reported for different 

dates. Adjustment of measurements to a common reference date is possible using decay 

relationships and has been performed in previous studies [6] and recommended in earlier 

guidance [1]. However, in order to perform these adjustments, measurements of any decay 

precursor nuclides (production terms) are also required. Because most samples used in this 

report were measured on the same date or on dates that are very similar relative to the half-lives 

of the nuclides and/or precursors, no adjustment of the measurements was performed in this 

study to avoid potentially introducing additional uncertainty in the measured concentrations. The 

impact of not performing time corrections was analyzed for the specific samples used in this study

and was found to be less than 30 pcm. This bias is insignificant relative to other uncertainties.

Measurements for Fukushima Daini 2 assembly 2F2DN23 were reported at the time of discharge 

from the reactor (no cooling time), with the exception of the samarium isotopes that were reported 

at the time of measurement (cooling time approximately 6 years). No adjustment of samarium is 

possible without also measuring the decay precursors of promethium. Care is needed in analyzing 

measurements reported at discharge to ensure that any short-lived precursors of the measured 

nuclide are included in the calculations, since these decay contributions cannot be subtracted 

from the measured concentration. For example, the concentration of 239Pu as reported at 

discharge includes the contribution from 239Np (T1/2 = 2.35 days) decay. Therefore, using 

calculated concentrations at discharge will miss this contribution to the 239Pu inventory. In the 

present study, a decay time of 500 days was added after irradiation to the calculations to ensure 

that short-lived precursors were included in the predictions to be consistent with measurements.
7-3

Figure 13 Uncertainty Analysis Methodology for Nuclide Compositions

 (Adapted from [19])

7.1.3 Partial Isotopic Datasets

The calculation procedure in Figure 13 was applied to all samples using the major actinide-only 

nuclides and the actinide-plus-fission product datasets. Measurements of all major actinides 

(Table 1) were available for most samples considered in this report, with the exception of 

14 samples that did not measure 241Am. Measurements for minor actinides and fission products 

(Table 2) were available for a limited number of samples.
7-4

To minimize inconsistencies that could arise by comparing keff values for samples with different 

sets of nuclides (as available in the measured samples), a common set of minor actinides and 

fission products (Table 2) were used in all criticality calculations. To account for nuclides not 

measured in some samples, calculated concentrations were used as a surrogate for measured 

data. To account for bias in the calculated concentrations, the median bias derived from other 

samples with measured data was applied. To account for uncertainty in this estimated 

concentration, the keff calculation was also performed using surrogate nuclide concentrations 

adjusted conservatively to maximize keff and also to minimize the keff in the cask model. The 

bounding keff values were obtained by performing adjustments for uncertainty in the bias-corrected 

calculated nuclide concentrations using the 10th percentile and the 90th percentile of the 

distribution (see Sect. 7.2) of deviations obtained between calculated and measured nuclide 

concentrations. Therefore, in addition to obtaining the keff bias value for each sample, this 

approach introduces errors bars for each sample due to the application of derived measurement 

data from the calculations to illustrate the potential impact of adding surrogate data to the keff

calculations.

7.2 Isotopic Bias and Uncertainty

The calculated isotopic concentrations of all nuclides considered in the burnup credit analysis 

methodology (Table 2) were compared to measured concentrations obtained by destructive 

radiochemical analysis of the fuel samples. These comparisons were performed at the time of 

measurement of each isotope (Appendix B). 

One sample from the Fukushima Daini-2 assembly 2F2DN23, sample SF99-10, was not included 

in the analysis due to its very close proximity to the end of the active fuel length. The results for 

this sample exhibit large biases that are attributed to the spectral change near the ends of the fuel 

rods which are not accounted for in the 2D models.

The deviations between the Polaris calculations (C) and measurements (M) are expressed as the 

relative percent difference (C/M – 1)%.The results for each individual sample and nuclide are 

listed in Appendix B, and a statistical analysis summary is presented in Table 19. The summary 

includes the total number of measurements available for each isotope, the mean deviation, the 

standard deviation, the median value, minimum and maximum deviations, the 1st and 3rd quartile 

(range contains 50% of the data points), and the percentiles for 10% (p10) and 90% (p90) of the 

data (range contains 80% of the data points).

The statistical analysis presented in Table 19 provides a summary of the isotopic results and 

illustrates the performance of the Polaris code used in the present study. However, it is important 

to note that these isotopic results are not essential for the uncertainty methodology used here, 

since the calculated and measured nuclide concentrations are used directly in the application 

model without any need to characterize the bias, uncertainty, or statistical distribution of individual 

nuclides. An exception is for nuclides (minor actinides and fission products) that are not measured 

in a particular sample. In these cases, surrogate nuclide concentrations are used in lieu of 

measurements to ensure that all criticality calculations use a consistent set of burnup credit 

nuclides.

In this study, the surrogate data are nuclide concentrations calculated with Polaris and adjusted to 

correct for the median bias using deviations observed in other samples. To account for the 

additional uncertainty in these surrogate data, additional criticality calculations are performed 
7-5

using surrogate data adjusted to the p10 and p90 percentiles to provide a conservative 

quantitative estimate of the additional uncertainty associated with samples that do not measure all 

burnup credit isotopes.

The isotopic distributions are presented as box plots in Figures 14, 15, and 16, showing the mean, 

median, quartiles, and box whiskers that represent the p10 and p90 percentiles and the min/max 

values in the distributions. The individual values for each sample are also shown. Maximum

values in 234U, 238Pu, 242Pu, 241Am, 243Am and 109Ag percent differences are above 60%, and not 

shown in the plots in order to display distribution details. These plots are nonparametric and make 

no assumptions of the statistical isotopic distributions for the samples and are based on the actual 

observed distributions. An outlier analysis of these distributions can be performed; however, in 

this study no data were rejected based on outlier analysis. 

Table 19 Statistical Analysis of Predicted Isotopic Concentrations (C/M-1) (%)

Data

Number of 

Measurements Mean

Standard

Deviation Median Minimum Maximum

1st

Quartile

(Q1)

3rd

Quartile

(Q3)

Percentile

p10a

Percentile

p90a

234U 76 6.8% 13.4% 5.5% -37.0% 66.6% 1.4% 9.9% -5.4% 20.6%

235U 76 4.3% 11.2% 2.8% -15.1% 36.5% -2.8% 9.1% -8.1% 22.0%

236U 76 1.6% 4.8% 1.2% -6.0% 15.7% -1.5% 3.2% -3.4% 7.7%

238U 76 -0.1% 0.3% -0.1% -0.8% 0.5% -0.2% 0.1% -0.6% 0.2%

238Pu 76 9.5% 21.1% 6.7% -38.8% 93.6% -1.9% 20.1% -18.1% 33.2%

239Pu 76 -0.9% 8.7% -1.0% -22.8% 22.7% -6.6% 4.4% -11.6% 10.2%

240Pu 76 -3.1% 8.4% -2.4% -28.3% 31.4% -6.9% -0.5% -12.6% 6.5%

241Pu 76 -3.3% 11.6% -2.4% -34.5% 40.1% -8.8% 2.9% -17.6% 9.2%

242Pu 76 1.2% 17.2% 1.2% -42.6% 87.9% -4.5% 9.6% -19.8% 15.1%

241Am 62 2.8% 17.4% 3.6% -50.3% 69.1% -5.0% 10.5% -15.2% 15.6%

243Am 62 1.3% 33.5% -7.5% -44.5% 122.8% -16.6% 4.9% -26.7% 49.9%

237Np 29 -2.4% 11.9% -5.7% -19.8% 46.4% -7.7% -1.1% -11.3% 6.1%

143Nd 50 4.5% 4.0% 3.9% -4.1% 13.1% 2.1% 7.0% 0.0% 10.8%

145Nd 50 2.5% 3.2% 1.4% -2.5% 11.8% 0.7% 3.6% -0.8% 8.0%

95Mo 23 2.0% 7.7% -0.4% -11.5% 17.6% -3.1% 5.0% -4.8% 14.5%

99Tc 16 25.5% 15.5% 23.3% -4.3% 49.5% 14.6% 38.7% 6.0% 45.3%

101Ru 14 5.5% 13.4% 4.2% -4.7% 48.8% -2.2% 7.4% -3.3% 9.5%

109Ag 15 31.0% 38.1% 20.2% -17.8% 147.2% 12.7% 46.4% -4.5% 54.5%

133Cs 16 -3.2% 7.2% -2.9% -24.0% 7.7% -5.0% 1.7% -11.9% 3.6%

147Sm 35 0.2% 8.2% 1.6% -17.0% 17.0% -4.8% 6.0% -10.8% 7.7%

149Sm 32 -6.6% 12.2% -6.7% -34.0% 20.2% -16.6% 1.6% -20.2% 5.6%

150Sm 34 2.6% 6.6% 3.3% -10.4% 15.8% -4.2% 7.5% -8.1% 9.6%

151Sm 35 -0.5% 11.9% -0.2% -18.2% 37.9% -10.0% 4.9% -12.7% 14.4%

152Sm 35 4.7% 6.5% 6.0% -8.5% 13.6% 0.7% 10.2% -7.1% 12.2%

151Eu 15 -9.2% 21.8% 3.2% -48.4% 11.7% -32.7% 7.7% -39.3% 9.3%

153Eu 25 6.3% 3.7% 6.0% -3.2% 14.0% 4.7% 9.2% 0.6% 10.3%

155Gd 25 13.9% 12.8% 10.0% -8.5% 6.0% 10.0% 19.9% 50.4% 4.0%

103Rh 15 5.2% 9.0% 2.5% -6.2% 31.1% 1.1% 5.0% -0.3% 20.2% a p10 and p90 are the 10% and 90% percentiles of the distribution of the C/M-1 values.
7-6

Quartiles (box)

Median (horizontal line in box)

10th and 90th percentiles (whiskers)

Mean (shown as a small square in the box)

Max/Min data points (×) and individual data points (+) for each sample

Max data points above 60% are not shown in the plots

Figure 14 Box Plot of the Major Actinide Isotopes
7-7

Quartiles (box)

Median (horizontal line in box)

10th and 90th percentiles (whiskers)

Mean (shown as a small square in the box)

Max/Min data points (×) and individual data points (+) for each sample

Max data points above 60% are not shown in the plots

Figure 15 Box Plot of the Minor Actinides and Fission Products (Mo, Tc, Ru, Ag, and

 Rh) 
7-8

Quartiles (box)

Median (horizontal line in box)

10th and 90th percentiles (whiskers)

Mean (shown as a small square in the box)

Max/Min data points (×) and individual data points (+) for each sample

Max data points above 60% are not shown in the plots

Figure 16 Box Plot of the Fission Products (Nd, Cs, Sm, Eu, and Gd) 
7-9

7.3 Criticality Calculations

The measured isotopic concentrations were applied in the GBC-68 application model, and the keff

values were calculated with KENO V.a using data from the 77 spent fuel sample measurements. 

As discussed previously, three separate criticality calculations were performed using the 

measured data for each sample:

1. Measured isotopic data plus calculated surrogate data for isotopes not measured in the 

sample, with surrogate data calculated based on the median isotopic bias (Table 19)

2. Measured isotopic data with surrogate data calculated based on the p10 percentile bias 

(Table 19)

3. Measured isotopic data with surrogate data calculated based on the p90 percentile bias

(Table 19)

For the application of measurement data to the criticality model, the measured isotopic 

concentrations in units of mg/gUi were converted to atom number densities using a fuel density in 

the application model of 10.42 g/cm3

.

The potential uncertainty introduced by using calculated surrogate data to augment missing 

measurement data is estimated by analyzing the uncertainties in the surrogate data based on the 

range of deviations observed in the samples with measurements. In this case, it is important to 

note that uncertainties are applied such that the calculated concentrations of all surrogate data 

are simultaneously increased (when the p90 percentile bias was applied) or simultaneously

decreased (when the p10 percentile bias was applied), providing a conservative estimate of the 

uncertainty range. In an earlier study using a similar approach [6], the impact of surrogate data 

uncertainty was estimated to be small; however, no quantitative analysis of the uncertainty was 

performed in that study.

The keff values obtained using measured nuclide concentrations are compared with the values 

obtained by calculation using the Polaris code for the actinide-only nuclide set (Table 1) and 

actinides plus fission products (Table 2). Table 20 lists the keff values and the difference in keff

between the measured and calculated actinide-only isotope set concentrations in absolute units of

pcm (1 pcm = 10-5

). Since keff is calculated by KENO V.a with a stochastic uncertainty of less than 

10 pcm, this Monte Carlo sampling uncertainty contribution can be neglected. The uncertainties 

for the ∆keff values in Table 20 correspond to the uncertainties associated with the use of 

surrogate data. The plus and minus error bars were not equal due to the non-normal distributions 

of the nuclide deviations used in developing the p10 and p90 percentile values. However, in most 

cases the error bars were similar and therefore only the average error bar is listed in Table 20.

Values of the sample burnup and the average void fraction are also listed in Table 20. Sample 

identifiers are based on the reactor, assembly, fuel rod, and sample name. The reactor names are 

• FOR (Forsmark), 

• DOD (Dodewaard), 

• FD2 (Fukushima Daini), 

• LBT (Leibstadt), and 

• LIM (Limerick).
7-10

Measured uranium and plutonium isotopes were available for all samples. The Forsmark 3 

sample F3F6 and samples from Fukushima Daini 1 assemblies 2F1ZN2 and F21ZN3 did not 

include measurements of major actinide 241Am. Therefore, the keff values for these samples were 

calculated using surrogate concentrations for 241Am and include bias and bias uncertainty 

estimates.

Table 21 provides keff results for the actinide-plus-fission product isotope dataset with the 

uncertainties associated with surrogate data for the minor actinides and fission products as 

estimated using the p10 and p90 percentiles from Table 19.

The keff bias results are plotted in Figure 17. Each sample is labeled, and the data points are 

colored to identify the different experimental datasets. In general, the actinide-only results are 

similar to the actinide-plus-fission product results.
7-11

Table 20 Criticality Results for the GBC-68 Cask Model for Major Actinide-Only

 Compositions

No. Sample ID Burnup

(GWd/MTU)

Void

(%)

Measured

kkeeeeee

mm

Calculated

kkeeeeee

cc

∆kkeeeeee

(pcm) a

1 FD2-2|2F2DN23|SF98|1 4.2 0.0 0.6456 0.6376 -803

2 FD2-2|2F2DN23|SF98|2 26.5 0.0 0.8401 0.8165 -2,358

3 FD2-2|2F2DN23|SF98|3 36.9 3.0 0.7418 0.7474 565

4 FD2-2|2F2DN23|SF98|4 42.4 11.0 0.7151 0.7194 438

5 FD2-2|2F2DN23|SF98|5 44.0 32.0 0.7314 0.7318 42

6 FD2-2|2F2DN23|SF98|6 39.9 54.5 0.7835 0.7787 -482

7 FD2-2|2F2DN23|SF98|7 39.4 68.0 0.7920 0.7986 666

8 FD2-2|2F2DN23|SF98|8 27.2 73.0 0.8452 0.8538 858

9 FD2-2|2F2DN23|SF99|1 7.5 0.0 0.6417 0.6238 -1,790

10 FD2-2|2F2DN23|SF99|2 22.6 1.4 0.8129 0.8210 808

11 FD2-2|2F2DN23|SF99|3 32.4 5.8 0.7543 0.7604 614

12 FD2-2|2F2DN23|SF99|4 35.4 10.8 0.7312 0.7463 1,510

13 FD2-2|2F2DN23|SF99|5 37.4 27.7 0.7510 0.7519 95

14 FD2-2|2F2DN23|SF99|6 32.4 54.7 0.8027 0.8062 355

15 FD2-2|2F2DN23|SF99|7 32.1 66.5 0.8125 0.8214 892

16 FD2-2|2F2DN23|SF99|8 21.8 71.7 0.8493 0.8671 1,776

17 FD2-2|2F2DN23|SF99|9 16.7 72.9 0.8697 0.8845 1,478

18 FD2-2|2F2D1|B3|TU103 10.0 64.0 0.9156 0.9116 -404

19 FD2-2|2F2D1|B3|TU104 9.4 60.2 0.8993 0.9150 1,564

20 FD2-2|2F2D1|B3|TU105 12.3 17.3 0.9030 0.8866 -1,647

21 FD2-2|2F2D1|F6|TU101 14.0 64.3 0.9663 0.9503 -1,606

22 FD2-2|2F2D1|F6|TU102 18.2 12.9 0.9238 0.9160 -779

23 FD2-2|2F2D1|F6|TU106 16.1 59.8 0.9462 0.9381 -808

24 FD2-2|2F2D2|B3|TU203 24.5 63.1 0.8530 0.8387 -1,426

25 FD2-2|2F2D2|B3|TU204 23.5 58.5 0.8385 0.8401 154

26 FD2-2|2F2D2|B3|TU205 22.8 10.4 0.7922 0.8159 2,367

27 FD2-2|2F2D2|F6|TU201 29.1 63.1 0.8935 0.8763 -1,717

28 FD2-2|2F2D2|F6|TU202 32.9 7.0 0.8294 0.8246 -483

29 FD2-2|2F2D3|A4|TU304 37.8 61.0 0.7434 0.7585 1,511

30 FD2-2|2F2D3|A4|TU306 32.3 6.0 0.7011 0.7350 3,393

31 FD2-2|2F2D3|B3|TU308 30.2 63.5 0.7963 0.8114 1,505

32 FD2-2|2F2D3|B3|TU309 34.8 60.5 0.7942 0.7868 -733

33 FD2-2|2F2D3|B3|TU311 33.5 9.1 0.7159 0.7463 3,033

34 FD2-2|2F2D3|H5|TU301 34.6 60.6 0.7490 0.7718 2,285

35 FD2-2|2F2D3|H5|TU302 31.4 5.2 0.7047 0.7406 3,593

36 FD2-2|2F2D8|A4|TU505 59.1 54.9 0.6746 0.6695 -503

37 FD2-2|2F2D8|A4|TU506 57.5 23.0 0.6171 0.6248 775
7-12

Table 21 Criticality Results for the GBC-68 Cask Model for Major Actinide-Only

 Compositions (Con’t)

No. Sample ID

Burnup

(GWd/MTU)

Void

(%)

Measured

kkeeeeee

mm

Calculated

kkeeeeee

cc

∆kkeeeeee

(pcm) a

38 FD2-2|2F2D8|B3|TU510 53.1 62.2 0.7222 0.7136 -855

39 FD2-2|2F2D8|B3|TU511 48.1 14.0 0.6479 0.6632 1,536

40 FD2-2|2F2D8|H5|TU501 53.2 63.2 0.7004 0.7000 -46

41 FD2-2|2F2D8|H5|TU502 58.9 58.0 0.6661 0.6758 975

42 FD2-2|2F2D8|H5|TU503 55.6 20.6 0.6049 0.6272 2,222

43 FD2-1|2F1ZN2|C2|GdB 35.6 18.4 0.7079 0.7162 734 ± 93

44 FD2-1|2F1ZN2|C2|GdT 27.9 73.7 0.7754 0.7876 890 ±244

45 FD2-1|2F1ZN2|C3|UB 47.5 18.4 0.7575 0.7623 348 ± 105

46 FD2-1|2F1ZN2|C3|UT 38.2 73.7 0.8268 0.8402 969 ±272

47 FD2-1|2F1ZN3|A9|UB 61.2 11.4 0.5674 0.5720 312 ± 109

48 FD2-1|2F1ZN3|A9|UM 64.2 38.3 0.5892 0.5731 -1,903 ± 211

49 FD2-1|2F1ZN3|A9|UT 56.3 60.1 0.6276 0.6321 262 ± 144

50 FD2-1|2F1ZN3|C2|GdB 54.6 18.4 0.6147 0.6203 402 ± 131

51 FD2-1|2F1ZN3|C2|GdM 54.5 38.3 0.6329 0.6269 -1,006 ± 293

52 FD2-1|2F1ZN3|C2|GdT 46.3 60.1 0.7120 0.7175 369 ± 136

53 FD2-1|2F1ZN3|C3|UB 68.3 11.4 0.6357 0.6407 329 ± 130

54 FD2-1|2F1ZN3|C3|UM 68.4 38.3 0.6607 0.6549 -876 ± 220

55 FD2-1|2F1ZN3|C3|UT 59.1 60.1 0.7411 0.7542 1,120 ± 134

56 DOD-1|Y013|B2|DU1 b 55.0 49.8 0.7050 0.7139 891

57 DOD-1|Y013|B2|DU1 c 55.0 49.8 0.6955 0.7033 773

58 FOR-3|14595|F6|F3F6 61.0 58.4 0.6577 0.6463 -1485 ± 274

59 FOR-3|GN592|J8|ENUSA-1 50.4 51.0 0.7257 0.7077 -1,801

60 FOR-3|GN592|J8|ENUSA-2 51.1 51.0 0.7269 0.7077 -1,920

61 FOR-3|GN592|J8|ENUSA-3 51.5 13.3 0.6842 0.6757 -849

62 FOR-3|GN592|J8|ENUSA-4 56.0 60.9 0.7096 0.6824 -2,722

63 FOR-3|GN592|J8|ENUSA-5 43.6 67.0 0.7531 0.7339 -1,916

64 FOR-3|GN592|J8|ENUSA-6 43.1 2.2 0.7058 0.6992 -667

65 FOR-3|GN592|J8|ENUSA-7 47.8 13.3 0.6822 0.6757 -644

66 FOR-3|GN592|J8|ENUSA-8 38.3 66.5 0.7665 0.7628 -371

67 LBT|AIA003|KLU1 60.5 8.4 0.6158 0.6261 1,032

68 LBT|AIA003|KLU2 62.9 51.2 0.6539 0.6649 1,104

69 LBT|AIA003|KLU3 56.5 70.4 0.6729 0.6870 1,407

70 LIM-1|FYJ1433|D8|3D2B 54.8 54.8 0.6972 0.7213 2,412

71 LIM-1|FYJ1433|D8|4G3 37.0 68.8 0.7548 0.7889 3,408

72 LIM-1|FYJ1433|D9|1D2 62.1 12.1 0.6237 0.6255 182

73 LIM-1|FYJ1433|D9|2D2 65.5 44.1 0.6554 0.6581 271

74 LIM-1|FYJ1433|D9|4D4 65.0 65.4 0.6748 0.6709 -391

75 LIM-1|FYJ1433|D9|4G1 56.5 69.1 0.7014 0.7075 607

76 LIM-1|FYJ1433|H5|3A1C 57.9 54.8 0.7175 0.7200 246

77 LIM-1|FYJ1433|H5|3A1G 57.8 57.7 0.7188 0.7200 118

a ∆kkeeeeee = kkeeeeee

cc − kkeeeeeemm ± average uncertainty due to use of surrogate nuclide compositions.

