The "Summary Report of SNF Isotopic Comparisons for the Disposal Criticality Analysis Methodology" contains a summary of the analyses that compare SNF measured isotopic concentrations (radiochemical assays) to calculated SNF isotop~c concentrations (SAS2H module ·orScale4.3). The results of these analyses are used to support the validation of the isotopic models for spent commercial light water reactor (LWR) fuel.

The purpose of this calculation is to document the Crystal River Unit 3 pressurized water reactor (PWR) fuel depletion calculations performed as part of the commercial reactor critical (CRC) evaluation program. The CRC evaluations support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel in a geologic repository.

This report presents the results of computer code benchmark simulations against spent fuel radiochemical assay
measurements from the Kansai Electric Ltd. Takahama-3 reactor published by the Japan Atomic Energy
Research Institute. Takahama-3 is a pressurized-water reactor that operates with a 17 × 17 fuel-assembly design.
Spent fuel samples were obtained from assemblies operated for 2 and 3 cycles and achieved a maximum burnup
of 47 GWd/MTU. Radiochemical analyses were performed on two rods having an initial enrichment of

Pressurized water reactor (PWR) burnup credit validation is
demonstrated using the benchmarks for quantifying fuel reactivity
decrements, published as Benchmarks for Quantifying Fuel Reactivity
Depletion Uncertainty, Electric Power Research Institute (EPRI)
report 1022909. This demonstration uses the depletion module
TRITON (Transport Rigor Implemented with Time-Dependent
Operation for Neutronic Depletion) available in the SCALE 6.1
(Standardized Computer Analyses for Licensing Evaluations) code

This report presents a methodology for validation of the isotopic
contents of spent light water reactor fuel for actinide-only burnup
credit with additional high-quality radiochemistry assay (RCA) data
obtained from the Yankee Rowe pressurized water reactor. The
additional Yankee Rowe RCA data were not included in previous
isotopic validation studies for burnup credit due to the difficulty of
accurately modeling the complex Yankee Rowe fuel assembly design
using the SAS2H one-dimensional sequence of the earlier SCALE

This paper provides insights into the neutronic similarities between a representative high-capacity rail-transport cask containing typical pressurized water reactor (PWR) spent nuclear fuel assemblies and critical reactor state-points, referred to as commercial reactor critical (CRC) state-points. Forty CRC state-points from five PWRs were analyzed, and the characteristics of CRC state-points that may be applicable for validation of burnup-credit criticality safety calculations for spent fuel transport/storage/disposal systems were identified.

This report is part of a report series designed to document benchmark-quality radiochemical isotopic
assay data against which computer code accuracy can be quantified to establish the uncertainty and bias
associated with the code predictions. The experimental data included in the report series were acquired
from domestic and international programs and include spent fuel samples that cover a large burnup range.
The measurements analyzed in the current report, for which experimental data is publicly available,

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
(AFR) criticality safety analyses be validated against experimental measurements. If credit for the
negative reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark
computational methods against spent fuel critical configurations. This report summarizes a portion
of the ongoing effort to benchmark AFR criticality analysis methods using selected critical
configurations from commercial pressurized-water reactors (PWR).

STARBUCS is a new prototypic analysis sequence for performing automated criticality safety analyses of spent fuel systems employing burnup credit. A depletion analysis calculation for each of the burnup-dependent regions of a spent fuel assembly, or other system containing spent fuel, is performed using the ORIGEN-ARP sequence of SCALE. The spent fuel compositions are then used to generate resonance self-shielded cross sections for each region of the problem, which are applied in a three-dimensional criticality safety calculation using the KENO V.a code.

The purpose of this design analysis is to determine the accuracy of the SAS2H module of SCALE 4.3 in predicting isotopic concentrations of spent fuel assemblies. The objective is to develop a methodology for modeling assemblies similar to those evaluated within this analysis and to establish the consistency of SAS2H predictions. The results of this analysis may then be applied to future depletion calculations using SAS2H in which no measurements are available.

The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor
criticality safety analyses be validated against experimental measurements. If credit for the negative
reactivity of the depleted (or spent) fuel isotopics is desired, it is necessary to benchmark
computational methods against spent fuel critical configurations. This report summarizes a portion
of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical
configurations from commercial pressurized-water reactors.

The purpose of this design analysis is to determine the accuracy of the SAS2H module of SCALE 4.3 in predicting isotopic concentrations of spent fuel assemblies. The objective is to develop a methodology for modeling assemblies similar to those evaluated within this analysis and to establish the consistency of SAS2H predictions. The results of this analysis may then be applied to future depletion calculations using SAS2H in which no measurements are available.

This calculation file provides comparisons of one- and two-dimensional methods for calculating the isotopic content of spent nuclear fuel. The one-dimensional methods use the SAS2H sequence of SCALE 4.4a (Reference 7.1) and the SAS2 sequence of SCALE 5.0 (Reference 7.2). The two-dimensional method uses the TRITON control module along with the T-DEPL sequence of SCALE 5.0 (Reference 7.3). The SAS2H results for SCALE 4.4a are taken from Reference 7.4. Data from previous two-dimensional calculations (Reference 7.5) using CASM03 will also be used for comparisons with TRITON.

Thirty spent fuel samples obtained from boiling-water-reactor (BWR) fuel pins have been
modeled at Oak Ridge National Laboratory using the SAS2H sequence of the SCALE code system.
The SAS2H sequence uses transport methods combined with the depletion and decay capabilities
of the ORIGEN-S code to estimate the isotopic composition of fuel as a function of its burnup
history. Results of these calculations are compared with chemical assay measurements of spent fuel
inventories for each sample. Results show reasonable agreement between measured and predicted

Disposal Criticality Analysis Methodology Topical Report describes a methodology for performing postclosure criticality analyses within the repository at Yucca Mountain, Nevada. An important component of the postclosure criticality analysis is the calculation of conservative isotopic concentrations for spent nuclear fuel. This report documents the isotopic calculation methodology. The isotopic calculation methodology is shown to be conservative based upon current data for pressurized water reactor and boiling water reactor spent nuclear fuel.

The purpose of this design analysis is to determine the accuracy of the SAS2H module of SCALE 4.3 in predicting isotopic concentrations of spent fuel assemblies. The objective is to develop a methodology for modeling assemblies similar