b sample measured by laboratories at SCK•CEN. c sample measured by laboratories at PSI.
7-13

Table 22 Criticality Results for the GBC-68 Cask Criticality Analysis Model for Actinide

 and Fission Product Spent Fuel Compositions

No. Sample ID

Burnup

(GWd/MTU)

Void

(%)

Measured

kkeeeeee

mm

Calculated

kkeeeeee

cc

∆kkeeeeee

(pcm) a

1 FD2-2|2F2DN23|SF98|1 4.2 0.0 0.6317 0.6238 -785 ± 27

2 FD2-2|2F2DN23|SF98|2 26.5 0.0 0.7881 0.7628 -2532 ± 138

3 FD2-2|2F2DN23|SF98|3 36.9 3.0 0.6744 0.6798 547 ± 176

4 FD2-2|2F2DN23|SF98|4 42.4 11.0 0.6425 0.6463 382 ± 168

5 FD2-2|2F2DN23|SF98|5 44.0 32.0 0.6569 0.6567 -21 ± 181

6 FD2-2|2F2DN23|SF98|6 39.9 54.5 0.7121 0.7071 -502 ± 180

7 FD2-2|2F2DN23|SF98|7 39.4 68.0 0.7212 0.7282 694 ± 172

8 FD2-2|2F2DN23|SF98|8 27.2 73.0 0.7886 0.7993 1070 ± 156

9 FD2-2|2F2DN23|SF99|1 7.5 0.0 0.6205 0.6035 -1697 ± 64

10 FD2-2|2F2DN23|SF99|2 22.6 1.4 0.7631 0.7707 757 ± 292

11 FD2-2|2F2DN23|SF99|3 32.4 5.8 0.6912 0.6970 581 ± 341

12 FD2-2|2F2DN23|SF99|4 35.4 10.8 0.6645 0.6792 1462 ± 349

13 FD2-2|2F2DN23|SF99|5 37.4 27.7 0.6762 0.6758 -38 ± 290

14 FD2-2|2F2DN23|SF99|6 32.4 54.7 0.7399 0.7421 223 ± 335

15 FD2-2|2F2DN23|SF99|7 32.1 66.5 0.7491 0.7574 827 ± 350

16 FD2-2|2F2DN23|SF99|8 21.8 71.7 0.7985 0.8169 1846 ± 315

17 FD2-2|2F2DN23|SF99|9 16.7 72.9 0.8275 0.8430 1546 ± 253

18 FD2-2|2F2D1|B3|TU103 10.0 64.0 0.8757 0.8695 -614 ± 345

19 FD2-2|2F2D1|B3|TU104 9.4 60.2 0.8528 0.8674 1455 ± 405

20 FD2-2|2F2D1|B3|TU105 12.3 17.3 0.8698 0.8510 -1880 ± 278

21 FD2-2|2F2D1|F6|TU101 14.0 64.3 0.9336 0.9153 -1824 ± 263

22 FD2-2|2F2D1|F6|TU102 18.2 12.9 0.8831 0.8742 -888 ± 293

23 FD2-2|2F2D1|F6|TU106 16.1 59.8 0.9095 0.8997 -979 ± 277

24 FD2-2|2F2D2|B3|TU203 24.5 63.1 0.8008 0.7836 -1719 ± 366

25 FD2-2|2F2D2|B3|TU204 23.5 58.5 0.7872 0.7864 -84 ± 370

26 FD2-2|2F2D2|B3|TU205 22.8 10.4 0.7403 0.7632 2286 ± 363

27 FD2-2|2F2D2|F6|TU201 29.1 63.1 0.8384 0.8183 -2010 ± 385

28 FD2-2|2F2D2|F6|TU202 32.9 7.0 0.7687 0.7611 -754 ± 407

29 FD2-2|2F2D3|A4|TU304 37.8 61.0 0.6762 0.6897 1350 ± 471

30 FD2-2|2F2D3|A4|TU306 32.3 6.0 0.6400 0.6735 3345 ± 408

31 FD2-2|2F2D3|B3|TU308 30.2 63.5 0.7353 0.7491 1378 ± 427

32 FD2-2|2F2D3|B3|TU309 34.8 60.5 0.7315 0.7190 -1257 ± 414

33 FD2-2|2F2D3|B3|TU311 33.5 9.1 0.6506 0.6801 2951 ± 442

34 FD2-2|2F2D3|H5|TU301 34.6 60.6 0.6860 0.7071 2111 ± 404

35 FD2-2|2F2D3|H5|TU302 31.4 5.2 0.6445 0.6801 3563 ± 425

36 FD2-2|2F2D8|A4|TU505 59.1 54.9 0.5937 0.5857 -798 ± 488

37 FD2-2|2F2D8|A4|TU506 57.5 23.0 0.5398 0.5448 503 ± 490
7-14

Table 23 Criticality Results for the GBC-68 Cask Criticality Analysis Model for Actinide

 and Fission Product Spent Fuel Compositions (Con’t)

No. Sample ID Burnup

(GWd/MTU)

Void

(%)

Measured

kkeeeeee

mm

Calculated

kkeeeeee

cc

∆kkeeeeee

(pcm) a

38 FD2-2|2F2D8|B3|TU510 53.1 62.2 0.6433 0.6311 -1217 ± 519

39 FD2-2|2F2D8|B3|TU511 48.1 14.0 0.5732 0.5863 1,,318 ± 473

40 FD2-2|2F2D8|H5|TU501 53.2 63.2 0.6206 0.6199 -69 ± 506

41 FD2-2|2F2D8|H5|TU502 58.9 58.0 0.5831 0.5921 900 ± 524

42 FD2-2|2F2D8|H5|TU503 55.6 20.6 0.5263 0.5484 2,212 ± 472

43 FD2-1|2F1ZN2|C2|GdB 35.6 18.4 0.6427 0.6489 622 ± 548

44 FD2-1|2F1ZN2|C2|GdT 27.9 73.7 0.7164 0.7254 04 ± 265

45 FD2-1|2F1ZN2|C3|UB 47.5 18.4 0.6804 0.6827 230 ± 547

46 FD2-1|2F1ZN2|C3|UT 38.2 73.7 0.7625 0.7699 742 ± 279

47 FD2-1|2F1ZN3|A9|UB 61.2 11.4 0.4936 0.4968 311 ± 659

48 FD2-1|2F1ZN3|A9|UM 64.2 38.3 0.5102 0.4917 -1856± 326

49 FD2-1|2F1ZN3|A9|UT 56.3 60.1 0.5518 0.5536 176 ± 706

50 FD2-1|2F1ZN3|C2|GdB 54.6 18.4 0.5379 0.5408 289 ± 647

51 FD2-1|2F1ZN3|C2|GdM 54.5 38.3 0.5552 0.5447 -1,056± 346

52 FD2-1|2F1ZN3|C2|GdT 46.3 60.1 0.6384 0.6396 119 ± 638

53 FD2-1|2F1ZN3|C3|UB 68.3 11.4 0.5476 0.5490 143 ± 672

54 FD2-1|2F1ZN3|C3|UM 68.4 38.3 0.5702 0.5586 -1,156 ± 315

55 FD2-1|2F1ZN3|C3|UT 59.1 60.1 0.6556 0.6654 985 ± 676

56 DOD-1|Y013|B2|DU1 b 55.0 49.8 0.6166 0.6258 925

57 DOD-1|Y013|B2|DU1 c 55.0 49.8 0.6108 0.6142 344

58 FOR-3|14595|F6|F3F6 61.0 58.4 0.5774 0.5605 -1688 ± 519

59 FOR-3|GN592|J8|ENUSA-1 50.4 51.0 0.6467 0.6266 -2,011 ± 473

60 FOR-3|GN592|J8|ENUSA-2 51.1 51.0 0.6462 0.6266 -1,964 ± 506

61 FOR-3|GN592|J8|ENUSA-3 51.5 13.3 0.6075 0.5964 -1,111 ± 524

62 FOR-3|GN592|J8|ENUSA-4 56.0 60.9 0.6301 0.6007 -2,935 ± 472

63 FOR-3|GN592|J8|ENUSA-5 43.6 67.0 0.6795 0.6589 -2,057 ± 548

64 FOR-3|GN592|J8|ENUSA-6 43.1 2.2 0.6331 0.6245 -861 ± 265

65 FOR-3|GN592|J8|ENUSA-7 47.8 13.3 0.6067 0.5964 -1,033 ± 547

66 FOR-3|GN592|J8|ENUSA-8 38.3 66.5 0.6979 0.6933 -464 ± 279

67 LBT|AIA003|KLU1 60.5 8.4 0.5316 0.5408 918 ± 659

68 LBT|AIA003|KLU2 62.9 51.2 0.5672 0.5749 766 ± 326

69 LBT|AIA003|KLU3 56.5 70.4 0.5880 0.5999 1,193 ± 706

70 LIM-1|FYJ1433|D8|3D2B 54.8 54.8 0.6121 0.6347 2,259 ± 647

71 LIM-1|FYJ1433|D8|4G3 37.0 68.8 0.6817 0.7184 3,672 ± 346

72 LIM-1|FYJ1433|D9|1D2 62.1 12.1 0.5395 0.5380 -153 ± 638

73 LIM-1|FYJ1433|D9|2D2 65.5 44.1 0.5671 0.5664 -71 ± 672

74 LIM-1|FYJ1433|D9|4D4 65.0 65.4 0.5862 0.5792 -699 ± 315

75 LIM-1|FYJ1433|D9|4G1 56.5 69.1 0.6164 0.6203 391 ± 676

76 LIM-1|FYJ1433|H5|3A1C 57.9 54.8 0.6331 0.6325 -58 ± 519

77 LIM-1|FYJ1433|H5|3A1G 57.8 57.7 0.6345 0.6325 -203 ± 473

a ∆kkeeeeee = kkeeeeee

cc − kkeeeeeemm ± average uncertainty due to use of surrogate nuclide compositions.

b sample measured by laboratories at SCK•CEN. c sample measured by laboratories at PSI.
7-15

Figure 17 keff Bias for Actinide-Only and Actinide-Plus-Fission Products Results
7-16

7.4 Experimental Correlations

Correlations are inherent in both the measurements and calculations. Correlations in spent fuel 

isotopic measurements by laboratory may occur due to biases associated with using a common 

measurement technique, instrument, or reference standard to calibrate the measurements. 

Similarly, correlations in the simulations can occur for samples obtained from the same fuel rod, 

assembly, or reactor due to common variations in the design data or estimated operating 

parameters and void fraction. Also, any error in the sample burnup will impact biases in the 

calculated nuclide compositions for the sample in a highly correlated way.

The approach used in this study inherently accounts for isotope correlations within the same 

sample since the measured and calculated nuclide concentrations are used directly in the 

criticality model without the need to address the bias and uncertainty of each individual nuclide. 

However, this approach also assumes that each sample is independent.

Sobes et al. [64] implemented modifications to the USLSTATS statistical analysis code, used to 

predict the upper subcritical limit (USL) for critical experiments, to account for potential 

correlations in critical experiments. However, the degree of correlation for samples within the 

same rod or samples in the same assembly is not known, and for this analysis, all samples are 

assumed to be independent and uncorrelated.

7.5 Isotopic Model Uncertainties

7.5.1 Measurements

Measurement uncertainties are reported by the laboratories for all isotopic concentration values 

used in this study. The uncertainty depends to a large extent on the measurement method, the 

type of instrument used for mass spectrometry, the use and accuracy of reference standards, and 

the isotopic concentration of the isotope in the fuel. The reported uncertainties can vary 

significantly by laboratory and depend on the uncertainty analysis methods and rigor, reliance on 

past experience, and which components of the measurement uncertainty have been included in 

the process (i.e., reproducibility of a measurement or analysis of all steps starting with cutting and 

dissolution of fuel samples). Due to the inconsistency of current uncertainty estimates, 

measurement uncertainties were not used to weight the individual sample results in this study. 

One sample from the Dodewaard reactor, DU1, was measured at independent laboratories at

SCK•CEN and PSI. A comparison of the keff results using these two measurement sets shows a 

difference of about 550 pcm attributed to the measurements alone. 

7.5.2 Void

Void fraction information is calculated by the operator with time steps shorter than the cycle 

length. The uncertainty in the void fraction has been estimated by comparison of calculated to 

measured average void fractions. Measurements analyzed by Morooka [65] suggest a relative 

standard deviation of 5.3% and 6.3% for the predictive codes COBRA/BWR and THERMIT. 

These values apply to average void fraction within an axial segment of an assembly (node).

The void distribution within the assembly flow channel is not uniform, and the uncertainty in local 

void in the vicinity of any single fuel rod can be much larger than the uncertainty in the average 

node void level. Studies suggest that the void fraction distribution in regions near the channel or 
7-17

corner and water rods [66] can be 25% less than the average void fraction for some conditions. 

However, this radial variability can depend significantly on the axial location within the assembly.

The impact of the void fraction during depletion on the kinf of the fuel in out-of-reactor conditions 

was previously evaluated by Wagner [67]. For core average void fractions of typically 40%, a 10% 

uncertainty in the void fraction was shown to have a corresponding uncertainty in kinf of about 

±300 pcm. Further studies performed in this work investigated void fraction uncertainties for fuel 

rod C3 of Fukushima Daini 1 assembly 2F1ZN3. Three samples, UB (bottom), UM (middle), and 

UT (top) were irradiated with void fractions of nominally 10%, 40%, and 70%. Reanalysis of these 

samples using a ±10% change on void fraction uncertainty resulted in a keff uncertainty up to ±30, 

±600, and ±300 pcm at the respective axial locations, with the largest sensitivity to void 

uncertainty observed for the middle sample (40% void).

The variability in the keff results (Figure 17) is larger than that expected on the basis of a 5–6% 

uncertainty in the void fraction alone, as determined from previous studies for node average 

values. Large deviations observed for some samples may be attributed to larger void uncertainties 

in local void fraction values that have been estimated to be as large as 25%. In such a case, 

errors in the calculated keff of up to 1,500 pcm would be expected. Two of the largest deviations 

are observed for Limerick 1 samples from rod D8. This rod contained Gd2O3 and had a lower 

average power and therefore may have experienced a lower void fraction compared to other rods 

in the assembly. The calculations significantly overestimate the keff for these samples, consistent 

with an overestimate of the local void conditions for this rod.

7.5.3 Fuel Temperature

The fuel temperature is generally reported with the operating history data as obtained by core 

code calculations. The uncertainty in these values has been estimated to be ±50°C when data are 

provided by the operator and ±100°C when values are estimated from other sources of 

information [68]. An analysis of Forsmark 3 GE14 assembly GN592 samples was performed by 

increasing the average fuel temperature from 792 K to 950 K during the depletion analysis. The 

impact for all axial sample positions was nominally 2 pcm/°C in the application model. Therefore, 

even assuming large uncertainties of 100°C in the fuel temperature, the uncertainty in keff is no 

greater than 200 pcm. This indicates that while fuel temperature is important, the impact is likely to 

be less than that due to many other sources of uncertainty present in BWR depletion models.

7.5.4 Sample Burnup 

The reported burnup of each sample is derived from measurements of 148Nd [69], 148Nd plus other 

fission products [70], and also using uranium and plutonium in limited cases [52]. Consequently,

the burnup is not known precisely due to uncertainties in the isotopic measurements, the 

derivation methods and nuclear data in the burnup derivation. Uncertainty in the burnup, an input 

parameter in the depletion calculations, can affect the nuclide concentrations and the keff of the 

application model.

The impact of burnup uncertainty on keff was estimated using a sample UM (mid-axial height) from 

rod C3 of Fukushima Daini 1 assembly 2F1ZN3 with an average void of 38%. The uncertainty 

was evaluated at the end of each cycle of irradiation, for five cycles, to cover a range of sample 

burnups. An uncertainty in fuel burnup of nominally 2% was found to have a 200 pcm effect at low 

burnup, and up to 600 pcm at high burnup, for the GBC-68 application model. 
7-18

In the case of the Fukushima Daini 2 data, where uncertainties in the sample burnup values of up 

to 6.5% were reported, the potential impact of keff can be as large as 2,000 pcm. These samples 

(Figure 17) exhibit some of the larger variations in the analyzed data.

7.5.5 Summary of Model Uncertainties

Uncertainties in both the measurements and the calculations contribute to the total uncertainty in 

the criticality model. Calculation uncertainties were assessed in this study for the input parameters 

of void fraction, fuel temperature, and burnup of the measured sample using models for several 

samples. The impact of the uncertainty parameters on the calculated keff of the cask model are 

summarized in Table 22. The different parameter values reflect the typical uncertainty and the 

maximum uncertainty values. The range of keff uncertainty values for some parameters reflects 

different sample burnup and void values. These uncertainties, when combined (assuming they are 

independent) result in a total uncertainty in the application model from about 600 to 1,900 pcm.

The measurements represent a large source of the overall uncertainty both in terms of the nuclide 

concentration values and the estimation of the sample burnup that is also derived from the 

measurements.

Table 24 Summary of Uncertainties

Parameter Parameter uncertainty keff uncertainty (pcm)

RCA measurements 1%–5% 550

Fuel temperature 50 K

100 K

100

200

Void fraction 6 %

25 %

10–90

50–375

Sample burnup

2 %

6 %

200–600

600–1,800

Combined 600–1,900

7.6 Trending Analysis

The keff results presented in Section 7.3 were further analyzed by trending on parameters that 

have the largest impact on the isotopic concentrations. This analysis included trending on sample 

burnup and moderator void fraction. Trending was also performed on the keff of the application 

model. This parameter includes a measure of both the fuel burnup and initial enrichment that is 

not reflected by trending on fuel burnup alone.

Trending analysis was performed using linear regression analysis of the Δkeff results in Tables 20 

and 21. Sample uncertainties associated with the use of surrogate data described in Section 7.2 

were analyzed to illustrate the impact of surrogate data and provide a potential means of 

weighting different samples using these uncertainties. However, other uncertainties associated 

with each sample including the measurement uncertainties and the sample burnup uncertainties 

(usually based on the measured 148Nd) are not included in the error bars. Because there is usually 

insufficient information reported on the how measurement uncertainties are estimated by different 

laboratories, and inconsistent approaches are frequently applied for uncertainty analysis, no 

weighting was applied to avoid potential incorrect weighting of the samples and biasing the 

results.
7-19

A statistical analysis of the trending data was performed to estimate the lower one-sided 95% 

tolerance limit (LTL) of the population at the 95% confidence level. An analysis of the residuals of 

the linear regression fit was performed using the Shapiro-Wilk test to verify normality of the 

distributions. The data were determined at the 0.05 level to be significantly drawn from a normally 

distributed population for all trending parameters.

7.6.1 Trends with Fuel Burnup

The keff bias associated with the isotopic predictions as a function of fuel sample burnup is shown 

in Figure 18 for the case of major actinides and actinides plus fission products, with the linear 

regression fit and the 95% LTL shown in the graph. The LTL uses the same statistical methods as 

are used in the calculation of the Upper Subcritical Limit (USL) for critical systems for which no 

statistically significant trend is present. The regression analysis indicates that the slope of the 

trend is not statistically significant, and the slope was therefore set to zero for the fit.

The mean bias corresponds to the keff bias term, βi, in Eq. (1). The LTL can be used to estimate 

the keff bias uncertainty, ∆ki, associated with the depletion code. In this report, the reported total 

margin for uncertainty associated with the depletion code includes both the bias and the bias 

uncertainty terms (βi + ∆ki). Positive bias is not credited, consistent with common criticality safety 

practice. In the case that positive biases were to be credited, any positive bias would be 

subtracted from the reported total uncertainty values.

For the actinide-only calculations, there is a positive bias, βi, of 253 pcm. The LTL band is 

relatively constant, ranging from a maximum value of -2,150 to -2,074 pcm. For actinide plus 

fission products, the mean bias is 161 pcm, with an LTL band of -2,368 to -2,288 pcm over the 

range of all fuel burnups. The maximum total margin for uncertainty, not crediting the positive 

bias, is 2,403 pcm for major actinide only burnup credit, and 2,529 pcm for actinide plus fission 

product credit.
7-20

Figure 18 keff Bias for Actinide-Only (top) and Actinide-Plus-Fission Products (bottom)

 Results as a Function of fuel Burnup Showing Mean (red) and 95% LTL Band

 (blue)
7-21

7.6.2 Trends with Moderator Void

The keff bias as a function of average moderator void fraction is shown in Figure 19 for the case of 

major actinides and actinides plus fission products, with the linear regression fit and the 95% LTL 

shown in the graph. Again, the slope of the linear regression fit is not statistically significant and is 

set to zero.

The major actinide-only results show a positive bias of 253 pcm and an LTL band range from -

2,114 to -2,074 pcm. For actinides plus fission products, the mean bias is 161 pcm with an LTL 

band from -2,330 to -2,288 pcm. The maximum total margin for uncertainty, not crediting the 

positive bias, is 2,367 pcm for major actinide only burnup credit, and 2,491 pcm for actinide plus 

fission product credit.

7.6.3 Trends with Fuel Reactivity

Trending with the keff bias was also investigated as a function of the keff of the GBC-68 application 

model to assess both the burnup and the initial enrichment of the samples. The keff bias with 

reactivity is shown in Figure 20 for the case of major actinides and actinides plus fission products.

Again, the slope of the linear regression fit is not statistically significant and is set to zero.

The linear regression fit for the major actinide results shows a positive bias of 253 pcm and LTL

band range from -2,166 to -2,074 pcm. For actinides plus fission products, the mean bias is 

161 pcm with an LTL band from -2,288 to -2,390 pcm. The maximum total margin for uncertainty, 

not crediting the positive bias, is 2,419 pcm for major actinide only burnup credit, and 2,551 pcm 

for actinide plus fission product credit.
7-22

Figure 19 keff Bias for Actinide-Only (bottom) and Actinide-Plus-Fission Products (top)

 Results as a Function of Moderator Void Showing Mean (red) and 95% LTL

 Band (blue)
7-23

Figure 20 keff Bias for Actinide-Only (bottom) and Actinide-Plus-Fission Products (top)

 Results as a Function of Application keff Showing Mean (red) and 95% LTL

 Band (blue)
7-24

7.7 Margins for Isotopic Uncertainty

Margins for uncertainty in keff calculations due to the predicted nuclide concentrations are 

developed from the trending analysis and the lower one-sided 95% LTL for the distribution. 

Analysis of the trending parameters shows that the slope of the mean bias is not statistically 

significant. Therefore, a constant bias is appropriate. The average bias, the maximum 95% LTL 

values, and the total margins for uncertainty are listed in Table 22. The bias and minimum LTL 

values are seen to be the same for all trending parameters because there is no significant slope in 

the fit of the data. The maximum LTL band value is only weakly dependent on the trending 

parameter. The LTL values do not include the bias component. Consistent with past criticality 

safety practice, positive bias is added to the LTL value and negative bias is set to zero.

The KENO V.a criticality calculations in this study were performed using axially and radially 

uniform fuel compositions. In practice, burnup credit criticality analyses are usually performed 

using models with axially varying burnup and fuel compositions. In these cases, the region with 

the peak fission densities and therefore the region dominant in the keff calculation is near the top 

end of the BWR fuel assembly. This region has a higher fissile material content a greater reactivity 

than the assembly average due to the lower burnup and higher void fractions during irradiation. To 

ensure an appropriate margin for isotopic uncertainty, the margin for uncertainty corresponding to 

the maximum over the range of fuel parameters for the assembly should be used. The LTL is 

observed to be only weakly correlated to the trending parameters, and the margin for uncertainty 

is relatively constant over the parameter ranges. 

Table 25 Summary of Bias and Uncertainties for Trending Parameters

Isotope set Trending parameter

Mean bias 

(pcm)

LTL min. 

(pcm)

Total margin for 

uncertainty (pcm)

Major actinides 

only Fuel burnup 253 -2,150 2,403

Void fraction 253 -2,114 2,367

System keff 253 -2,166 2,419

Actinides plus 

fission products Fuel burnup 161 -2,368 2,529

Void fraction 161 -2,330 2,491

System keff 161 -2,390 2,551
8-1

8 SUMMARY AND CONCLUSIONS

Experimental data from the destructive assay of 77 BWR spent fuel samples have been evaluated 

in this report to calculate margins for uncertainty in the predictions of isotopic inventories as 

applied to burnup credit criticality calculations. Measurements cover a wide range of modern 

assemblies including 8 × 8-2, 8 × 8-4, 9 × 9-7, GE11 9 × 9, 9 × 9-9, GE14 10 × 10, SVEA-96, and 

SVEA-100 10 × 10 fuel assembly designs. The data cover a range of void conditions up to 74% 

and a burnup range from 7 to 68 GWd/MTU. Most of the measurement data used in this report 

are obtained from public references and additional information compiled and documented as part 

of the OECD/NEA SFCOMPO spent nuclear fuel measurement database. Several datasets used 

in this study are from proprietary programs. These data may be made available to support 

licensing applications through nondisclosure agreements. 

The uncertainty analysis methodology used in this study is independent of the application model 

or the computational methods. The methodology is based on the validation of separate 

components of the criticality calculation. The margins for isotopic uncertainty are developed by the 

direct application of measurement and calculated nuclide concentrations to the application model 

and statistical analysis of the results. This procedure, as applied to the major actinide-only 

calculations, requires only minimal analysis of the isotopic distributions of individual nuclides since 

most samples include measurements for all major actinide isotopes. For minor actinide and fission 

product credit, analysis of individual isotopic bias and uncertainty was used to develop surrogate 

isotope concentration data with uncertainties for isotopes not measured in a fuel sample. 

This uncertainty analysis approach is demonstrated in this report using SCALE 6.2.2. with 

ENDF/B-VII.1 cross section data. Specifically, depletion calculations were performed using the 

Polaris code and criticality calculations were performed using KENO V.a Margins for uncertainty 

associated with the predicted nuclide compositions are developed for a GBC-68 dry storage cask 

model. Therefore, the results presented in this report are specific to this code system and 

application model but are expected to be similar for other dry storage and transportation cask 

designs, when using same computer codes and cross section data. 

The use of measured isotopic concentrations to develop margins for uncertainty is expected to be 

conservative. Significant uncertainties in the measured isotopic concentrations can be introduced 

due to the complexity of the radiochemical analysis procedures. Biases of more than 500 pcm 

were observed using measurements performed by two different laboratories for the same sample. 

Measurements are also used to estimate the sample burnup, used in the depletion calculations for 

the sample, adding to uncertainty. Moreover, the input data associated with the operating history 

obtained for core operating and fuel management calculations can also have large uncertainties. 

In particular, the void conditions for an assembly are obtained from core thermal-hydraulic 

calculations for the axial node of an assembly, with uncertainties of 5–6% for average void 

conditions and larger uncertainties for local void conditions within the assembly. These 

uncertainties in the input data contribute to uncertainty in keff of more than 1,000 pcm. These 

uncertainties are not associated with the accuracy of the code calculations themselves but 

nevertheless contribute to the estimated margins for code uncertainty since these sources of error 

are not easily separated from other sources.

A margin for uncertainty is developed from the bias and 95% one-sided lower tolerance limit for 

the population of measurement data, and trending analysis was performed for sample burnup, 

average void, and keff of the application model. For the Polaris calculations and the GBC-68 

application model used in the present study, the keff bias for actinide-only calculations is 253 pcm 

with a conservative margin for isotopic uncertainty of 2,419 pcm for the range of fuel samples 
8-2

included in the current analysis. This margin is observed to be largely independent of fuel burnup 

or void fraction based on the analysis of available experimental data. For actinide-plus-fission 

product calculations, the keff bias is 161 pcm and the most conservative margin for isotopic 

uncertainty is 2,551 pcm. These values do not credit the positive bias observed for all calculations 

performed in this study.

These results are similar to those shown in previous studies for PWR burnup credit [6], where 

average biases for a dry storage cask of 320 to 720 pcm and uncertainties of 1,430 to 2,050 pcm 

were reported for fuel with burnup less than 50 GWd/MTU. The larger uncertainties seen in the 

present study are likely attributed to the increased complexity of the BWR fuel and reactor 

modeling.
9-1

9 REFERENCES

[1] Interim Staff Guidance 8, Revision 3, Burnup Credit in the Criticality Safety Analyses of 

PWR Spent Fuel in Transportation and Storage Casks. US Nuclear Regulatory 

Commission, September 2012. < https://www.nrc.gov/docs/ML1226/ML12261A186.pdf >

[2] J. Hu, J. L. Peterson, I. C. Gauld, and S. M. Bowman, U.S. Commercial Spent Nuclear 

Fuel Assembly Characteristics – 1968–2013, US Nuclear Regulatory Commission report 

NUREG/CR-7227, September 2016. < https://doi.org/10.2172/1330516 >

[3] B. J. Marshall et al., Validation of keff Calculations for Extended BWR Burnup Credit, 

NUREG/CR-7252, prepared for the US Nuclear Regulatory Commission by Oak Ridge 

National Laboratory, Oak Ridge, Tennessee, 2018.

[4] O. W. Hermann and M. D. DeHart, Validation of Scale (SAS2H) Isotopic Predictions for 

BWR Spent Fuel, Oak Ridge National Laboratory report ORNL/TM-13315, September 

1998. < https://doi.org/10.2172/814258 >

[5] U. Mertyurek, M. W. Francis, and I. C. Gauld, SCALE 5 Analysis of BWR Spent Nuclear 

Fuel Isotopic Compositions for Safety Studies, Oak Ridge National Laboratory report 

ORNL/TM-2010/286, December 2010. < https://doi.org/10.2172/1081684 > 

[6] G. Radulescu, I. C. Gauld, G. Ilas, and J. C. Wagner, An Approach for Validating Actinide 

and Fission Product Burnup Credit Criticality Safety Analyses—Isotopic Composition 

Predictions, NUREG/CR-7108, prepared for the US Nuclear Regulatory Commission by 

Oak Ridge National Laboratory, Oak Ridge, Tennessee, 2012.

[7] L. Wimmer, Isotopic Generation and Confirmation of the BWR Application Model, AREVA 

Framatome ANP, Document 32-5035847-01 Yucca Mountain Site Characterization Office, 

2004, DOC 20040630.0007. 

[8] Y. Nakahara, K. Suyama, J. Inagawa, R. Nagaishi, S. Kurosawa, N. Kohno, M. Onuki, and 

H. Mochizuki, “Nuclide Composition Benchmark Data Set for Verifying Burnup Codes on 

Spent Light Water Reactor Fuels,” Nuclear Technology 137, pp. 111–126 (2002). 

<https://doi.org/10.13182/NT02-2 >

[9] OECD/NEA Working Party on Nuclear Criticality Safety, Expert Group on Assay Data for 

Spent Nuclear Fuel. < https://www.oecd-nea.org/science/wpncs/ADSNF/index.html >

[10] F. Michel-Sendis, I. Gauld, J. S. Martinez, et al., “SFCOMPO-2.0: An OECD/NEA 

Database of Spent Nuclear Fuel Isotopic Assays, Reactor Design Specifications, and 

Operating Data,” Annals of Nuclear Energy 110, pp. 779–788 (2017). 

<https://doi.org/10.1016/j.anucene.2017.07.022>

[11] J. Conde, C. Alejano, and J. M. Rey, “Nuclear Fuel Research Activities of the Consejo De 

Seguridad Nuclear,” Transactions of the 2006 International Meeting on LWR Fuel 

Performance, TopFuel 2006, Salamanca, Spain, 22–26 October 2006.

[12] D. Boulanger, M. Lippens, L. Mertens, J. Basselier, and B. Lance, “High Burnup PWR and 

BWR MOX Fuel Performance: A Review of Belgonucleaire Recent Experimental 

Programs,” Proceedings of the 2004 International Topical meeting on LWR Fuel 

Performance, September 19–22, 2004, American Nuclear Society, 2004.

[13] H. R. Radulescu, Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H 

Calculations, CAL-DSU-NU-000002 Rev 00A, Office of Civilian Radioactive Waste 

Management, August 2003.
9-2

[14] M. A. Jessee, W. A. Wieselquist, et al., “Polaris: A New Two-Dimensional Lattice Physics

Analysis Capability for the SCALE Code System,” Proceedings of the International

Conference on Physics of Reactors, PHYSOR 2014, Kyoto, Japan, 2014.

[15] B. T. Rearden and M. A. Jessee, Eds., SCALE Code System, ORNL/TM-2005/39, Version

6.2.2, Oak Ridge National Laboratory, Oak Ridge, Tennessee, 2017. Available from

Radiation Safety Information Computational Center as CCC-834.

[16] M. B. Chadwick, M. Herman, and P. Obložinský et al., “ENDF/B-VII.1 Nuclear Data for

Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay

Data,” Nuclear Data Sheets 112:12, pp. 2887–2996 (2011).

< https://doi.org/10.1016/j.nds.2011.11.002>

[17] Code of Federal Regulations, Title 10, “Energy,” 2011.

[18] Burnup Credit for LWR Fuel, ANSI/ANS-8.27-2015, American Nuclear Society, La Grange

Park, IL, 2015.

[19] A. H. Wells, Isotopic Model for Commercial SNF Burnup Credit, US Department of Energy,

Office of Civilian Radioactive Waste Management report CAL-DSU-NU-000007 REV 00B,

November 2004. < https://doi.org/10.2172/862151 >

[20] I. C. Gauld, Strategies for Application of Isotopic Uncertainties in Burnup Credit,

NUREG/CR-6811, prepared for the US Nuclear Regulatory Commission by Oak Ridge

National Laboratory, Oak Ridge, Tennessee, 2003.

[21] C. V. Parks, M. D. DeHart, and J. C. Wagner, Review and Prioritization of Technical

Issues Related to Burnup Credit for LWR Fuel, NUREG/CR-6665, prepared for the US

Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge,

Tennessee, 2000.

[22] J. Rhodes, K. Smith, and D. Lee, CASMO-5/CASMO-5M, A Fuel Assembly Burnup

Program, User’s Manual, SSP-07/431 Rev. 0, Studsvik Scandpower, Inc., 2007.

[23] M. Ouisloumen et al., “PARAGON: The New Westinghouse Assembly Lattice Code,” in

Proceedings of the ANS International Meeting on Mathematical Methods for Nuclear

Applications, Salt Lake City, Utah, USA, 2001.

[24] MCNP—Monte Carlo N-Particle Transport Code System, Version 5, Vol. 1: LA-UR-03-

1987, Vol. 2: LA-CP-03-0245, and Vol. 3: LA-CP-03-0284, Los Alamos National

Laboratory, 2003. Available from Radiation Safety Information Computational Center at

Oak Ridge National Laboratory as CCC-740.

[25] M. L. Williams and K.-S. Kim, “The Embedded Self-Shielding Method,” in PHYSOR 2012:

Advances in Reactor Physics, Knoxville, TN, 2012.

[26] U. Mertyurek, B. R. Betzler, M. A. Jessee, and S. M. Bowman, “SCALE 6.2 Lattice Physics

Code Accuracy Assessment for Light Water Reactor Fuel,” Proceedings of the

International Conference on Physics of Reactors, PHYSOR 2018, Cancun, Mexico, 2018.

[27] J-Ch. Sublet, A. J. Koning, R. A. Forrest, and J. Kopecky, “The JEFF-3.0/A Neutron

Activation File—EAF-2003 into ENDF-6 Format,” JEFDOC-982, Commissariat à l’Energie

Atomique, France, November 2003.

[28] T. R. England and B. F. Rider, Evaluation and Compilation of Fission Product Yields 1993,

LA-UR-94-3106 (ENDF-349), Los Alamos National Laboratory, 1994.
9-3

[29] M. T. Pigni, M. W. Francis, and I. C. Gauld, “Investigation of Inconsistent ENDF/B-VII.1

Independent and Cumulative Fission Product Yields with Proposed Revisions,” Nuclear

Data Sheets 123, pp. 231–236 (2015).

[30] R. J. Guenther, D.E. Blahnik, T. K. Campbell, U. P. Jenquin, J. E. Mendel, L. E. Thomas,

and C.K. Thornhill, Characterization of Spent Fuel Approved Testing Material-ATM-105,

PNL-5109-105/UC-802, December 1991. < https://doi.org/10.2172/138325 >

[31] P. Barbero et al., “Post Irradiation Analysis of the Gundremmingen BWR Spent Fuel,”

EUR 6301en, Joint Research Center Ispra and Karlsruhe Establishments, 1978.

[32] H. Natsume et al., “Gamma-Ray Spectrometry and Chemical Analysis Data of JPDR-I

Spent Fuel,” Journal of Nuclear Science and Technology 14(10), pp. 745–761 (1977).

[33] T. Suzaki et al., “Non-Destructive and Destructive Measurements on Burnup

Characteristics of Japan Power Demonstration Reactor-I Full-Core Fuel Assemblies,”

Journal of Nuclear Science and Technology 23(1), pp. 53–72 (1986).

[34] R. T. Primm III, ARIANE International Programme Final Report, Oak Ridge National

Laboratory report ORNL/SUB/97-XSV750-1, April 2002.

[35] P. Ortego and A. Rodríguez, Evaluation of Dodewaard DU1 Sample, report prepared for

the OECD/NEA Expert Group on Assay Data for Spent Nuclear Fuel.

[36] G. Rowlands, “Resonance absorption and non-uniform temperature distributions,” Journal

of Nuclear Energy Parts A/B Reactor Science and Technology 16, pp. 235–236 (1964).

[37] D. J. Kelly, “Depletion of a BWR lattice using the RACER continuous energy Monte Carlo

code,” Proc. Int. Natl. Conf. on Mathematics and Computations, Reactor Physics, and

Environmental Analyses, Portland, Vol. II, p. 1011, 1995.

[38] H.-U. Zwicky, Isotopic Data of Sample F3F6 from a Rod Irradiated in the Swedish Boling

Water Reactor FORSMARK 3, ZC-08/001, September 2008. < https://www.oecdnea.org/science/wpncs/ADSNF/reports/Forsmark3/F3F6.pdf>

[39] F. Lindström, Forsmark 3 – Characteristic Data of Fuel Rod J8 in Bundle GN592, VNF1002152744/02, Vattenfall Nuclear Fuel AB report, 2011.

[40] H.-U. Zwicky, J. Low, and M. Granfors, Fuel Pellet Isotopic Analyses of Samples from a

GE14 Fuel Rod Irradiated in Forsmark 3— Final Report, STUDSVIK/N-10/007, Studsvik

Nuclear AB report, 2010.

[41] D. E. Mueller, J. M. Scaglione, J. C. Wagner, and S. M. Bowman, Computational

Benchmark for Estimated Reactivity Margin from Fission Products and Minor Actinides in

BWR Burnup Credit, NUREG/CR-7157, prepared for the US Nuclear Regulatory

Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn., February 2013).

[42] T. Yamamoto and Y. Kanayama, “Lattice Physics Analysis of Burnups and Isotope

Inventories of U, Pu, and Nd of Irradiated BWR 9×9-9 UO2 Fuel Assemblies,” Journal of

Nuclear Science and Technology, 45:6, pp. 547–566 (2008).

<https://doi.org/10.1080/18811248.2008.9711879 >

[43] M. Suzuki, T. Yamamoto, H. Fukaya, K. Suyama, and G. Uchiyama, “Lattice physics

analysis of measured isotopic compositions of irradiated BWR 9 × 9 UO2 fuel,” Journal of

Nuclear Science and Technology, 50:12, pp. 1161–1176 (2013), DOI:

10.1080/00223131.2013.837845.

[44] T. Yamamoto, Irradiation Report of Fuel Samples of BWR 9 × 9-9 Fuel, report prepared for

the OECD NEA Data Bank contribution to SFCOMPO, February 2014.
9-4

[45] International Atomic Energy Agency, In-core fuel management code package validation for

BWRs, IAEA-TECDOC-849, 1995.

[46] Y. Nakahara, K. Suyama, and T. Suzaki, Technical Development on Burn-Up Credit for

Spent LWR Fuels, JAERI-Tech 2000-071, Japan Atomic Energy Research Institute, 2000

(in Japanese). English translation published as report ORNL/TR-2001/01, 2002).

[47] T. Yamamoto and M. Yamamoto, “Nuclear Analysis of PIE Data of Irradiated BWR 8×8-2

and 8×8-4 UO2 Fuel Assemblies,” Journal of Nuclear Science and Technology, 45:11, pp.

1193–1214 (2008), DOI: 10.1080/18811248.2008.9711908.

[48] T. Yamamoto, Compilation of Measurement and Analysis Results of Isotopic Inventories of

Spent BWR Fuels, report contributed to the OECD/NEA and included as part of the

SFCOMPO spent fuel database, February 2012.

[49] MALIBU Program, Final Report on the Extension Scope, SCK•CEN report MA 2015/20,

November 2015.

[50] MALIBU Program, Irradiation Data Report of the MALIBU Extension Program, SCK•CEN

report MA 2010/05, March 2010.

[51] “Fuel design data,” Nuclear Engineering International, September 2004.

[52] R. Reager, BWR Spent Fuel Isotopic Characterization, NEDO-33094, Revision 0, GE

Nuclear Energy, February 2003, MOL.20030528.0184.2003.

[53] C. W. Mays, Code to Code Comparison of One- and Two-Dimensional Methods, AREVA

document 32-5048840-00, 2004, DOC.20041015.0003.

[54] R. Gauntt et al., Fukushima Daiichi Accident Study, SAND2012-6173, Sandia National

Laboratories, 2012.

[55] R. S. Moore, Physical Characteristics of GE Fuel Assemblies, ORNL/TM-10902, Oak

Ridge National Laboratory, April 1989.

[56] R. D. Reager and R. B. Adamson, TRW Yucca Mountain Project: Test Report Phase I,

Reference no. A09112CC8A, 1999.

[57] R. D. Reager and R.B. Adamson, TRW Yucca Mountain Project: Test Report Phase II,

Reference no. A09112CC8A, 1999.

[58] S. F. Wolf, D. L. Bowers and J. C. Cunnane, Analysis of Spent Nuclear Fuel Samples from

Three Mile Island and Quad Cities Reactors: Final Report, Argonne National Laboratory

report (November 2000).

[59] Quad Cities Updated Final Safety Analysis Report, Revision 11, October 2011.

[60] D. E. Mueller, S. M. Bowman, W. J. Marshall, and J. M. Scaglione, Review and

Prioritization of Technical Issues Related to Burnup Credit for BWR Fuel, NUREG/CR7158, prepared for the US Nuclear Regulatory Commission by Oak Ridge National

Laboratory, Oak Ridge, Tenn., February 2013.

[61] W. J. Marshall, B. J. Ade, S. Bowman, and J. S. Martinez-Gonzalez, Axial Moderator

Density Distributions, Control Blade Usage, and Axial Burnup Distributions for Extended

BWR Burnup Credit, NUREG/CR-7224 (ORNL/TM-2015/544), prepared for the US

Nuclear Regulatory Commission by Oak Ridge National Laboratory, Oak Ridge, Tenn.,

August 2016.
9-5

[62] M. D. DeHart, Sensitivity and Parametric Evaluations oF Significant Aspects of Burnup

Credit for PWR Spent Fuel Packages, ORNL/TM-12973, Oak Ridge National Laboratory,

May 1996.

[63] M. D. DeHart, A Stochastic Method for Estimating the Effect of Isotopic Uncertainties in

Spent Nuclear Fuel, ORNL/TM-2001/83, Oak Ridge National Laboratory, September

2001.

[64] V. Sobes, B. T. Rearden, D. E. Mueller, W.J. Marshall, J. M. Scaglione, and M. Dunn,

“Upper Subcritical Calculations Based on Correlated Data,” International Conference on

Nuclear Criticality Safety, ICNC 2015, 2015.

[65] S. Morooka, T. Ishizuka, M. Iizuka, and K. Yoshimura, “Experimental Study on Void

Fraction in a Simulated BWR Fuel Assembly (Evaluation of Cross-Sectional Averaged

Void Fraction),” Nuclear Engineering and Design 114, pp. 91–98 (1989).

[66] A. Inoue, T. Kurosu, T. Aoki, M. Yagi, T. Mitsutake, and S. Morooka, “Void Fraction

Distribution in BWR Fuel Assembly and Evaluation of Subchannel Code,” Journal of

Nuclear Science and Technology 32(7), pp. 629–640 (July 1995).

[67] J. C. Wagner, M. D. DeHart, and B. L. Broadhead, Investigation of Burnup Credit Modeling

Issues Associated with BWR Fuel, ORNL/TM-1999/193, Oak Ridge National Laboratory.

[68] OECD Nuclear Energy Agency, Evaluation Guide for the Evaluated Spent Nuclear Fuel

Assay Database (SFCOMPO), NEA/NSC/R 2015, 8, February 2016.

[69] “Standard Test Method for Atom Percent Fission in Uranium and Plutonium Fuel

(Neodymium-148 Method),” ASTM Standard, ASTM E321 – 96, 2012.

[70] C. Devida, M. Betti, P. Peerani, E. H. Toscano, and W. Goll, “A Quantitative Burn-up

Determination: A Comparision of Different Experimental Methods, pp. 106–113 in

HOTLAB: European Hot Laboratories Research Capacities and Needs, “HOTLAB”

Plenary Meeting, Norway, September 6–8, 2004.

APPENDIX A MEASURED NUCLIDE CONCENTRATIONS IN SPENT

FUEL SAMPLES

A-1

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial)

Nuclide

FDN-2

|2F2DN23|

SF98|1 a

FDN-2

|2F2DN23|

SF98|2 a

FDN-2

|2F2DN23

|SF98|3 a

FDN-2

|2F2DN23

|SF98|4 a

FDN-2

|2F2DN23

|SF98|5 a

FDN-2

|2F2DN23

|SF98|6 a

FDN

|2F2D

|SF9

U-234 4.880E-02 2.680E-01 2.180E-01 1.980E-01 1.900E-01 1.860E-01 1.960

U-235 4.13E+00 1.74E+01 8.14E+00 5.97E+00 6.32E+00 9.06E+00 9.36E

U-236 4.86E-01 3.55E+00 4.99E+00 5.28E+00 5.31E+00 5.14E+00 5.14E

U-238 9.88E+02 9.46E+02 9.41E+02 9.36E+02 9.33E+02 9.33E+02 9.33E

Np-237 2.38E-02 1.48E-01 3.35E-01 4.32E-01 3.86E-01 5.16E-01 4.57E

Pu-238 3.13E-03 2.83E-02 1.17E-01 1.68E-01 1.94E-01 1.69E-01 2.08E

Pu-239 2.30E+00 3.37E+00 3.69E+00 3.79E+00 4.26E+00 5.30E+00 5.63E

Pu-240 5.47E-01 1.12E+00 2.14E+00 2.46E+00 2.61E+00 2.63E+00 2.67E

Pu-241 1.33E-01 4.31E-01 8.95E-01 1.03E+00 1.17E+00 1.29E+00 1.36E

Pu-242 1.69E-02 9.29E-02 4.62E-01 6.62E-01 6.94E-01 5.43E-01 5.44E

Am-241 1.03E-02 2.30E-02 3.27E-02 3.42E-02 3.73E-02 4.09E-02 4.39E

Am-243 5.84E-04 6.99E-03 6.68E-02 1.14E-01 1.27E-01 1.12E-01 1.09E

Mo-95 N/A N/A N/A N/A N/A N/A N/

Tc-99 N/A N/A N/A N/A N/A N/A N/

Ru-101 N/A N/A N/A N/A N/A N/A N/

Rh-103 N/A N/A N/A N/A N/A N/A N/

Ag-109 N/A N/A N/A N/A N/A N/A N/

Cs-133 N/A N/A N/A N/A N/A N/A N/

Nd-143 1.21E-01 7.57E-01 8.23E-01 8.49E-01 9.04E-01 9.20E-01 9.18E

Nd-145 9.19E-02 5.97E-01 7.65E-01 8.42E-01 8.67E-01 7.95E-01 7.85E

Sm-147 4.78E-02 2.30E-01 3.09E-01 3.21E-01 3.02E-01 2.89E-01 2.80E

Sm-149 6.37E-04 2.20E-03 2.55E-03 2.50E-03 3.70E-03 3.37E-03 4.20E

Sm-150 3.34E-02 1.79E-01 3.31E-01 3.87E-01 3.81E-01 N/A 3.51E

Sm-151 2.55E-03 8.20E-03 9.19E-03 9.74E-03 1.04E-02 1.27E-02 1.31E

Sm-152 2.23E-02 9.02E-02 1.42E-01 1.56E-01 1.43E-01 1.23E-01 1.22E

Eu-151 N/A N/A N/A N/A N/A N/A N/

Eu-153 N/A N/A N/A N/A N/A N/A N/

Gd-155 N/A N/A N/A N/A N/A N/A N/

a Measurements reported at the time of discharge from the reactor (no cooling time), on 11/14/1992, except for samarium isot

the time of measurement on 5/10/1998.
A2

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

FDN-2|

2F2DN23|

SF99|10 b

FDN-2|

2F2DN23|

SF99|1 b

FDN-2|

2F2DN23|

SF99|2 b

FDN-2|

2F2DN23|

SF99|3 b

FDN-2|

2F2DN23|

SF99|4 b

FDN-2|

2F2DN23|

SF99|5 b

FDN

2F2D

SF99

U-234 4.090E-02 3.900E-02 2.010E-01 1.780E-01 1.670E-01 1.600E-01 1.650

U-235 3.04E+00 2.91E+00 1.40E+01 8.66E+00 6.98E+00 7.38E+00 1.05E

U-236 6.74E-01 6.87E-01 3.47E+00 4.25E+00 4.48E+00 4.52E+00 4.30E

U-238 9.84E+02 9.84E+02 9.52E+02 9.45E+02 9.43E+02 9.39E+02 9.41E

Np-237 5.49E-02 5.67E-02 2.18E-01 3.63E-01 3.67E-01 4.62E-01 4.15E

Pu-238 1.10E-02 1.13E-02 3.96E-02 9.70E-02 1.15E-01 1.23E-01 1.37E

Pu-239 3.01E+00 3.01E+00 3.91E+00 3.98E+00 3.86E+00 4.55E+00 5.63E

Pu-240 1.05E+00 1.06E+00 1.52E+00 2.13E+00 2.29E+00 2.53E+00 2.45E

Pu-241 3.36E-01 3.60E-01 6.76E-01 9.45E-01 1.01E+00 1.20E+00 1.26E

Pu-242 7.33E-02 8.24E-02 1.90E-01 4.37E-01 5.57E-01 6.07E-01 4.33E

Am-241 1.61E-02 2.70E-02 2.11E-02 3.95E-02 3.41E-02 4.36E-02 4.56E

Am-243 5.62E-03 5.79E-03 1.89E-02 6.50E-02 9.04E-02 1.13E-01 8.51E

Mo-95 N/A N/A N/A N/A N/A N/A N/

Tc-99 N/A N/A N/A N/A N/A N/A N/

Ru-101 N/A N/A N/A N/A N/A N/A N/

Rh-103 N/A N/A N/A N/A N/A N/A N/

Ag-109 N/A N/A N/A N/A N/A N/A N/

Cs-133 N/A N/A N/A N/A N/A N/A N/

Nd-143 1.89E-01 1.95E-01 6.14E-01 7.63E-01 7.81E-01 N/A 7.98E

Nd-145 1.52E-01 1.58E-01 4.92E-01 6.73E-01 7.20E-01 N/A 6.52E

Sm-147 N/A 7.78E-02 N/A N/A N/A 2.77E-01 N/

Sm-149 N/A 8.84E-04 N/A N/A N/A 2.72E-03 N/

Sm-150 N/A 6.17E-02 N/A N/A N/A 3.25E-01 N/

Sm-151 N/A 3.41E-03 N/A N/A N/A 1.03E-02 N/

Sm-152 N/A 3.95E-02 N/A N/A N/A 1.27E-01 N/

Eu-151 N/A N/A N/A N/A N/A N/A N/

Eu-153 N/A N/A N/A N/A N/A N/A N/

Gd-155 N/A N/A N/A N/A N/A N/A N/

b Measurements reported at the time of discharge from the reactor (no cooling time), on 11/14/1992, except for Sm isotopes th

date of analysis of 5/21/1999.
A3

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

FDN-2|

2F2DN23|

SF99|8 b

FDN-2|

2F2DN23|

SF99|9 b

LIM-1|

FYJ1433|

D8|3D2B c

LIM-1|

FYJ1433|

D8|4G3 c

LIM-1|

FYJ1433|

D9|1D2 c

LIM-1|

FYJ1433|

D9|2D2 c

LIM

FYJ1

D9|4

U-234 1.960E-01 2.180E-01 1.490E-01 1.820E-01 1.53E-01 1.50E-01 1.51E

U-235 1.58E+01 1.91E+01 3.94E+00 8.18E+00 1.57E+00 2.01E+00 2.47E

U-236 3.46E+00 2.83E+00 5.29E+00 4.86E+00 5.54E+00 5.52E+00 5.55E

U-238 9.49E+02 9.54E+02 N/A N/A N/A N/A N/

Np-237 2.76E-01 1.98E-01 7.33E-01 5.06E-01 7.08E-01 8.26E-01 8.13E

Pu-238 6.48E-02 3.43E-02 4.06E-01 2.39E-01 3.69E-01 4.66E-01 4.98E

Pu-239 5.45E+00 4.73E+00 5.05E+00 5.18E+00 3.64E+00 4.38E+00 4.86E

Pu-240 1.65E+00 1.18E+00 3.43E+00 2.72E+00 2.88E+00 3.21E+00 3.31E

Pu-241 8.33E-01 5.37E-01 1.28E+00 1.07E+00 9.79E-01 1.20E+00 1.31E

Pu-242 1.72E-01 8.33E-02 1.10E+00 5.94E-01 1.42E+00 1.49E+00 1.43E

Am-241 3.62E-02 2.88E-02 3.60E-01 2.97E-01 2.61E-01 2.99E-01 3.50E

Am-243 2.57E-02 9.26E-03 2.81E-01 1.22E-01 3.34E-01 3.60E-01 3.84E

Mo-95 N/A N/A 1.18E+00 9.32E-01 1.32E+00 1.30E+00 1.26E

Tc-99 N/A N/A 1.18E+00 9.37E-01 1.28E+00 1.25E+00 1.28E

Ru-101 N/A N/A 1.19E+00 8.97E-01 1.38E+00 1.39E+00 1.35E

Rh-103 N/A N/A 6.90E-01 5.69E-01 6.67E-01 6.97E-01 7.34E

Ag-109 N/A N/A 1.49E-01 1.03E-01 1.35E-01 1.40E-01 1.42E

Cs-133 N/A N/A N/A N/A N/A N/A N/

Nd-143 6.14E-01 5.01E-01 9.44E-01 8.62E-01 7.87E-01 9.04E-01 9.54E

Nd-145 4.67E-01 3.71E-01 9.35E-01 7.49E-01 1.04E+00 1.04E+00 1.06E

Sm-147 N/A N/A 2.67E-01 2.42E-01 2.72E-01 2.72E-01 2.66E

Sm-149 N/A N/A 2.49E-03 2.73E-03 1.54E-03 2.16E-03 2.58E

Sm-150 N/A N/A 4.45E-01 3.13E-01 4.60E-01 5.06E-01 4.98E

Sm-151 N/A N/A 1.25E-02 1.15E-02 9.32E-03 1.14E-02 1.24E

Sm-152 N/A N/A 1.41E-01 1.13E-01 1.65E-01 1.62E-01 1.52E

Eu-151 N/A N/A 4.42E-04 4.05E-04 3.16E-04 3.71E-04 4.02E

Eu-153 N/A N/A 1.80E-01 1.32E-01 1.93E-01 2.03E-01 1.97E

Gd-155 N/A N/A 1.10E-02 9.08E-03 7.20E-03 8.06E-03 8.42E

b Measurements reported at the time of discharge from the reactor (no cooling time), on 11/14/1992, except for Sm that is rep

analysis of 5/21/1999. 

c Date of analysis: 7/15/2002.
A4

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

DOD-1|

Y013|

B2|DU1 d

DOD-1|

Y013|

B2|DU1 e

FOR-3|

GN592|

J8|ENUSA-1 f

FOR-3|

GN592|

J8|ENUSA-2 f

FOR-3|

GN592|

J8|ENUSA-3 f

FOR-3|

GN592|

J8|ENUSA-4 f

FOR

GN5

J8|ENU

U-234 1.49E-01 1.41E-01 1.80E-01 1.83E-01 1.89E-01 1.75E-01 2.00E

U-235 6.13E+00 5.76E+00 5.72E+00 5.79E+00 4.56E+00 4.97E+00 8.02E

U-236 7.18E+00 7.11E+00 6.08E+00 5.96E+00 5.97E+00 4.98E+00 5.63E

U-238 9.21E+02 9.14E+02 9.31E+02 9.31E+02 9.36E+02 9.29E+02 9.34E

Np-237 6.75E-01 7.08E-01 N/A N/A N/A N/A N/

Pu-238 2.69E-01 3.05E-01 3.55E-01 3.37E-01 2.86E-01 3.73E-01 2.67E

Pu-239 3.74E+00 3.88E+00 5.14E+00 5.16E+00 4.02E+00 4.83E+00 5.05E

Pu-240 2.56E+00 2.71E+00 2.97E+00 2.99E+00 2.67E+00 2.99E+00 2.67E

Pu-241 7.99E-01 9.84E-01 1.25E+00 1.26E+00 9.70E-01 1.22E+00 1.12E

Pu-242 7.99E-01 8.54E-01 9.77E-01 9.64E-01 9.34E-01 1.06E+00 7.07E

Am-241 3.68E-01 2.05E-01 3.40E-01 3.46E-01 2.64E-01 3.28E-01 2.91E

Am-243 1.80E-01 1.72E-01 2.12E-01 2.31E-01 1.81E-01 2.46E-01 1.29E

Mo-95 1.09E+00 1.36E+00 1.14E+00 1.10E+00 1.12E+00 1.01E+00 1.03E

Tc-99 9.82E-01 1.29E+00 N/A N/A N/A N/A N/

Ru-101 8.65E-01 1.13E+00 N/A N/A N/A N/A N/

Rh-103 5.20E-01 5.73E-01 N/A N/A N/A N/A N/

Ag-109 4.50E-02 5.31E-02 N/A N/A N/A N/A N/

Cs-133 1.72E+00 1.70E+00 1.89E+00 2.19E+00 1.72E+00 1.76E+00 1.56E

Nd-143 9.81E-01 1.05E+00 9.27E-01 9.53E-01 8.64E-01 9.37E-01 8.98E

Nd-145 1.08E+00 1.09E+00 9.33E-01 9.47E-01 9.39E-01 9.59E-01 8.49E

Sm-147 3.57E-01 3.10E-01 2.66E-01 2.64E-01 2.63E-01 2.45E-01 3.29E

Sm-149 2.44E-03 3.01E-03 2.30E-03 2.50E-03 1.80E-03 2.50E-03 2.50E

Sm-150 4.55E-01 4.42E-01 4.18E-01 4.17E-01 3.84E-01 4.09E-01 3.41E

Sm-151 1.08E-02 1.08E-02 1.20E-02 1.30E-02 9.00E-03 1.10E-02 9.00E

Sm-152 1.67E-01 1.61E-01 1.47E-01 1.50E-01 1.58E-01 1.49E-01 1.33E

Eu-151 6.19E-04 6.85E-04 N/A N/A N/A N/A N/

Eu-153 1.70E-01 1.76E-01 1.61E-01 1.57E-01 1.55E-01 1.67E-01 1.51E

Gd-155 8.11E-03 5.65E-03 5.40E-03 5.20E-03 5.20E-03 6.00E-03 4.00E

d Measurements performed by PSI. Date of analysis: 4/22/1999 (U), 4/12/1999 (Pu), 9/15/1999 (Np), 12/1/1999 (Am), 5/28/19

Eu, Gd), 9/6/1999 (Cs), 1/11/2000 (Mo, Ru, Ag), 12/13/1999 (Rh), 1/20/2000 (Tc).

e Measurements performed by SCK•CEN. Date of analysis: 5/6/1996 (U, Np), 4/12/1996 (Pu), 6/18/1996 (Am), 10/16/1996 (N

11/28/1996 (Eu), 6/24/1996 (Gd), 10/1/1996 (Cs), 4/10/2000 (Mo, Tc, Ru, Rh, Ag). f Date of analysis: 6/15/2009 (U), 6/16/2009 (Pu, Am), 7/1/2009 (Nd, Sm, Eu, Gd), 8/31/2009 (Cs), 10/28/2009 (Mo).
A5

T Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

FOR-3|

GN592|

J8|ENUSA-7 f

FOR-3|

GN592|J8|

ENUSA-8 f

FOR-3|

14595|

F6|F3F6 g

LIM-1|

FYJ1433|

H5|3A1C c

LIM-1|

FYJ1433|

H5|3A1G c

LBT|

AIA003|

KLU1 h

LB

AIA0

KLU

U-234 1.84E-01 2.07E-01 1.50E-01 1.66E-01 1.66E-01 1.51E-01 1.49E

U-235 4.44E+00 9.43E+00 3.16E+00 4.42E+00 4.50E+00 1.91E+00 2.76E

U-236 5.80E+00 5.24E+00 5.66E+00 5.72E+00 5.72E+00 5.99E+00 6.06E

U-238 9.37E+02 9.39E+02 9.17E+02 N/A N/A 9.21E+02 9.15E

Np-237 N/A N/A N/A 7.92E-01 7.97E-01 6.58E-01 7.52E

Pu-238 2.57E-01 2.05E-01 3.88E-01 4.67E-01 4.68E-01 3.31E-01 4.04E

Pu-239 3.98E+00 4.90E+00 4.62E+00 5.66E+00 5.69E+00 3.32E+00 4.06E

Pu-240 2.65E+00 2.41E+00 3.05E+00 3.46E+00 3.47E+00 2.75E+00 3.18E

Pu-241 9.62E-01 9.89E-01 8.90E-01 1.38E+00 1.38E+00 8.63E-01 1.09E

Pu-242 9.11E-01 5.41E-01 1.26E+00 1.07E+00 1.05E+00 1.31E+00 1.31E

Am-241 2.61E-01 2.71E-01 N/A 3.76E-01 3.81E-01 3.01E-01 3.89E

Am-243 1.78E-01 1.19E-01 N/A 2.73E-01 2.74E-01 2.98E-01 3.50E

Mo-95 1.06E+00 9.11E-01 N/A 1.18E+00 1.18E+00 1.09E+00 N/

Tc-99 N/A N/A N/A 1.12E+00 1.09E+00 1.10E+00 1.11E

Ru-101 N/A N/A N/A 1.22E+00 1.25E+00 N/A N/

Rh-103 N/A N/A N/A 6.95E-01 7.22E-01 5.79E-01 N/

Ag-109 N/A N/A N/A 1.23E-01 1.39E-01 1.15E-01 N/

Cs-133 1.70E+00 1.42E+00 N/A N/A N/A 1.83E+00 2.03E

Nd-143 8.24E-01 8.61E-01 9.69E-01 1.02E+00 1.02E+00 8.77E-01 1.07E

Nd-145 8.97E-01 7.64E-01 1.02E+00 9.86E-01 9.86E-01 1.11E+00 1.16E

Sm-147 2.53E-01 3.14E-01 N/A 2.71E-01 2.72E-01 3.05E-01 2.94E

Sm-149 1.90E-03 N/A N/A 2.71E-03 2.72E-03 1.58E-03 N/

Sm-150 3.76E-01 2.98E-01 N/A 4.67E-01 4.64E-01 4.74E-01 4.80E

Sm-151 7.60E-03 1.00E-02 N/A 1.42E-02 1.41E-02 9.63E-03 8.83E

Sm-152 1.53E-01 1.24E-01 N/A 1.37E-01 1.36E-01 1.77E-01 1.60E

Eu-151 N/A N/A N/A 4.78E-04 4.85E-04 5.49E-04 N/

Eu-153 1.55E-01 1.19E-01 N/A 1.81E-01 1.81E-01 2.00E-01 1.97E

Gd-155 4.50E-03 3.80E-03 N/A 7.92E-03 7.45E-03 7.59E-03 6.16E

g Date of analysis 12/14/2006.

h Date of analysis 12/31/2009.
A6

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

FDN-2|

2F2D1|

F6|TU101 i

FD2-2|

2F2D1|

F6|TU102 j

FD2-2|

2F2D1|

B3|TU103 k

FD2-2|

2F2D1|

B3|TU104 k

FD2-2|

2F2D1|

B3|TU105 k

FD2-2|

2F2D1|

F6|TU106 

U-234 3.33E-01 3.41E-01 2.46E-01 2.56E-01 2.55E-01 3.12E-01 

U-235 3.33E+01 2.76E+01 2.51E+01 2.33E+01 2.40E+01 2.93E+0

U-236 2.57E+00 2.96E+00 1.77E+00 2.00E+00 2.00E+00 3.06E+0

U-238 9.43E+02 9.45E+02 9.57E+02 9.59E+02 9.55E+02 9.43E+0

Np-237 N/A N/A N/A N/A N/A N/A 

Pu-238 1.96E-02 1.95E-02 9.84E-03 1.97E-02 9.81E-03 2.93E-02 

Pu-239 4.76E+00 4.04E+00 4.73E+00 4.54E+00 4.55E+00 5.63E+0

Pu-240 7.15E-01 8.78E-01 7.18E-01 8.17E-01 9.42E-01 1.02E+0

Pu-241 2.74E-01 3.03E-01 2.66E-01 3.25E-01 3.73E-01 4.59E-01 

Pu-242 2.94E-02 3.90E-02 1.97E-02 3.94E-02 3.93E-02 5.86E-02 

Am-241 5.82E-02 6.82E-02 5.74E-02 6.97E-02 6.82E-02 9.95E-02 

Am-243 1.70E-03 1.59E-03 1.51E-03 1.84E-03 1.50E-03 1.78E-03 

Mo-95 N/A N/A N/A N/A N/A N/A 

Tc-99 N/A N/A N/A N/A N/A N/A 

Ru-101 N/A N/A N/A N/A N/A N/A 

Rh-103 N/A N/A N/A N/A N/A N/A 

Ag-109 N/A N/A N/A N/A N/A N/A 

Cs-133 N/A N/A N/A N/A N/A N/A 

Nd-143 N/A N/A N/A N/A N/A N/A 

Nd-145 N/A N/A N/A N/A N/A N/A 

Sm-147 N/A N/A N/A N/A N/A N/A 

Sm-149 N/A N/A N/A N/A N/A N/A 

Sm-150 N/A N/A N/A N/A N/A N/A 

Sm-151 N/A N/A N/A N/A N/A N/A 

Sm-152 N/A N/A N/A N/A N/A N/A 

Eu-151 N/A N/A N/A N/A N/A N/A 

Eu-153 N/A N/A N/A N/A N/A N/A 

Gd-155 N/A N/A N/A N/A N/A N/A 

i Date of analysis 9/5/1994.

j Date of analysis 8/31/1994.

k Date of analysis 10/31/1994.
A7

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

FD2-2|

2F2D2|

F6|TU202 k

FD2-2|

2F2D2|

B3|TU203 j

FD2-2|

2F2D2|

B3|TU204 k

FD2-2|

2F2D2|

B3|TU205 k

FD2-2|

2F2D3|

H5|TU301 l

FD2-2|

2F2D3|

H5|TU302 l

FD2

2F2

A4|TU

U-234 2.68E-01 1.93E-01 2.03E-01 2.03E-01 1.62E-01 1.63E-01 1.61E

U-235 1.56E+01 1.55E+01 1.43E+01 1.24E+01 6.46E+00 6.29E+00 6.21E

U-236 4.83E+00 3.44E+00 3.42E+00 3.57E+00 4.61E+00 4.47E+00 4.53E

U-238 9.37E+02 9.45E+02 9.48E+02 9.53E+02 9.42E+02 9.49E+02 9.39E

Np-237 N/A N/A N/A N/A N/A N/A N/

Pu-238 8.62E-02 8.68E-02 9.66E-02 6.78E-02 2.67E-01 1.25E-01 2.56E

Pu-239 4.47E+00 6.49E+00 6.30E+00 4.19E+00 5.59E+00 3.52E+00 5.34E

Pu-240 1.90E+00 2.00E+00 2.08E+00 1.80E+00 2.84E+00 2.11E+00 2.71E

Pu-241 8.14E-01 1.00E+00 9.85E-01 7.07E-01 1.33E+00 7.68E-01 1.27E

Pu-242 2.68E-01 2.22E-01 2.61E-01 2.52E-01 7.53E-01 4.99E-01 7.31E

Am-241 1.37E-01 1.55E-01 1.73E-01 1.25E-01 2.37E-01 1.36E-01 2.09E

Am-243 1.84E-02 1.96E-02 3.40E-02 2.28E-02 1.41E-01 5.28E-02 1.41E

Mo-95 N/A N/A N/A N/A N/A N/A N/

Tc-99 N/A N/A N/A N/A N/A N/A N/

Ru-101 N/A N/A N/A N/A N/A N/A N/

Rh-103 N/A N/A N/A N/A N/A N/A N/

Ag-109 N/A N/A N/A N/A N/A N/A N/

Cs-133 N/A N/A N/A N/A N/A N/A N/

Nd-143 N/A N/A N/A N/A N/A N/A N/

Nd-145 N/A N/A N/A N/A N/A N/A N/

Sm-147 N/A N/A N/A N/A N/A N/A N/

Sm-149 N/A N/A N/A N/A N/A N/A N/

Sm-150 N/A N/A N/A N/A N/A N/A N/

Sm-151 N/A N/A N/A N/A N/A N/A N/

Sm-152 N/A N/A N/A N/A N/A N/A N/

Eu-151 N/A N/A N/A N/A N/A N/A N/

Eu-153 N/A N/A N/A N/A N/A N/A N/

Gd-155 N/A N/A N/A N/A N/A N/A N/

l Date of analysis 8/1/1995.
A8

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

FD2-2|

2F2D3|

B3|TU308 l

FD2-2|

2F2D3|

B3|TU309 l

FD2-2|

2F2D3|

B3|TU311 l

FDN-2|

2F2D8|

H5|TU501 m

FDN-2|

2F2D8|

H5|TU502 m

FDN-2|

2F2D8|

H5|TU503 m

FDN

2F2

A4|TU

U-234 2.11E-01 2.00E-01 2.11E-01 1.40E-01 1.39E-01 1.21E-01 1.39E

U-235 1.07E+01 8.86E+00 6.84E+00 3.71E+00 2.32E+00 1.40E+00 2.27E

U-236 4.22E+00 4.51E+00 4.58E+00 4.89E+00 4.90E+00 4.86E+00 4.85E

U-238 9.43E+02 9.37E+02 9.46E+02 9.25E+02 9.20E+02 9.27E+02 9.19E

Np-237 N/A N/A N/A N/A N/A N/A N/

Pu-238 1.63E-01 2.76E-01 1.34E-01 4.48E-01 4.73E-01 3.18E-01 5.28E

Pu-239 5.82E+00 7.22E+00 3.75E+00 4.75E+00 4.20E+00 2.97E+00 4.81E

Pu-240 2.43E+00 3.44E+00 2.30E+00 3.02E+00 3.01E+00 2.44E+00 3.49E

Pu-241 1.04E+00 1.49E+00 8.33E-01 1.32E+00 1.33E+00 8.68E-01 1.39E

Pu-242 3.93E-01 7.13E-01 5.46E-01 1.03E+00 1.26E+00 1.16E+00 1.49E

Am-241 1.90E-01 2.57E-01 1.49E-01 1.83E-01 2.42E-01 1.41E-01 2.60E

Am-243 6.90E-02 1.02E-01 6.03E-02 2.08E-01 3.05E-01 2.50E-01 3.55E

Mo-95 N/A N/A N/A N/A N/A N/A N/

Tc-99 N/A N/A N/A N/A N/A N/A N/

Ru-101 N/A N/A N/A N/A N/A N/A N/

Rh-103 N/A N/A N/A N/A N/A N/A N/

Ag-109 N/A N/A N/A N/A N/A N/A N/

Cs-133 N/A N/A N/A N/A N/A N/A N/

Nd-143 N/A N/A N/A N/A N/A N/A N/

Nd-145 N/A N/A N/A N/A N/A N/A N/

Sm-147 N/A N/A N/A N/A N/A N/A N/

Sm-149 N/A N/A N/A N/A N/A N/A N/

Sm-150 N/A N/A N/A N/A N/A N/A N/

Sm-151 N/A N/A N/A N/A N/A N/A N/

Sm-152 N/A N/A N/A N/A N/A N/A N/

Eu-151 N/A N/A N/A N/A N/A N/A N/

Eu-153 N/A N/A N/A N/A N/A N/A N/

Gd-155 N/A N/A N/A N/A N/A N/A N/

m Date of analysis 7/6/1999.

n Date of analysis 7/8/1999.
A9

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

FD2-2|

2F2D8|B3| 

TU510 o

FD2-2|

2F2D8|B3|

TU511 p

FD2-1|

2F1ZN2|

C2|GdB q

FD2-1|

2F1ZN2|

C2|GdT r

FD2-1|

2F1ZN2|

C3|UB q

FD2-1|

2F1ZN2|

C3|UT s

FD2

2F1Z

A9|U

U-234 1.40E-01 1.50E-01 1.88E-01 1.61E-01 2.78E-01 3.08E-01 9.14E

U-235 4.50E+00 2.94E+00 5.85E+00 1.07E+01 9.59E+00 1.64E+01 9.18E

U-236 5.09E+00 5.12E+00 4.07E+00 3.63E+00 6.55E+00 5.92E+00 2.67E

U-238 9.22E+02 9.32E+02 9.45E+02 9.48E+02 9.26E+02 9.29E+02 9.26E

Np-237 N/A N/A N/A N/A N/A N/A N/

Pu-238 5.03E-01 3.10E-01 1.56E-01 1.24E-01 2.07E-01 1.72E-01 2.73E

Pu-239 5.74E+00 3.64E+00 4.01E+00 5.63E+00 4.05E+00 5.83E+00 2.91E

Pu-240 3.86E+00 2.94E+00 2.47E+00 2.26E+00 2.43E+00 2.29E+00 2.50E

Pu-241 1.51E+00 9.69E-01 9.55E-01 7.23E-01 9.57E-01 6.70E-01 8.64E

Pu-242 1.15E+00 1.14E+00 6.21E-01 3.44E-01 5.99E-01 3.63E-01 1.97E

Am-241 2.78E-01 2.26E-01 N/A N/A N/A N/A N/

Am-243 2.47E-01 2.10E-01 N/A N/A N/A N/A N/

Mo-95 N/A N/A N/A 6.80E-01 N/A 9.33E-01 N/

Tc-99 N/A N/A N/A 4.96E-01 N/A 6.59E-01 N/

Ru-101 N/A N/A N/A 6.73E-01 N/A 8.93E-01 N/

Rh-103 N/A N/A N/A 4.32E-01 N/A 5.19E-01 N/

Ag-109 N/A N/A N/A 4.51E-02 N/A 4.47E-02 N/

Cs-133 N/A N/A N/A N/A N/A 1.30E+00 N/

Nd-143 N/A N/A 7.61E-01 7.08E-01 1.05E+00 9.88E-01 5.15E

Nd-145 N/A N/A 7.08E-01 5.67E-01 9.70E-01 7.91E-01 9.41E

Sm-147 N/A N/A N/A 2.38E-01 N/A 3.17E-01 N/

Sm-149 N/A N/A N/A 3.07E-03 N/A 3.65E-03 N/

Sm-150 N/A N/A N/A 2.30E-01 N/A 3.04E-01 N/

Sm-151 N/A N/A N/A 1.09E-02 N/A 1.34E-02 N/

Sm-152 N/A N/A N/A 8.94E-02 N/A 1.09E-01 N/

Eu-151 N/A N/A N/A 1.41E-03 N/A 1.28E-03 N/

Eu-153 N/A N/A N/A 9.58E-02 N/A 1.13E-01 N/

Gd-155 N/A N/A N/A 1.27E-02 N/A 6.35E-03 N/

o Date of analysis: 8/2/1999; p Date of analysis: 8/4/1999; q Date of analysis: 9/10/2002. r Date of analysis: 1/12/2010 (U, Pu, Nd, Gd), 1/9/2011 (Sm, Eu, Cs), 12/29/2011 (Tc, Mo, Ru, Rh, Ag). s Date of analysis: 7/20/2012 (U, Pu, Nd), 10/28/2010 (Eu, Sm, Gd), 12/29/2011 (Cs), 12/29/2011 (Tc, Mo, Ru, Rh, Ag). t Date of analysis: 3/10/2005.

u Date of analysis: 1/11/2010 (U, Pu, Nd), 12/20/2010 (Sm, Eu, Gd), 12/31/2011 (Cs, Tc, Mo, Ru, Rh, Ag).
A-1

0

Table A.1 Measured Nuclide Concentrations in Spent Fuel Samples (mg/g uranium initial) (con’t)

Nuclide

FD2-1|

2F1ZN3|

A9|UT t

FD2-1|

2F1ZN3|

C2|GdB v

FD2-1|

2F1ZN3|

C2|GdM w

FD2-1|

2F1ZN3|

C2|GdT v

FD2-1|

2F1ZN3|

C3|UB v

FD2-1|

2F1ZN3|

C3|UM u

U-234 9.17E-02 9.20E-02 1.29E-01 9.25E-02 1.81E-01 2.11E-01 

U-235 6.45E-01 1.29E+00 2.14E+00 4.18E+00 2.18E+00 3.35E+00 

U-236 2.87E+00 4.36E+00 4.44E+00 4.39E+00 7.12E+00 7.19E+00 

U-238 9.29E+02 9.30E+02 9.27E+02 9.33E+02 9.11E+02 9.08E+02 

Np-237 N/A N/A N/A N/A N/A N/A 

Pu-238 3.81E-01 3.33E-01 3.60E-01 3.81E-01 4.44E-01 5.29E-01 

Pu-239 4.07E+00 3.44E+00 4.43E+00 5.42E+00 3.52E+00 4.32E+00 

Pu-240 2.94E+00 2.85E+00 3.37E+00 3.44E+00 2.94E+00 3.32E+00 

Pu-241 1.24E+00 1.01E+00 9.40E-01 1.43E+00 1.04E+00 1.04E+00 

Pu-242 1.66E+00 1.41E+00 1.41E+00 1.02E+00 1.43E+00 1.39E+00 

Am-241 N/A N/A N/A N/A N/A N/A 

Am-243 N/A N/A N/A N/A N/A N/A 

Mo-95 N/A N/A 1.18E+00 N/A N/A 1.53E+00 

Tc-99 N/A N/A 8.44E-01 N/A N/A 1.08E+00 

Ru-101 N/A N/A 1.27E+00 N/A N/A 1.60E+00 

Rh-103 N/A N/A 6.55E-01 N/A N/A 7.48E-01 

Ag-109 N/A N/A 9.82E-02 N/A N/A 1.06E-01 

Cs-133 N/A N/A 1.74E+00 N/A N/A 2.09E+00 

Nd-143 7.34E-01 7.36E-01 8.53E-01 8.95E-01 9.31E-01 1.08E+00 

Nd-145 9.21E-01 9.54E-01 9.56E-01 8.52E-01 1.23E+00 1.23E+00 

Sm-147 N/A N/A 3.13E-01 N/A N/A 3.76E-01 

Sm-149 N/A N/A 2.02E-03 N/A N/A 2.19E-03 

Sm-150 N/A N/A 4.44E-01 N/A N/A 5.30E-01 

Sm-151 N/A N/A 1.04E-02 N/A N/A 1.17E-02 

Sm-152 N/A N/A 1.48E-01 N/A N/A 1.69E-01 

Eu-151 N/A N/A 1.21E-03 N/A N/A 1.15E-03 

Eu-153 N/A N/A 1.85E-01 N/A N/A 2.09E-01 

Gd-155 N/A N/A 1.42E-02 N/A N/A 1.16E-02 

v Date of analysis: 8/25/2005.

w Date of analysis: 7/19/2012 (U, Pu, Nd), 10/11/2011 (Eu, Sm, Gd), 12/31/2011 (Cs, Tc, Mo, Ru, Rh, Ag).
APPENDIX B COMPARISON OF POLARIS-CALCULATED NUCLIDE 

COMPOSITIONS WITH MEASUREMENTS

B-1

Table B.1 Comparisons of Polaris Calculations and Measurements for Burnup Credit Isotopes (C/M – 1)

Nuclide

FD2-2|

2F2DN23|

SF98|1

FD2-2|

2F2DN23|

SF98|2

FD2-2|

2F2DN23|

SF98|3

FD2-2|

2F2DN23|

SF98|4

FD2-2|

2F2DN23|

SF98|5

FD2-2|

2F2DN23|

SF98|6

FD2-2|

2F2DN23|

SF98|7

FD2-2|

2F2DN23|

SF98|8

2

U-234 -9.3% 11.4% 18.0% 18.7% 19.2% 28.0% 20.8% 19.9% 

U-235 -1.7% -15.1% 5.2% 5.6% 0.8% -0.9% 5.2% 4.4% 

U-236 -2.1% 13.6% -1.3% -1.0% -0.5% -1.2% -1.6% -3.4% 

U-238 0.0% 0.1% 0.0% 0.1% 0.1% 0.2% 0.1% 0.0% 

Np-237 -11.4% 46.4% 2.0% -3.9% 20.8% -11.3% 5.0% 6.1% 

Pu-238 -13.3% 93.6% 6.5% 5.4% 10.7% 19.1% 4.2% 1.1% 

Pu-239 -7.6% 9.6% 0.4% -0.9% -1.2% -5.4% 0.8% 3.8% 

Pu-240 -5.1% 31.4% -3.1% -4.8% -4.8% -8.1% -7.1% -5.3% 

Pu-241 -16.5% 40.1% -2.6% -3.1% -2.5% -5.1% -2.4% -0.4% 

Pu-242 -16.0% 87.9% -5.3% -5.4% -1.0% 0.0% -4.4% -4.6% 

Am-241 -50.3% 3.4% -14.3% -17.3% -13.1% -2.7% 1.7% 10.9% 

Am-243 -36.6% 122.8% -16.0% -16.6% -8.2% -16.9% -13.3% -12.2% 

Mo-95

Tc-99

Ru-101

Rh-103

Ag-109

Cs-133

Nd-143 -0.1% -4.1% 3.7% 4.6% 2.3% 1.1% 2.6% 2.4% 

Nd-145 -0.3% -1.3% 1.4% 2.1% 1.2% 1.4% 0.7% 0.8% 

Sm-147 -8.5% 2.0% -10.4% -10.8% -4.8% -4.8% -5.7% -9.6% 

Sm-149 -10.0% 5.4% 1.1% 11.9% -17.5% -3.2% -15.7% -21.4% 

Sm-150 -8.1% 15.8% -9.9% -10.4% -4.2% -5.8% -9.2% 

Sm-151 -14.3% -3.2% -10.9% -11.5% -7.0% -12.7% -4.9% -6.2% 

Sm-152 -7.1% 12.0% -7.8% -8.5% -1.8% 1.1% -3.1% -7.6% 

Eu-151

Eu-153

Gd-155
B2

Table B.1 Comparisons of Polaris Calculations and Measurements for Burnup Credit Isotopes (C/M – 1)

Nuclide

FD2-2|

2F2DN23|

SF99|3

FD2-2|

2F2DN23|

SF99|4

FD2-2|

2F2DN23|

SF99|5

FD2-2|

2F2DN23|

SF99|6

FD2-2|

2F2DN23|

SF99|7

FD2-2|

2F2DN23|

SF99|8

FD2-2|

2F2DN23|

SF99|9

F

F

F6|

U-234 7.1% 8.8% 9.1% 12.1% 11.9% 7.9% 4.6% 

U-235 5.7% 12.7% 2.9% 4.4% 6.3% 8.8% 5.1% 

U-236 2.3% 1.2% 2.4% 0.4% 1.5% 0.2% 2.9% 

U-238 0.0% 0.0% 0.1% 0.1% 0.0% -0.1% -0.2% 

Np-237 -11.2% -1.1% -11.3% -6.0% -8.2% -2.1% -2.2%

Pu-238 24.9% 30.0% 48.6% 18.5% 26.9% 28.0% 32.5% 

Pu-239 0.8% 5.2% -2.1% -2.2% 1.7% 10.1% 13.7% 

Pu-240 -1.3% -1.0% -4.2% -6.1% -5.1% 0.9% 2.0% 

Pu-241 -1.0% 1.0% -3.6% -3.0% 0.0% 6.5% 12.1% 

Pu-242 -0.9% -3.9% 0.1% 0.6% 0.3% 2.1% 6.4% 

Am-241 -13.2% 2.8% -10.6% 4.3% 7.1% 14.0% 1.3% 

Am-243 -13.6% -15.1% -13.9% -18.1% -15.3% -13.1% -3.9% 

Mo-95

Tc-99

Ru-101

Rh-103

Ag-109

Cs-133

Nd-143 2.2% 4.1% 1.7% 0.9% 0.5% -0.5%

Nd-145 0.8% 1.1% 1.3% 0.0% -1.1% -1.6%

Sm-147 -1.7%

Sm-149 5.0%

Sm-150 -4.4%

Sm-151 -8.3%

Sm-152 0.2%

Eu-151

Eu-153

Gd-155
B3

Table B.1 Comparisons of Polaris Calculations and Measurements for Burnup Credit Isotopes (C/M – 1)

Nuclide

FD2-2|

2F2D1|

B3|TU103

FD2-2|

2F2D1|

B3|TU104

FD2-2|

2F2D1|

B3|TU105

FD2-2|

2F2D1|

F6|TU106

FD2-2|

2F2D2|

F6|TU201

FD2-2|

F2D2|

F6|TU202

FD2-2|

2F2D2|

B3|TU203

FD2-2|

2F2D2|

B3|TU204B

U-234 1.3% -1.6% -3.0% 5.3% 5.2% 5.0% 7.7% 4.2% 

U-235 0.1% 10.1% -5.1% -0.3% -4.3% -0.1% -1.5% 9.7% 

U-236 14.1% -4.6% 14.1% 4.1% 7.8% 6.8% 7.7% 5.5% 

U-238 0.0% -0.1% 0.2% 0.2% 0.2% 0.1% 0.2% 0.0% 

Np-237

Pu-238 44.6% -38.8% 46.1% -23.2% -4.9% 3.9% 3.2% -18.1% 

Pu-239 -7.5% -8.0% -17.2% -20.5% -12.5% -6.8% -14.4% -15.2% 

Pu-240 -1.3% -21.1% -12.6% -20.2% -8.0% -8.0% -9.0% -16.8% 

Pu-241 -8.6% -34.5% -30.8% -33.2% -20.5% -16.1% -17.6% -22.5% 

Pu-242 40.0% -42.6% 3.4% -28.5% 20.5% -3.8% 4.6% -19.4% 

Am-241 15.6% -16.5% 3.3% -16.6% 7.2% -0.4% 4.0% -9.2% 

Am-243 10.8% -31.8% 49.9% 68.6% 120.6% 55.5% 55.3% -23.9% 

Mo-95

Tc-99

Ru-101

Rh-103

Ag-109

Cs-133

Nd-143

Nd-145

Sm-147

Sm-149

Sm-150

Sm-151

Sm-152

Eu-151

Eu-153

Gd-155
B4

Table B.1 Comparisons of Polaris Calculations and Measurements for Burnup Credit Isotopes (C/M – 1)

Nuclide

FD2-2|

2F2D3|

H5|TU302

FD2-2|

2F2D3|

A4|TU304

FD2-2|

2F2D3|

A4|TU306

FD2-2|

2F2D3|

B3|TU308

FD2-2|

2F2D3|

B3|TU309

FD2-2|

2F2D3|

B3|TU311

FD2-2|

2F2D8|

H5|TU501

FD2-2|

2F2D8|

H5|TU502H

U-234 20.6% 8.5% 6.5% -8.3% -8.8% -9.8% 4.4% -2.7% 

U-235 34.7% 22.0% 34.9% 14.3% 12.2% 27.1% -11.3% -7.3% 

U-236 -5.0% 0.5% -3.3% -0.8% -0.6% -3.3% 1.0% 0.8% 

U-238 -0.2% 0.0% -0.2% -0.1% 0.3% -0.2% 0.0% 0.0% 

Np-237

Pu-238 -17.9% -18.3% -18.2% -8.9% -29.8% -0.3% 11.8% 19.2% 

Pu-239 6.9% -1.2% 4.9% -2.3% -22.8% 6.7% 9.6% 15.7% 

Pu-240 -8.9% -9.6% -8.8% -9.4% -28.3% -5.4% 2.2% 5.2% 

Pu-241 -5.9% -9.7% -6.5% -2.4% -23.9% 0.8% 4.9% 3.0% 

Pu-242 -27.1% -19.8% -27.0% -6.0% -29.1% -14.3% 15.1% 13.9% 

Am-241 -0.5% 2.4% -5.2% 5.1% -15.2% 6.6% 69.1% 19.8% 

Am-243 -18.4% -26.7% -33.3% -16.2% -14.3% 4.9% 30.5% 13.1% 

Mo-95

Tc-99

Ru-101

Rh-103

Ag-109

Cs-133

Nd-143

Nd-145

Sm-147

Sm-149

Sm-150

Sm-151

Sm-152

Eu-151

Eu-153

Gd-155
B5

Table B.1 Comparisons of Polaris Calculations and Measurements for Burnup Credit Isotopes (C/M – 1)

Nuclide

FD2-2|

2F2D8|

A4|TU506

FD2-2|

2F2D8|

B3|TU510

FD2-2|

2F2D8|

B3|TU511

FD2-1|

2F1ZN2|

C2|GdB

FD2-1|

2F1ZN2|

C2|GdT

FD2-1|

2F1ZN2|

C3|UB

FD2-1|

2F1ZN2|

C3|UT

FD2-1|

2F1ZN3|

C2|GdB

2

C

U-234 1.3% 8.4% 5.3% -17.7% 8.9% -5.4% -3.6% 27.0% 

U-235 -12.3% -7.5% 17.6% 5.4% 3.5% 1.6% 4.6% 13.0% 

U-236 4.5% 0.6% -1.8% 1.0% 1.7% 1.8% 2.0% 3.2% 

U-238 0.0% 0.3% 0.1% -0.1% -0.2% -0.1% -0.2% -0.1% 

Np-237

Pu-238 30.0% 2.8% 10.7% 6.2% 8.3% 4.8% 2.8% 14.3% 

Pu-239 7.3% -5.6% 3.5% 2.9% 6.3% 1.8% 3.9% 2.1% 

Pu-240 -1.1% -15.7% -6.9% -0.6% -0.1% -1.6% -3.8% 3.7% 

Pu-241 8.2% -6.8% 2.1% 1.5% 4.5% 0.6% 1.4% 1.2% 

Pu-242 10.1% -0.6% -10.3% 1.2% 4.2% 1.4% -0.8% 5.6% 

Am-241 -5.5% 19.0% 0.1%

Am-243 28.4% 7.0% -7.7%

Mo-95 -7.6% -4.4% 

Tc-99 37.4% 39.9% 

Ru-101 -3.1% -2.1% 

Rh-103 0.9% 1.5% 

Ag-109 46.3% 50.5% 

Cs-133 1.4% 

Nd-143 2.6% 2.1% 2.1% 2.4% 6.2% 

Nd-145 0.6% 0.5% 0.9% 0.8% 1.9% 

Sm-147 -0.7% -1.0% 

Sm-149 5.6% 14.5% 

Sm-150 1.4% 3.9% 

Sm-151 2.5% 4.0% 

Sm-152 6.3% 7.4% 

Eu-151 -30.7% -7.4% 

Eu-153 11.4% 5.4% 

Gd-155 -8.5% 10.9% 
B6

Table B.1 Comparisons of Polaris Calculations and Measurements for Burnup Credit Isotopes (C/M – 1)

Nuclide

FD2-1|

2F1ZN3|

A9|UB

FD2-1|

2F1ZN3|

A9|UM

FD2-1|

2F1ZN3|

A9|UT

FD2-1|

2F1ZN3|

C3|UB

FD2-1|

2F1ZN3|

C3|UM

FD2-1|

2F1ZN3|

C3|UT

DOD|

Y013|

U0004|DU1 a

DODI|

Y013|

U0004|DU1 b

U-234 -37.0% 18.3% -21.9% 8.7% 2.3% 23.7% 66.6% 

U-235 26.0% -4.7% 2.2% 13.9% -0.7% 10.6% 0.4% 

U-236 3.2% 3.5% 1.3% 2.9% 3.2% 2.6% -0.7% 

U-238 -0.2% -0.1% -0.2% -0.1% -0.2% -0.2% -0.3% 

Np-237 -3.6%

Pu-238 27.2% 10.0% 15.0% 9.7% 6.5% 2.5% 23.3% 

Pu-239 4.8% -9.8% 3.3% -0.8% -6.3% 2.2% 10.2% 

Pu-240 12.5% 0.1% 6.5% -0.8% -3.0% -2.9% 7.7% 

Pu-241 4.5% -10.3% 2.9% -2.2% -5.9% 1.8% 9.2% 

Pu-242 13.9% 3.8% 11.2% 0.0% 5.8% 2.6% 13.4% 

Am-241 8.2% 

Am-243 -5.7% 

Mo-95 -3.1% -3.8% 16.8% 

Tc-99 49.5% 43.2% 35.4%

Ru-101 -3.3% -2.2% 48.8%

Rh-103 1.1% 2.5% 31.1%

Ag-109 20.2% 46.4% 147.2%

Cs-133 1.9% 2.7% 7.7% 

Nd-143 5.7% 1.9% 3.4% 6.5% 3.5% 4.7% 12.2% 

Nd-145 3.0% 1.5% 1.2% 1.9% 2.8% 1.4% 2.6% 

Sm-147 1.6% 0.6% 0.9% 

Sm-149 -15.2% 0.7% 2.2% 

Sm-150 2.2% 4.7% 1.6% 

Sm-151 -10.0% -4.7% -0.1% 

Sm-152 9.8% 13.1% 0.7% 

Eu-151 -39.3% -38.0% -1.9%

Eu-153 0.1% 5.3% 6.1% 

Gd-155 4.4% 10.7% 15.9% 
B7

Table B.1 Comparisons of Polaris Calculations and Measurements for Burnup Credit Isotopes (C/M – 1)

Nuclide

FOR-3|

GN592|

J8|ENUSA2

FOR-3|

GN592|

J8|ENUSA-3

FOR-3|

GN592|

J8|ENUSA-4

FOR-3|

GN592|

J8|ENUSA-5

FOR-3|

GN592|

J8|ENUSA-6

FOR-3|

GN592|

J8|ENUSA-7

FOR-3|

GN592|

J8|ENUSA-8

LBT|

AIA003|

KLU1

U-234 3.3% 1.5% 5.8% 3.3% 1.5% 4.3% 6.7% 6.9% 

U-235 -8.7% -6.5% -12.5% -8.7% -4.7% -4.0% 0.8% 2.0% 

U-236 -4.2% -5.2% 15.7% -2.6% -3.2% -2.5% -0.5% -4.1% 

U-238 -0.5% -0.7% -0.4% -0.2% -0.6% -0.8% -0.3% -0.3% 

Np-237 -1.1% 

Pu-238 3.4% -1.9% -7.2% -5.7% 6.9% 9.2% -10.4% 33.6% 

Pu-239 -11.6% -4.6% -15.6% -10.3% -3.4% -3.8% -5.3% 10.5% 

Pu-240 -2.4% -1.5% -3.5% -1.9% -1.4% -0.5% -1.9% 8.1% 

Pu-241 -9.6% -3.8% -13.9% -8.5% -2.2% -3.0% -5.6% 9.3% 

Pu-242 3.3% 0.9% 1.3% 2.7% 1.5% 3.4% -0.6% 12.3% 

Am-241 -5.0% 2.5% -8.6% 1.1% 5.0% 3.7% -0.9% 10.5% 

Am-243 -17.1% -7.4% -17.3% -5.0% -14.3% -5.8% -31.8% 4.6% 

Mo-95 1.9% -0.8% 13.4% -3.0% 14.5% 5.0% -1.5% 17.6%

Tc-99 25.0% 

Ru-101

Rh-103 20.2%

Ag-109 29.4%

Cs-133 -24.0% -4.5% -2.8% -4.5% -6.7% -3.5% -5.4% 3.6% 

Nd-143 6.2% 6.9% 4.6% 7.6% 8.2% 12.1% 7.0% 2.2% 

Nd-145 3.5% 3.1% 4.7% 4.1% 0.0% 7.9% 4.7% -0.6% 

Sm-147 6.8% 9.1% 17.0% -16.7% -15.6% 13.5% -17.0% 2.9% 

Sm-149 -25.4% -13.8% -34.0% -18.5% -6.9% -18.3% -10.1%

Sm-150 2.7% 6.3% 7.0% 7.5% 9.6% 8.5% 7.7% 2.3% 

Sm-151 -18.2% 0.5% -10.9% 12.1% 23.8% 19.0% -0.2% -2.7% 

Sm-152 3.9% 0.9% 9.8% 7.3% 2.5% 4.2% 5.1% 2.3% 

Eu-151 -32.7%

Eu-153 9.2% 5.7% 4.7% -3.2% 6.0% 5.7% 5.2% 0.6% 

Gd-155 24.5% 13.1% 9.4% 32.2% 20.6% 30.7% 14.4% 9.4% 
B8

Table B.1 Comparisons of Polaris Calculations and Measurements for Burnup Credit Isotopes (C/M – 1)

Nuclide

LIM-1|

FYJ1433|

D9|1D2

LIM-1|

FYJ1433|

D9|4G1

LIM-1|

FYJ1433|

D9|4D4

LIM-1|

FYJ1433|

D9|2D2

LIM-1|

FYJ1433|

H5|3A1C

LIM-1|

FYJ1433|

H5|3A1G

LIM-

FYJ14

D8|4

U-234 1.3% 0.9% 3.2% 2.3% 4.3% 4.3% 

U-235 3.0% 3.8% -2.9% 2.4% 4.3% 2.5% 

U-236 3.5% 3.9% 4.2% 4.3% 3.0% 3.0% 

U-238 -0.6% -0.6% -0.7% -0.7% -0.5% -0.5% 

Np-237 -6.9% -6.8% -5.7% -8.7% -5.2% -5.8% 

Pu-238 21.9% 16.0% 16.6% 21.1% 14.9% 14.6% 

Pu-239 0.7% 3.0% -2.9% 1.4% -0.7% -1.3% 

Pu-240 0.6% 0.2% -1.6% -0.5% -0.8% -1.1% 

Pu-241 2.2% 4.2% -0.2% 3.0% 3.5% 3.5% 

Pu-242 9.2% 6.7% 11.4% 10.2% 12.9% 15.1% 

Am-241 5.1% 9.8% 4.2% 15.2% 10.1% 8.6% 

Am-243 2.8% -0.7% 0.1% 9.4% 3.2% 2.8% -

Mo-95 -0.4% 3.6% 6.9% 4.4% 3.8% 3.8% -

Tc-99 9.7% 21.5% 12.8% 16.4% 18.1% 21.3% 

Ru-101 3.6% 5.5% 9.5% 7.4% 8.3% 5.7% 

Rh-103 5.0% 2.3% 2.1% 7.1% 3.6% -0.3% 

Ag-109 12.7% 23.7% 16.1% 18.5% 17.8% 4.2% -

Cs-133

Nd-143 11.1% 10.5% 9.2% 11.0% 9.8% 9.8% 

Nd-145 8.1% 7.7% 9.0% 11.8% 8.2% 8.2% 

Sm-147 7.6% 3.1% 7.7% 5.6% 6.0% 5.6% 

Sm-149 20.2% -6.6% -8.0% 3.0% -3.0% -3.4% 

Sm-150 8.3% 4.8% 8.7% 7.0% 5.0% 5.6% 

Sm-151 5.7% 2.8% 1.7% 4.9% 0.4% 1.1% 

Sm-152 10.2% 6.4% 13.6% 9.2% 12.2% 13.0% 

Eu-151 7.6% 6.8% 8.9% 11.7% 4.8% 3.2% 

Eu-153 6.5% 4.1% 10.3% 7.6% 9.2% 9.2% 

Gd-155 10.2% 8.7% 5.3% 9.2% 3.9% 10.4% 

a Sample measured by PSI

b Sample measured by SCK•CEN
APPENDIX C POLARIS MODEL INPUT FILES

C-1

Table C.1 Polaris Input File for the Dodewaard Assembly Y013 (6 × 6)

=polaris

title "Dodeward DU1 model "

lib "xn56v7.1" 

sys BWR 

geom FuelNode : ASSM 6 1.793 

channel COOL 

hGAP 0.476 0.872 : MOD.1 MOD.1 

box hspan=5.505 rad=0.9652 thick=0.17

opt KEFF UpscatterSuperGroup=1 

shield ALL=N FUEL=P GAD=R

basis ALL=no FUEL.100=yes

deplete FUEL=yes GAD=yes

%A

mat FUEL.1 : uo2_A temp=900 dens=10.412

comp uo2_A : UOX 3.2

%X

mat FUEL.100 : uo2_du1 temp=900 dens=10.462306

comp uo2_du1 : UOX 4.941

%B

mat GAD.1 : uo2gd_B temp=900 dens=10.412

comp uox_B : UOX 3.2

comp uo2gd_B : WT GD2O3=2.653 uox_B=-100

%C

mat FUEL.2 : uo2_C temp=900 dens=10.412

comp uo2_C : UOX 2.6

%D

mat FUEL.3 : uo2_D temp=900 dens=10.412

comp uo2_D : UOX 1.8

%mat FUEL.3 : uo2_A temp=900 dens=10.412

% comp uo2_E : UOX 3.2

mat FUEL.4 : mox_M temp=900 dens=10.34003664

comp uvec : WT scale=PCT

U234=0.003

U235=0.239

U235=0.001

U238=-100 

comp puvec : WT scale=PCT

PU238=1.405

PU239=61.844

PU240=23.380

PU241=8.826

PU242=4.545

comp uo2_mox : FORM uvec=1 O=2

comp puo2_mox : FORM puvec=1 O=2

comp mox_M : WT scale=PCT

 uo2_mox = 93.50514401

puo2_mox = 6.426611176

am241 = 0.068244814

mat GAP.1 : he_GAP temp=564

comp he_GAP : CONC

2003=1.50456E-11

APPENDIX C POLARIS MODEL INPUT FILES 
C-2

 2004=1.50456E-05

mat COOL.1 : H2O temp=564 

% void=32

mat MOD.1 : H2O temp=564 

% void=0

pin X : 0.5176 0.5275 0.6135 : FUEL.100 GAP.1 CLAD.1

pin A : 0.574 0.5855 0.6745 : FUEL.1 GAP.1 CLAD.1

pin B : 0.574 0.5855 0.6745 : GAD.1 GAP.1 CLAD.1

pin C : 0.574 0.5855 0.6745 : FUEL.2 GAP.1 CLAD.1

pin D : 0.574 0.5855 0.6745 : FUEL.3 GAP.1 CLAD.1

pin E : 0.574 0.5855 0.6745 : FUEL.1 GAP.1 CLAD.1

pin M : 0.5178 0.5275 0.6135 : FUEL.4 GAP.1 CLAD.1

pin G : 0.5855 0.6745 : MOD.1 CAN.1

pinmap D C E E A C 

 C X B A A A

 E B A A B E

 E A A G M E

 A A B M B A

 C A E E A A

mesh COOL : nx=9 ny=9 ns=16 nr=2

%<specialfield2>

mesh MOD : nf=4 nd=5 nx=10 ny=10 ns=16 nr=4

mesh FUEL : nr=3 ns=16

mesh GAD : nr=7 ns=16

%mesh CLAD : ns=16

mesh TUBE : ns=16

state

ALL : temp=564

FUEL : temp=900

MOD : void=0 temp=564

COOL : void=32 temp=564

read history 

%power is reduced by 1.02 void is reduced by 1.0

%cycle 19

STATE COOL : void = 53.1 52.4 52.1 51.9 52.7 52.7 52.3 52.0 52.8 53.3 53.8 52.9 53.6 53.8 53.5 

53.4 53.5 53.5 54.1 53.5 52.7 52.1 52.1

 FUEL : temp = 924.7 955.3 963.3 988.6 975.6 979.8 1011.7 1016.8 1019.0 1024.1 1026.8 1057.3 

1065.9 1080.8 1101.9 1107.6 1115.6 1110.2 1094.6 1069.6 1050.2 1011.9 1011.9

 GAD : temp = 924.7 955.3 963.3 988.6 975.6 979.8 1011.7 1016.8 1019.0 1024.1 1026.8 1057.3 

1065.9 1080.8 1101.9 1107.6 1115.6 1110.2 1094.6 1069.6 1050.2 1011.9 1011.9

power 28.336 29.845 30.07 31.324 30.325 30.335 32.518 20.706 39.29 33.007 33.181 35.404 36.057 

37.046 38.546 38.964 39.464 39.148 37.913 35.751 33.701 32.426 0.0

dt 6.3 12.03 21.54 20.87 7.3 15.67 11.48 7.29 13.55 14.09 9.9 17.89 17.71 20.9 23.47 22.07 

20.03 20.74 10.6 16.89 6.75 11.04 73.0

%cycle 20

STATE COOL : void = 23.9 88.0 53.15 53.19 52.80 52.00 51.37 50.55 50.55

FUEL : temp = 1016.8 1080.8 1051.78 1046.44 1052.33 1066.48 1077.01 1037.75 1037.75

 GAD : temp = 1016.8 1080.8 1051.78 1046.44 1052.33 1066.48 1077.01 1037.75 1037.75

power 29.876 36.873 33.446 33.007 34.058 35.873 37.322 33.028 0.0

dt 4.26 13.32 64.65 48.6 27.1 29.8 55.64 44.63 50.0

%cycle 21

STATE COOL : void = 48.83 58.09 57.03 55.41 54.64 53.30 50.07 50.36 52.70 52.70
C-3

 FUEL : temp = 1051.39 1128.78 1090.98 1053.75 1027.34 1043.67 1064.28 1091.97 1027.44 

1027.44

 GAD : temp = 1051.39 1128.78 1090.98 1053.75 1027.34 1043.67 1064.28 1091.97 1027.44

1027.44

power 31.232 42.707 40.361 37.556 34.966 37.108 39.148 41.412 36.128 0.0

dt 8.1 32.84 68.82 41.91 28.97 19.9 63.14 27.14 25.07 54.0

%cycle 22

STATE COOL : void = 48.78 55.20 50.79 45.15 39.90 46.43 47.23 48.58 48.58

FUEL : temp = 1099.88 1114.99 1145.96 1132.02 1075.93 1131.26 1065.71 972.45 972.45

 GAD : temp = 1099.88 1114.99 1145.96 1132.02 1075.93 1131.26 1065.71 972.45 972.45

power 36.995 42.799 44.431 43.432 38.383 41.116 37.189 30.733 0.0

dt 20.43 19.64 45.71 53.76 47.67 64.08 33.87 32.13 39.0

%cycle 23

STATE COOL : void = 48.67 45.94 39.59 36.64 36.13 38.36 42.98 46.51

FUEL : temp = 976.53 992.19 977.68 931.56 922.97 938.25 909.21 814.07

GAD : temp = 976.53 992.19 977.68 931.56 922.97 938.25 909.21 814.07

power 30.253 31.079 29.876 27.234 26.571 27.173 25.469 23.093 

dt 48.95 57.93 64.76 48.5 36.85 31.23 24.42 11.35

STATE COOL : void = 46.51

power 0 0 0 0 0 0 0 0 0 0 0 0

dt 2295 25 11 77 24 9 4 19 54 12 10 19 9

end history

end
C-4

Table C.2 Polaris Input File for the Forsmark 3 Assembly 14595 (SVEA100) 

=polaris

title "SVEA-96 Optima2"

lib "xn56v7.1" 

sys BWR 

geom FuelNode : ASSM 10 1.24

channel COOL 

hgap 0.6775 0.7775 : MOD.1 MOD.1 

box hspan=6.867 rad=1.01 thick=0.14 

shield ALL=N FUEL=P GAD=R

mat CAN.1 : ZIRC2 temp=560

mat CLAD.1 : ZIRC2 temp=560

mat FUEL.9 : uo2_f9 10.62 temp=800

 comp uo2_f9 : UOX 4.0

mat FUEL.10 : uo2_f10 10.62 temp=800

 comp uo2_f10 : UOX 3.6

mat FUEL.11 : uo2_f11 10.62 temp=800

 comp uo2_f11 : UOX 3.6

mat FUEL.12 : uo2_f12 10.62 temp=800

 comp uo2_f12 : UOX 1.98

mat FUEL.13 : uo2_f13 10.62 temp=800

 comp uo2_f13 : UOX 2.22

mat FUEL.14 : uo2_f14 10.62 temp=800

 comp uo2_f14 : UOX 2.81

mat FUEL.15 : uo2_f15 10.62 temp=800

 comp uo2_f15 : UOX 3.17 

mat FUEL.16 : uo2_f16 10.62 temp=800

 comp uo2_f16 : UOX 3.6

mat FUEL.17 : uo2_f17 10.62 temp=800

 comp uo2_f17 : UOX 4.0

mat FUEL.100 : uo2_f17 10.62 temp=800

mat GAD.18 : uo2gd_f315 temp=800 dens=10.37

comp uox317 : UOX 3.17

comp uo2gd_f315 : WT GD2O3=3.15 uox317=-100 

mat gap.1 : he_gap temp=560

comp he_gap : CONC

2003=1.50456E-11

 2004=1.50456E-05

mat COOL.1 : H2O void=20 temp=560

mat MOD.1 : H2O void=0 temp=560

mat COOL.2 : H2O void=20 temp=560

pin W1 : 0.65 : COOL.1 : SQR

pin 9 : 0.4095 0.418 0.481 : FUEL.9 gap.1 clad.1

pin 10 : 0.4095 0.418 0.481 : FUEL.10 gap.1 clad.1

pin 11 : 0.4095 0.418 0.481 : FUEL.11 gap.1 clad.1

pin 12 : 0.4095 0.418 0.481 : FUEL.12 gap.1 clad.1

pin 13 : 0.4095 0.418 0.481 : FUEL.13 gap.1 clad.1

pin 14 : 0.4095 0.418 0.481 : FUEL.14 gap.1 clad.1

pin 15 : 0.4095 0.418 0.481 : FUEL.15 gap.1 clad.1
C-5

pin 16 : 0.4095 0.418 0.481 : FUEL.16 gap.1 clad.1

pin 17 : 0.4095 0.418 0.481 : FUEL.17 gap.1 clad.1

pin 18 : 0.4095 0.418 0.481 : GAD.18 gap.1 clad.1

pin A : 0.4095 0.418 0.481 : FUEL.100 gap.1 clad.1

pinmap 

12 

13 14 

14 18 10 

14 11 17 17 

14 16 17 17 A 

14 16 17 17 17 17 

14 11 17 17 17 17 17 

14 16 9 17 17 17 17 9 

14 18 16 11 16 16 11 16 18 

13 14 15 15 15 15 15 15 14 13

%0.242 0.689 0.931 4.077 4.505 5.185 5.711 6.192

cross 0.456 0.08 : 0 0.242 0.689 0.931 4.077 4.505 5.185 5.711 5.883 6.273 6.403 6.87 

 7.337 7.467 7.857 8.029 8.555 9.235 9.663 12.809 13.051 13.498 13.74

: 0.78 0.00 0.00 0.2 0.2 0.2 0.2 0.2 0.2 0.3 0.4 0.912

0.4 0.3 0.2 0.2 0.2 0.2 0.2 0.2 0.00 0.00 0.78 

: -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0

-0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0

: 1 1 1 8 2 2 2 1 2 1 2 

 2 1 2 1 2 2 2 8 1 1 1

dymap

-0.0

0.0 -0.0

0.0 0.0 0.0 

0.0 0.0 0.0 0.0 

0.0 0.0 0.0 0.0 0 

-0.0 -0.0 -0.0 -0.0 0 0

-0.0 -0.0 -0.0 -0.0 0 0 -0.0

-0.0 -0.0 -0.0 -0.0 0 0 -0.0 -0.0

-0.0 -0.0 -0.0 -0.0 0 0 -0.0 -0.0 -0.0

0.0 0.0 0.0 0.0 0 0 0.0 0.0 0.0 0.00

dxmap 

-0.0

0.0 -0.0

0.0 0.0 0.0 

0.0 0.0 0.0 0.0 

0.0 0.0 0.0 0.0 0

-0.0 -0.0 -0.0 -0.0 0 0

-0.0 -0.0 -0.0 -0.0 0 0 -0.0

-0.0 -0.0 -0.0 -0.0 0 0 -0.0 -0.0

-0.0 -0.0 -0.0 -0.0 0 0 -0.0 -0.0 -0.0

0.0 0.0 0.0 0.0 0 0 0.0 0.0 0.0 0.00

mesh COOL : nx=9 ny=9 ns=16 nr=2

mesh COOL.2 : nx=9 ny=9 ns=16 nr=4

mesh MOD : nf=4 nd=5 nx=10 ny=10 ns=16 nr=4

mesh MOD.2 : ns=4 nr=1

mesh FUEL : nr=3 ns=16

mesh GAD : nr=7 ns=16
C-6

mesh TUBE : ns=16

DEPLETE FUEL=yes GAD=yes

BASIS ALL=no FUEL.100=yes

read history

STATE COOL : VOID= 67.27 69.24 68.81 73.21 68.97 69.16 69.36 69.46 69.25 70.22 71.36 64.81 60.05 

38.34 

 FUEL : TEMP= 852.30 835.90 838.63 745.35 852.24 856.41 858.87 846.00 865.16 887.47 878.10 

862.32 828.09 734.56 

 GAD : TEMP= 852.30 835.90 838.63 745.35 852.24 856.41 858.87 846.00 865.16 887.47 878.10 

862.32 828.09 734.56 

power 34.60 32.53 32.65 21.41 33.88 34.18 34.30 32.77 34.98 37.68 36.59 34.92 31.04 20.36 

dt 5.25 8.13 23.72 37.92 26.56 25.76 27.38 6.91 26.72 27.28 27.69 27.46 27.47 79.75 

power 0.00 

dt 21.00 

STATE COOL : VOID= 65.78 64.20 66.85 67.52 65.38 63.62 65.19 62.07 64.85 66.90 64.98 57.40 

 FUEL : TEMP= 818.30 852.05 880.64 879.15 880.96 905.63 882.96 899.29 889.04 834.89 814.62 

703.20 

 GAD : TEMP= 818.30 852.05 880.64 879.15 880.96 905.63 882.96 899.29 889.04 834.89 814.62 

703.20 

power 30.26 34.22 37.26 37.13 37.42 40.35 37.73 39.80 38.53 32.16 29.84 16.92 

dt 6.85 8.35 28.02 22.19 18.59 8.15 32.38 12.13 24.49 39.12 29.07 50.65 

power 0.00 

dt 28.00 

STATE COOL : VOID= 62.23 62.72 62.14 61.46 61.03 59.13 58.42 57.66 57.34 58.12 51.44 59.07 52.19 

 FUEL : TEMP= 864.39 865.90 864.45 861.75 856.24 847.26 830.67 770.58 790.46 678.03 779.96 

805.88 667.08 

 GAD : TEMP= 864.39 865.90 864.45 861.75 856.24 847.26 830.67 770.58 790.46 678.03 779.96 

805.88 667.08 

power 36.92 37.09 36.90 36.57 35.90 34.82 32.82 25.49 27.95 14.12 26.75 29.53 12.88 

dt 38.96 25.85 30.16 28.43 24.95 35.79 28.54 36.22 12.32 36.29 18.57 31.61 23.32 

power 0.00 

dt 18.00 

STATE COOL : VOID= 59.07 59.45 59.89 59.24 59.22 60.02 58.99 57.81 58.01 58.74 58.19 54.57 51.22 

46.62 40.41 36.92 

 FUEL : TEMP= 757.10 827.04 839.57 832.60 826.96 829.23 835.94 847.86 847.96 835.43 832.85 

801.70 752.00 739.79 720.79 711.09 

 GAD : TEMP= 757.10 827.04 839.57 832.60 826.96 829.23 835.94 847.86 847.96 835.43 832.85 

801.70 752.00 739.79 720.79 711.09 

power 24.01 32.39 33.93 33.12 32.44 32.68 33.58 35.16 35.21 33.66 33.35 29.62 23.53 22.04 19.70 

18.53 

dt 7.16 27.00 27.22 26.07 26.88 22.33 29.19 27.13 21.17 28.15 5.72 29.56 31.37 32.35 22.91 16.78 

power 0.00 

dt 18.00 

STATE COOL : VOID= 56.88 59.31 59.24 58.21 57.49 56.71 55.88 55.31 54.52 56.19 52.00 

FUEL : TEMP= 831.33 815.79 808.05 803.02 800.52 799.59 799.49 812.47 814.44 821.55 746.62 

 GAD : TEMP= 831.33 815.79 808.05 803.02 800.52 799.59 799.49 812.47 814.44 821.55 746.62 

power 33.14 31.22 30.31 29.67 29.28 29.08 28.94 30.34 30.54 31.13 22.38 

dt 10.05 14.96 19.43 26.51 26.68 26.58 31.41 16.28 6.31 25.04 50.74 

power 0.00 

dt 34.00 

STATE COOL : VOID= 45.64 46.01 48.17 46.18 44.72 44.74 44.48 43.54 41.76 40.07 40.67 35.75 31.01 

FUEL : TEMP= 719.84 691.84 665.55 704.55 712.11 710.45 707.85 705.35 706.64 719.87 714.22 

677.15 693.47 

 GAD : TEMP= 719.84 691.84 665.55 704.55 712.11 710.45 707.85 705.35 706.64 719.87 714.22 

677.15 693.47 

power 18.60 15.37 12.34 16.75 17.56 17.32 16.98 16.66 16.76 18.17 17.48 13.45 15.22 

dt 6.42 31.17 26.89 19.04 29.03 28.81 29.73 29.53 30.12 30.28 29.88 35.00 27.11 

power 0.00 

dt 3775.00 
C-7

power 0.00 

dt 12.00 

power 0.00 

dt 15.00 

end history

end
C-8

Table C.3 Polaris Input File for Forsmark 3 Assembly GN592 (GE14 Dominant Lattice)

=polaris

title "FORSMARK N3 DOM"

lib "xn56v7.1"

sys BWR 

geom FuelNode : ASSM 10 [1.2954 ] 

channel COOL 

hgap [0.688 1.051] : MOD.1 MOD.1

box [ 0.165 1.143 6.703 : 0.089 : 4.87 : 4.87] 

mat CAN.1 : ZIRC2 temp=560 

mat CLAD.1 : ZIRC2 temp=560

[mat FUEL.1 : uo2_f1 temp=792 dens= 10.503

comp uo2_f1 : CONC

8016=4.68538E-02

92234=3.39244E-06

92235=3.79547E-04

92236=1.73851E-06

92238=2.30422E-02

mat GAD.10 : uo2gd_f10 temp=792 dens= 10.373

comp uo2gd_f10 : CONC

 8016=4.65010E-02

64152=2.75728E-06

64154=3.00543E-05

64155=2.04039E-04

64156=2.82207E-04

64157=2.15757E-04

64158=3.42454E-04

64160=3.01370E-04

92234=7.23675E-06

92235=8.09649E-04

92236=3.70858E-06

92238=2.13959E-02

mat GAD.11 : uo2gd_f10 temp=792 dens= 10.373

mat FUEL.2 : uo2_f2 temp=792 dens= 10.503

comp uo2_f2 : CONC

8016=4.68559E-02

92234=4.24052E-06

92235=4.74430E-04

92236=2.17312E-06

92238=2.29471E-02

mat FUEL.3 : uo2_f3 temp=792 dens= 10.503

comp uo2_f3 : CONC

8016=4.68602E-02

92234=5.93666E-06

92235=6.64194E-04

92236=3.04233E-06

92238=2.27569E-02

mat FUEL.4 : uo2_f4 temp=792 dens= 10.503

comp uo2_f4 : CONC

 8016=4.68624E-02

92234=6.78471E-06
C-9

92235=7.59075E-04

92236=3.47693E-06

92238=2.26619E-02

mat FUEL.5 : uo2_f5 temp=792 dens= 10.503

comp uo2_f5 : CONC

8016=4.68581E-02

92234=5.08860E-06

92235=5.69313E-04

92236=2.60773E-06

92238=2.28520E-02

mat FUEL.6 : uo2_f6 temp=792 dens= 10.503

comp uo2_f6 : CONC

8016=4.68645E-02

92234=7.63276E-06

92235=8.53954E-04

92236=3.91152E-06

92238=2.25668E-02

mat FUEL.7 : uo2_f7 temp=792 dens= 10.503

comp uo2_f7 : CONC

8016=4.68664E-02

92234=8.37478E-06

92235=9.36972E-04

92236=4.29178E-06

92238=2.24836E-02

mat FUEL.8 : uo2_f7 temp=792 dens= 10.503

mat FUEL.9 : uo2_f7 temp=792 dens= 10.503 

mat FUEL.100 : uo2_f7 temp=792 dens= 10.503] 

mat GAP.1 : FILLGAS temp=560

mat COOL.1 : H2O temp=560

mat COOL.2 : H2O temp=560

% comp h2o_mod : CONC

% 1001=3.06145E-02

% 8016=1.53073E-02

mat MOD.1 : H2O temp=560 

% comp h2o_mod1 : CONC

% 1001=4.93677E-02

% 8016=2.46838E-02

pin 1 : 0.43815 0.44704 0.51308 

: FUEL.1 GAP.1 CLAD.1

pin 2 : 0.43815 0.44704 0.51308 

: FUEL.2 GAP.1 CLAD.1

pin 3 : 0.43815 0.44704 0.51308 

: FUEL.3 GAP.1 CLAD.1

pin 4 : 0.43815 0.44704 0.51308 

: FUEL.4 GAP.1 CLAD.1

pin 5 : 0.43815 0.44704 0.51308 

: FUEL.5 GAP.1 CLAD.1

pin 6 : 0.43815 0.44704 0.51308 

: FUEL.6 GAP.1 CLAD.1

pin 7 : 0.43815 0.44704 0.51308 

: FUEL.7 GAP.1 CLAD.1
C-10

pin 8 : 0.43815 0.44704 0.51308 

: FUEL.8 GAP.1 CLAD.1

pin 9 : 0.43815 0.44704 0.51308 

: FUEL.9 GAP.1 CLAD.1

pin A : 0.43815 0.44704 0.51308 

: FUEL.100 GAP.1 CLAD.1

pin 10 : 0.43815 0.44704 0.51308 

: GAD.10 GAP.1 CLAD.1

pin 11 : 0.43815 0.44704 0.51308 

: GAD.11 GAP.1 CLAD.1

pin V : 0.51308 : COOL.1

pin W1 size=2 :[ 1.16840 1.24460 ] 

: MOD.1 CAN.1

deplete FUEL=yes GAD=yes

basis ALL=no FUEL.100=yes

shield ALL=N FUEL=P GAD=R 

opt KEFF UpscatterSuperGroup=1

pinmap

1 

2 V

3 6 10

4 V 7 7

4 10 7 7 V

4 6 7 W1 W1 V

4 V 7 W1 W1 7 7

4 11 7 7 7 7 7 7

3 V 11 V 7 7 V 10 V

2 3 6 6 7 7 7 A 4 5

mesh COOL.1 nr=2 ns=16 

mesh COOL.2 nr=3 ns=16 

mesh MOD : nf=3 nd=4 nx=6 ny=6 ns=16 nr=4

mesh FUEL nr=3 ns=16

mesh GAD nr=7 ns=16

state ALL : temp=560

FUEL : temp=792

GAD : temp=792

MOD : void=0

COOL : void=40

%------------------------------------------------------------------------------% 

% operating history provided by RCA benchmark

%------------------------------------------------------------------------------% 

read history

%power is reduced by 0.91 void is reduced by 1.0

%----------------% 

% cycle 16 

%----------------% 

STATE COOL : void = 72.3 71.0 71.9 72.2 72.5 73.0 72.8 72.0 71.4 69.5 67.4 65.3 61.5 55.2
C-11

power 4.45 32.778 38.62 42.515 38.029 40.832 43.453 42.752 46.064 44.19 45.8 38.73 35.263 

35.617

dt 24.24 19.76 22.36 25.4 28.39 34.38 24.84 27.78 28.13 29.32 16.5 13.94 27.56 39.41

power 0.0

dt 40.0

%----------------% 

% cycle 17 

%----------------% 

STATE COOL : void = 68.3 69.6 67.0 65.0 65.6 66.0 65.6 61.5 62.2 50.0

power 34.116 44.772 43.498 44.863 38.056 42.015 39.485 30.967 23.969 23.614

dt 9.49 28.94 27.3 28.88 28.37 28.27 27.34 27.89 4.5 64.01

power 0.0

dt 14.0

%----------------% 

% cycle 18 

%----------------% 

STATE COOL : void = 78.1 76.8 75.6 75.0 74.1 73.5 72.9 71.8 70.3 67.9 65.6 70.4 70.4 53.8

power 29.047 38.22 38.329 45.655 37.137 40.55 39.312 43.998 41.96 40.249 45.573 54.109 38.866 

39.694

dt 3.72 8.47 28.17 26.01 29.07 26.63 30.21 24.54 28.3 29.51 26.06 27.93 2.78 81.6

power 0.0

dt 12.0

%----------------% 

% cycle 19 

%----------------% 

STATE COOL : void = 53.8 53.6 52.9 51.6 51.0 49.4 48.8 47.2 45.9 40.6 38.0 31.9 30.1 19.2 13.4

power 13.659 19.893 26.053 19.064 18.71 18.118 17.881 21.904 21.458 18.946 19.838 22.377 18.791 

17.681 14.842

dt 31.62 32.56 16.58 28.32 23.08 29.8 12.08 24.64 25.15 28.49 27.22 28.95 22.99 24.42 29.1

power 0.0

dt 35.0

%----------------% 

% cycle 20 

%----------------% 

STATE COOL : void = 44.0 43.6 42.2 41.1 40.4 38.4 36.5 34.5 31.2 25.8 19.0 18.8

power 15.334 15.006 18.491 19.292 14.251 13.932 23.032 13.495 21.057 15.261 11.339 8.59

dt 7.04 21.58 29.2 22.39 30.3 31.01 23.44 32.0 25.64 28.3 9.52 12.57

power 0.0 0.0 0.0 0.0 0.0 0.0

dt 1479.0 1.0 15.0 61.0 58.0 85.0

end history 

end
C-12

Table C.4 Polaris Input File for the Fukushima Daini 1 Assembly 2F1ZN2 (ATRIUM9) 

=polaris

title "Atrium-9 model " 

lib "xn56v7.1"

sys BWR 

geom FuelNode : ASSM 9 1.45 

channel COOL 

hgap 0.65 : MOD.1 

box 0.25 0.87376 6.7 

opt KEFF UpscatterSuperGroup=1 

shield ALL=N FUEL=P GAD=P

deplete FUEL=YES GAD=YES

opt CRITSPEC Method="P1"

mat FUEL.1 : uo2_f100 temp=900 dens=10.63

 comp uo2_f100 : UOX 4.9

mat FUEL.2 : uo2_f102 temp=900 dens=10.63

 comp uo2_f102 :UOX 4.2

mat FUEL.21 : uo2_f102 temp=900 dens=10.63

mat FUEL.22 : uo2_f102 temp=900 dens=10.63

mat FUEL.3 : uo2_f103 temp=900 dens=10.63

 comp uo2_f103 : UOX 3.6 

mat FUEL.31 : uo2_f103 temp=900 dens=10.63

mat FUEL.4 : uo2_f104 temp=900 dens=10.63

 comp uo2_f104 : UOX 3.0

mat FUEL.41 : uo2_f104 temp=900 dens=10.63

mat FUEL.5 :uo2_f105 temp=900 dens=10.63

comp uo2_f105 : UOX 2.1

mat GAD.1 : uo2gd_f50 temp=900 dens=10.38

comp uox34 : UOX 3.0

comp uo2gd_f50 : WT GD2O3=5.0 uox34=-100 

mat GAP.1 : he_GAP temp=558.15

comp he_GAP : CONC

2003=1.50456E-11

2004=1.50456E-05

mat COOL.1 : H2O temp=558.15 :void=14

mat MOD.1 : H2O temp=558.15 :void=0

mat FUEL.100 : uo2_f100 temp=900 dens=10.63

mat GAD.100 : uo2gd_f50 temp=900 dens=10.38

pin C2 : 0.47000 0.49000 0.55000 

: GAD.100 gap.1 clad.1 

pin C3 : 0.47000 0.49000 0.55000 

: FUEL.100 gap.1 clad.1 

pin 1 : 0.47000 0.49000 0.55000 

: FUEL.1 gap.1 clad.1

pin 2 : 0.47000 0.49000 0.55000 

: FUEL.2 gap.1 clad.1

pin 3 : 0.47000 0.49000 0.55000 

: FUEL.3 gap.1 clad.1

pin 4 : 0.47000 0.49000 0.55000 
C-13

: FUEL.4 gap.1 clad.1

pin 5 : 0.47000 0.49000 0.55000 

: FUEL.5 gap.1 clad.1

pin 9 : 0.47000 0.49000 0.55000 

: GAD.1 gap.1 clad.1

pin 21 : 0.47000 0.49000 0.55000 

: FUEL.21 gap.1 clad.1 

pin 22 : 0.47000 0.49000 0.55000 

: FUEL.22 gap.1 clad.1 

pin 41 : 0.47000 0.49000 0.55000 

: FUEL.41 gap.1 clad.1 

pin 31 : 0.47000 0.49000 0.55000 

: FUEL.31 gap.1 clad.1 

pin W 3.0 : 1.8425 1.925 : MOD.1 CAN.1 : SQR SQR

pinmap 5

4 3

3 C2 C3

2 22 21 W

2 9 21 W W

2 22 21 W W W

3 9 1 21 21 21 1

4 3 9 22 9 22 9 3

5 4 3 2 2 2 3 4 5

mesh MOD : nf=3 nd=4 nx=6 ny=6 ns=8 nr=4

mesh FUEL : nr=3 ns=8

mesh GAD : nr=5 ns=8

%mesh CLAD : ns=8

mesh TUBE : ns=8

state ALL : temp=559

FUEL : temp=900

GAD : temp=900

MOD : void=0

COOL :void=14

%------------------------------------------------------------------------------% 

% Irradiation Report of Fuel Samples of BWR 9x9-9 Fuel 

% Japan Nuclear Energy Safety Organizaion Toru YAMAMOTO (JNES 2014)

%------------------------------------------------------------------------------% 

basis ALL=no GAD.100=yes

read history

STATE COOL : void = 67.6 67.6 67.6 69.1 69.1 69.2 69.2 70.2 70.2 71.2 71.2 72.5 72.5 73.6 73.6 

73.6

power 17.019 18.907 18.907 19.584 19.584 20.291 20.291 20.412 20.412 20.584 20.584 20.432 

20.432 20.17 20.17 20.17

dt 4.0 16.0 16.0 17.5 17.5 21.0 21.0 16.0 16.0 15.5 15.5 17.5 17.5 17.5 17.5 12.0

power 0.0

dt 6.0

STATE COOL : void = 74.2 74.3 74.3 74.0 74.0 74.4 74.4 73.2 73.2 73.2

power 16.887 20.432 20.432 21.21 21.21 22.745 22.745 24.876 24.876 24.876

dt 7.0 15.5 15.5 16.0 16.0 16.5 16.5 18.5 18.5 18.0

power 0.0

dt 40.0

STATE COOL : void = 77.0 77.0

power 16.998 16.998
C-14

dt 8.0 26.0

power 0.0

dt 71.0

STATE COOL : void = 76.5 76.3 76.3 76.4 76.4 76.5 76.5 76.6 76.6 76.8 76.8 76.9 76.9 76.8 76.3 

76.5 76.5

power 61.246 25.28 25.28 25.583 25.583 25.886 25.886 25.704 25.704 25.523 25.523 25.25 25.25 

25.068 25.381 26.078 26.078

dt 5.0 12.0 12.0 16.0 16.0 15.5 15.5 16.0 16.0 19.0 19.0 16.0 16.0 31.0 28.0 24.0 28.0

power 0.0

dt 86.0

STATE COOL : void = 72.9 73.3 74.3 73.4 73.7 73.8 73.6 73.4 72.9 73.0 73.0 71.2 71.4 70.8 70.8 

68.8 68.8

power 24.947 24.402 24.816 24.886 24.573 24.977 25.149 25.068 25.119 25.644 25.644 25.725 

26.826 28.411 28.411 28.401 28.401

dt 6.0 31.0 32.0 17.0 32.0 31.0 32.0 35.0 31.0 21.0 21.0 32.0 28.0 19.0 19.0 25.0 11.0

power 0.0

dt 3537.0

power 0.0

dt 15.0

power 0.0 

dt 347.0

power 0.0

dt 3.0

power 0.0

dt 4.0

power 0.0

dt 347.0

end history

end
C-15

Table C.5 Polaris Input File for Fukushima Daini-2 Assembly 2F2DN23 (8 × 8-2)

=polaris

title "Fukushima Daini -2 Assembly model "

lib "xn56v7.1" 

sys BWR 

geom FuelNode : ASSM 8 1.625 

channel COOL 

hGAP 0.6629 0.6629 : MOD.1 MOD.1 

box hspan=6.70305 rad=0.9652 thick=0.2032 

opt KEFF UpscatterSuperGroup=1 

deplete ALL=false FUEL=true GAD=true

shield ALL=N FUEL=P GAD=R

basis ALL=no GAD.100=yes

mat FUEL.1 : uo2_f100 temp=900 dens=10.42

comp uvec100 : WT scale=PCT

 U234=0.04

U235=3.910

U238=96.05

comp uo2_f100 : FORM uvec100=1 O=2 

mat FUEL.2 : uo2_f102 temp=900 dens=10.42

comp uvec102 : WT scale=PCT

 U234=0.031

U235=3.448

U238=96.521

comp uo2_f102 : FORM uvec102=1 O=2 

mat FUEL.3 : uo2_f103 temp=900 dens=10.42

comp uvec103 : WT scale=PCT

 U234=0.03

U235=3.405

U238=96.565

comp uo2_f103 : FORM uvec103=1 O=2

mat FUEL.4 : uo2_f104 temp=900 dens=10.42

comp uvec104 : WT scale=PCT

 U234=0.026

U235=2.903

U238=97.071

comp uo2_f104 : FORM uvec104=1 O=2

mat FUEL.5 :uo2_f105 temp=900 dens=10.42

comp uvec105 : WT scale=PCT

 U234=0.018

U235=1.997

U238=97.985

comp uo2_f105 : FORM uvec105=1 O=2

mat GAD.6 : uo2gd_f50 temp=900 dens=10.23

comp uox341 : UOX 3.41

 comp uo2gd_f50 : WT GD2O3=4.5 uox341=-100 

mat GAD.100 : uo2gd_f50 temp=900 dens=10.23 

mat GAP.1 : he_GAP temp=559.15

comp he_GAP : CONC

2003=1.50456E-11
C-16

 2004=1.50456E-05

mat COOL.1 : H2O temp=559.15 :void=54.7

mat MOD.1 : H2O temp=559.15 :void=0

pin B : 0.51500 0.52700 0.61300 : GAD.100 GAP.1 CLAD.1

pin 1 : 0.51500 0.52700 0.61300 : FUEL.1 GAP.1 CLAD.1

pin 2 : 0.51500 0.52700 0.61300 : FUEL.2 GAP.1 CLAD.1

pin 3 : 0.51500 0.52700 0.61300 : FUEL.3 GAP.1 CLAD.1

pin 4 : 0.51500 0.52700 0.61300 : FUEL.4 GAP.1 CLAD.1

pin 5 : 0.51500 0.52700 0.61300 : FUEL.5 GAP.1 CLAD.1

pin 6 : 0.51500 0.52700 0.61300 : GAD.6 GAP.1 CLAD.1

pin W1 : 0.67380 0.75000 : MOD.1 CAN.1

pinmap 5

4 1

3 B 2

3 2 4 3

3 2 4 W1 3

3 6 2 4 4 2

4 1 6 2 2 6 1

5 4 3 3 3 3 4 5

mesh COOL : nx=9 ny=9 ns=16 nr=2

%<specialfield2>

mesh MOD : nf=4 nd=5 nx=10 ny=10 ns=16 nr=4

mesh FUEL : nr=3 ns=16

mesh GAD : nr=7 ns=16

%mesh CLAD : ns=16

mesh TUBE : ns=16

state

ALL : temp=559.15

FUEL : temp=900

GAD : temp=900 

MOD : temp=559.15 void=0

COOL : temp=559.15 void=54.7

read history

power 9.89 24.97 30.78 30.78 30.78 30.78 30.78 30.78 0.0 

dt 6 3 22 22 22 22 22 22 21 

power 11.15 26.85 26.85 26.85 26.85 26.85 26.85 26.85 26.85 31.09 0.0 

dt 5 30.5 30.5 30.5 30.5 30.5 30.5 30.5 30.5 8 117

power 11.15 26.85 26.85 26.85 26.85 26.85 26.85 26.85 26.85 26.85 26.85 0.0 

dt 5 31.7 31.7 31.7 31.7 31.7 31.7 31.7 31.7 31.7 31.7 9 

power 11.62 27.33 27.33 27.33 30.78 0.0

dt 4 24 24 24 10 81

power 12.72 28.42 28.42 28.42 28.42 28.42 28.42 28.42 28.42

dt 3 45.625 45.625 45.625 45.625 45.625 45.625 45.625 45.625

end

power 0 0

dt 500 1881 

end history
C-17

Table C.6 Polaris Input File for Fukushima Daini-2 Assembly 2F2D1 (8 × 8-4)

=polaris

title "Fukushima Daini -2 Assembly model B3"

lib "xn56v7.1" 

sys BWR 

geom FuelNode : ASSM 8 1.625 

channel COOL 

hGAP 0.6629 0.6629 : MOD.1 MOD.1 

box hspan=6.70305 rad=0.9652 thick=0.2032 

opt KEFF UpscatterSuperGroup=1 

shield ALL=N FUEL=P GAD=R

basis ALL=no GAD.100=yes

DEPLETE FUEL=yes GAD=yes

mat FUEL.1 : uo2_f100 temp=900 dens=10.61

comp uo2_f100 : UOX 4.5

mat FUEL.2 : uo2_f102 temp=900 dens=10.61

comp uo2_f102 :UOX 3.9

mat FUEL.3 : uo2_f103 temp=900 dens=10.61

comp uo2_f103 : UOX 3.4 

mat FUEL.4 : uo2_f104 temp=900 dens=10.61

comp uo2_f104 : UOX 2.9

mat FUEL.5 :uo2_f105 temp=900 dens=10.61

comp uo2_f105 : UOX 2.0

mat GAD.6 : uo2gd_f50 temp=900 dens=10.48

comp uox341 : UOX 3.41

comp uo2gd_f50 : WT GD2O3=4.5 uox341=-100 

mat GAD.100 : uo2gd_f50 temp=900 dens=10.48 

mat GAP.1 : he_GAP temp=559

comp he_GAP : CONC

2003=1.50456E-11

 2004=1.50456E-05

mat COOL.1 : H2O temp=559 :void=60.48

mat MOD.1 : H2O temp=559 :void=0

%<specialfield1>

%<specialfield3>

pin B : 0.519 0.529 0.615 :GAD.100 GAP.1 CLAD.1

pin 1 : 0.519 0.529 0.615 :FUEL.1 GAP.1 CLAD.1

pin 2 : 0.519 0.529 0.615 :FUEL.2 GAP.1 CLAD.1

pin 3 : 0.519 0.529 0.615 :FUEL.3 GAP.1 CLAD.1

pin 4 : 0.519 0.529 0.615 :FUEL.4 GAP.1 CLAD.1

pin 5 : 0.519 0.529 0.615 :FUEL.5 GAP.1 CLAD.1

pin G : 0.519 0.529 0.615 :GAD.6 GAP.1 CLAD.1

pin W 2 : 1.623 1.7 : MOD.1 CAN.1
C-18

pinmap 5

4 2

3 B 1

3 2 4 W

3 2 4 W W

3 G 1 4 4 1

4 2 G 2 2 G 2

5 4 3 3 3 3 4 5

mesh COOL : nx=9 ny=9 ns=16 nr=2

%<specialfield2>

mesh MOD : nf=4 nd=5 nx=10 ny=10 ns=16 nr=4

mesh FUEL : nr=3 ns=16

mesh GAD : nr=7 ns=16

%mesh CLAD : ns=16

mesh TUBE : ns=16

state

ALL : temp=559

FUEL : temp=900

GAD : temp=900

MOD : temp=559 void=0

COOL : temp=559 void=60.48

read history 

power 32.58 33.07 33.07 33.07 33.07 33.07 33.07 33.07 33.07 33.07 33.07 32.92 32.92 32.92 32.92 

32.92 32.92 32.60 32.60 32.60 32.60 32.60 32.60 32.61 32.61 32.61 32.61 32.61 32.61 32.61 32.61 

0.00

dt 4.6 14.0 14.0 14.0 14.0 14.0 14.0 14.0 14.0 14.0 14.0 12.6 12.6 12.6 12.6 12.6 12.6 13.8 13.8 

13.8 13.8 13.8 13.8 14.4 14.4 14.4 14.4 14.4 14.4 14.4 14.4 118.0

power 30.20 30.20 30.20 30.20 29.72 29.72 29.72 29.72 29.72 29.72 29.72 29.72 29.72 29.72 29.09 

29.09 29.09 29.09 29.09 29.09 29.09 29.09 29.09 28.93 28.93 28.93 28.93 28.93 28.93 28.93 28.93 

0.00

dt 13.3 13.3 13.3 13.3 11.5 11.5 11.5 11.5 11.5 11.5 11.5 11.5 11.5 11.5 14.4 14.4 14.4 14.4 

14.4 14.4 14.4 14.4 14.4 14.9 14.9 14.9 14.9 14.9 14.9 14.9 14.9 82.0

power 26.04 26.04 26.04 26.04 26.04 24.12 24.12 24.12 24.12 24.12 24.12 24.12 24.12 24.12 24.12 

22.67 22.67 22.67 22.67 22.67 22.67 0.00 0.00

dt 22.8 22.8 22.8 22.8 22.8 12.9 12.9 12.9 12.9 12.9 12.9 12.9 12.9 12.9 12.9 20.5 20.5 20.5 

20.5 20.5 20.5 990.0 60.0

end history

end
C-19

Table C.7 Polaris Input File for Leibstadt Assembly AIA003 (SVEA96 Cominant Lattice)

=polaris 

title "Leib N4 DOM"

lib "xn56v7.1"

sys BWR

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

% options

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

shield ALL=N FUEL=P GAD=R

opt KEFF UpscatterSuperGroup=1

basis ALL=no FUEL.9=yes

deplete FUEL=yes GAD=yes

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

% materials

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

[comp U35 : UOX 3.5

comp U39 : UOX 3.9

comp U47 : UOX 4.7

comp U42 : UOX 4.2

comp U21 : UOX 2.1

comp U40 : UOX 4.0

comp U32 : UOX 3.2

comp U32GD40 : WT GD2O3=4.0 U32 = -100

comp U40GD40 : WT GD2O3=4.0 U40 = -100

mat FUEL.1 : U21 10.52

mat FUEL.2 : U35 10.52

mat FUEL.3 : U39 10.52

mat GAD.4 : U40GD40 10.38

mat FUEL.5 : U42 10.52

mat FUEL.6 : U47 10.52

mat GAD.7 : U32GD40 10.38

mat FUEL.8 : U35 10.52

mat FUEL.9 : U39 10.52

] 

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

% geometry

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

geom myBWR : ASSM 10 [1.27] 

hgap 0.69 0.69 : MOD.1 MOD.1

%: MOD.2 MOD.1

% box thichness, radius, half width

box [ 0.14 1.13 6.79 ] 

pin 1 : [0.4095 0.418 0.481] : FUEL.1 GAP.1 CLAD.1

pin 2 : 0.4095 0.418 0.481 : FUEL.2 GAP.1 CLAD.1

pin 3 : 0.4095 0.418 0.481 : FUEL.3 GAP.1 CLAD.1

pin 4 : 0.4095 0.418 0.481 : GAD.4 GAP.1 CLAD.1

pin 5 : 0.4095 0.418 0.481 : FUEL.5 GAP.1 CLAD.1

pin 6 : 0.4095 0.418 0.481 : FUEL.6 GAP.1 CLAD.1 

pin 7 : [0.4385 0.447 0.515] : GAD.7 GAP.1 CLAD.1

pin 8 : [0.4385 0.447 0.515]: FUEL.8 GAP.1 CLAD.1

pin 9 : [0.4385 0.447 0.515]: FUEL.9 GAP.1 CLAD.1 

pin E : [0.63] : COOL.1 : SQR
C-20

mesh COOL.1 : nx=9 ny=9 nr=3

mesh FUEL : nr=2 ns=16

mesh GAD : nr=6 ns=16

mesh MOD.1 : nf=3 nd=4 nx=6 ny=6 ns=16 nr=4 

mesh MOD.2 : nf=3 nd=5 

pinmap

1 

2 2

3 4 6

3 5 6 7 

2 6 9 8 E 

2 6 9 8 E E 

3 5 6 7 8 8 7 

3 4 6 6 9 9 6 6

2 2 4 5 6 6 5 4 2 

1 2 3 3 2 2 3 3 2 1

%cross halfwidth=0.0

% 1a 1b 1c 2c 2b 2a halfwidth

cross [0.295 0.08 : 0.00 0.20 0.534 0.775 3.95 4.27 4.917 6.79 

8.663 9.3045 9.63 12.805 13.045 13.38 13.58 

: 0.4 0.0 0.00 0.12 0.12 0.0 0.0 1.873 

0.0 0.0 0.12 0.12 0.00 0.00 0.4 

: -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0

-0.0 -0.0 -0.0 -0.0 -0.0 -0.0 -0.0]

: 1 1 1 4 1 1 5 

 5 1 1 4 1 1 1

dxmap

-.[00

-.00 .0

-.00 .0 .0

-.00 .0 .0 .0

-.005 -.01 -.01 -.01 -.01] 

dymap

-.[00

-.00 .0

-.00 .0 .0

-.00 .0 .0 .0

-.005 -.005 -.005 -.005 .01] 

% state

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%% 

state ALL : temp=560

FUEL : temp=900

GAD : temp=900

MOD : void=0

COOL : void=10

read history

%cyc 15 

STATE COOL : void = 0.23 0.08 0.92 0.39 3.09 2.24 1.91 8.78 14.16 11.69 6.55 1.3 0.19 0 

COOL : temp=560 FUEL : temp=816 GAD : temp=816 

power 0 15.65 24.07 26.01 27.21 31.01 29.67 32.83 40.13 40.99 36.52 28.22 19.98 0 

dt 1 12 34 27 13 31 33 21 36 34 36 41 32 24 

%cyc16 

STATE COOL : void = 1.88 22.76 21.65 22.73 24.9 20.57 10.99 4.05 0.82 0 0 

COOL : temp=561.7 FUEL : temp=924.7 GAD : temp=924.7 
C-21

power 18.73 41.24 48.24 47.47 48.60 49.32 43.96 36.06 28.97 20.89 0.00 

dt 28 36 36 33 36 36 34 34 30 32 21 

%cyc17 

STATE COOL : void = 15.94 11.74 19.43 18.72 19.76 16.67 17.89 19.3 14.62 10.35 3.44 1.29 0 

COOL : temp=562 FUEL : temp=944 GAD : temp=944 

power 25.7 41.1 41.8 45.2 45.6 43.8 45.8 45.8 43.3 39.4 34.5 25.3 0.0 

dt 17 39 30 35 35 14 36 34 22 27 35 25 21 

%cyc18 

STATE COOL : void = 6.55 4.13 4.42 3.45 3.09 2.61 1.59 0.07 1.27 0 0 0 

COOL : temp=559 FUEL : temp=838 GAD : temp=838 

power 31.01 34.97 32.86 32.94 32.03 31.40 30.43 25.76 23.66 23.45 18.28 0 

dt 24 34 28 29 36 28 33 42 28 36 24 15 

%cyc19 

STATE COOL : void = 0 0 0 0 0 0 0 0 0 0 0 0 0 0

COOL : temp=554 FUEL : temp=731 GAD : temp=731 

power 18.10 22.19 20.85 22.35 22.28 22.92 23.06 22.59 19.81 15.84 12.82 12.13 10.08 0

dt 18 39 38 27 27 16 12 36 35 40 24 14 23 21

%cyc20 

STATE COOL : void = 0 0 0 0 0 0 0 0 0 0 0 0 

COOL : temp=554 FUEL : temp=695 GAD : temp=695 

power 11.05 15.21 14.40 13.16 12.68 12.36 11.81 10.69 8.47 8.52 6.43 0 

dt 16 22 36 40 36 38 39 36 23 32 32 42 

%cyc21 

STATE COOL : void = 0.07 0 0 0 0 0 0 0 

COOL : temp=556 FUEL : temp=687 GAD : temp=687 

power 6.59 11.19 10.99 10.66 10.61 10.44 10.44 10.99 

dt 10 30 33 36 21 29 27 5 

power 0 

dt 1739 

end history 

end
C-22

Table C.8 Polaris Input File for Limerick Assembly YJ1433 (GE11)

=polaris

title "2-D depletion based on Limerick Unit1 BWR, GE11 9x9 Assembly YJ1433 "

lib "xn56v7.1" 

sys BWR 

geom FuelNode : ASSM 9 [ 1.43764] 

channel COOL 

hgap [0.726 0.726] : MOD.1 MOD.1 

box hspan=[6.70306 rad=0.72644 thick=0.1905 

: 0.1179 : 4.021836 : 4.021836 ] 

opt KEFF UpscatterSuperGroup=1 

shield ALL=N FUEL=P GAD=R

deplete FUEL=YES GAD=YES

opt CRITSPEC Method="P1"

mat CAN.1 : ZIRC2 temp=560

mat CLAD.1 : ZIRC2 temp=560

mat [FUEL.22 : uo2_22 temp=1100 dens=10.45128

 comp uo2_22 : UOX 2.2

mat FUEL.28 : uo2_28 temp=1100 dens=10.45128

 comp uo2_28 : UOX 2.8 

mat FUEL.34 : uo2_34 temp=1100 dens=10.45128

 comp uo2_34 : UOX 3.4

mat FUEL.39 : uo2_39 temp=1100 dens=10.45128

comp uo2_39 : UOX 3.95

mat FUEL.100 : uo2_39 temp=1100 dens=10.45128

mat FUEL.36 : uo2_36 temp=1100 dens=10.45128

comp uo2_36 : UOX 3.6

mat GAD.1 : uo2gd_f50 temp=1100 dens=9.37605

comp uox34 : UOX 3.6

 comp uo2gd_f50 : WT GD2O3=5.0 uox34=-100 

mat GAD.100 : uo2gd_f50 temp=1100 dens=9.37605] 

mat COOL.1 : H2O temp=560 : dens=1.0

mat MOD.1 : H2O temp=560 : void=0

pin 1 : 0.4877 0.559 : FUEL.22 CLAD.1

pin 2 : 0.4877 0.559 : FUEL.28 CLAD.1

pin 3 : 0.4877 0.559 : FUEL.34 CLAD.1 

pin 4 : 0.4877 0.559 : FUEL.39 CLAD.1 

pin 5 : 0.4877 0.559 : FUEL.36 CLAD.1

pin 7 : 0.4877 0.559 : GAD.1 CLAD.1 

pin A : 0.4877 0.559 : FUEL.100 CLAD.1 

pin W1 [1.731 : 1.16840 1.2446] : MOD.1 can.1 COOL.1

pinmap 1

2 2

3 4 4

4 7 4 4

4 4 4 W1 _

4 7 5 W1 W1 4

3 4 5 5 4 4 7
C-23

2 2 4 7 A 7 4 2

1 2 3 4 4 4 5 2 1 

mesh COOL : nx=9 ny=9 ns=16 nr=2

mesh COOL.2 : nx=9 ny=9 ns=16 nr=4

mesh MOD : nf=4 nd=5 nx=10 ny=10 ns=16 nr=4

mesh MOD.2 : ns=4 nr=1

mesh FUEL : nr=3 ns=16

mesh GAD : nr=7 ns=16

%mesh CLAD : ns=16

mesh TUBE : ns=16

state ALL : temp=560

FUEL : temp=1100

GAD : temp=1100

COOL : temp=575 dens=0.3

%------------------------------------------------------------------------------% 

%------------------------------------------------------------------------------% 

basis ALL=no FUEL.100=yes

read history

%power is reduced by 1.0 void is reduced by 1.0

STATE COOL : dens = 0.3756 0.3294 0.3294 0.3305 0.3305 0.3673 0.3673 0.3298 0.3508 0.403 0.4777 

0.4777

power 27.832 28.364 28.364 29.381 29.381 29.271 29.271 33.895 31.112 35.509 29.422 0.0

dt 73.0 45.5 45.5 45.5 45.5 45.5 45.5 55.0 53.0 50.0 67.0 38.0

% 

STATE COOL : dens = 0.2614 0.2546 0.2723 0.2723 0.2926 0.2978 0.3076 0.3339 0.3181 0.3557 

0.3587 0.3587 0.4049 0.4856 0.4856

power 37.768 35.744 34.256 34.256 33.12 33.1 29.463 31.787 28.507 31.299 21.042 21.042 30.618 

28.601 0.0

dt 57.0 65.0 42.5 42.5 37.0 67.0 37.0 67.0 35.0 42.0 43.0 43.0 60.0 45.0 27.0

% 

STATE COOL : dens = 0.3478 0.3279 0.3358 0.3358 0.3575 0.3609 0.4173 0.4345 0.4135 0.3756 0.403 

0.4909

power 24.239 28.121 28.964 28.964 27.857 29.895 27.207 22.966 21.894 24.292 23.147 18.596

dt 57.0 78.0 60.5 60.5 71.0 74.0 71.0 81.0 41.0 48.0 48.0 63.0

power 0.0 0.0 0.0

dt 1518.0 13.0 33.0

end history

end

NUREG/CR-7251 

ORNL/TM-2018/782

Ian Gauld, Ugur Mertyurek

Oak Ridge National Laboratory

Managed by UT-Battelle, LLC

Oak Ridge, TN 37831-6170

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

This report describes research to develop a technical basis to expand burnup credit to boiling water reactors (BWRs). 

One of the largest components of uncertainty in burnup credit analyses is the predicted isotopic inventories of spent fuel 

as applied to the criticality safety analysis application model. The analysis of BWR fuel inventories is challenging due to 

the complexity of BWR assembly designs, the lack of publicly available radiochemical assay measurements, and limited 

access to documentation on fuel design and operating conditions. This study has compiled and evaluated experimental 

data on measured nuclide concentrations in commercial spent fuel for more than 75 fuel samples that cover a wide 

range of modern assembly designs and operating conditions. These data were applied to predict the net effect of 

isotopic uncertainties on the effective neutron multiplication factor for a representative spent nuclear fuel storage 

system. The experimental data, uncertainty analysis methodology, and results for a dry storage cask application system 

are described. The results are applicable to BWR burnup credit beyond peak reactivity where any initial gadolinium 

present in the fuel has been fully depleted.

boiling water reactor

burnup credit

nuclide uncertainty analysis

SCALE

spent nuclear fuel

isotopic validation

December 2018

Technical

Margins for Uncertainty in the Predicted Spent Fuel Isotopic Inventories 

for BWR Burnup Credit



NUREG/CR-7251 Margins for Uncertainty in the Predicted Spent Fuel Isotopic Inventories for BWR Burnup Credit December 2018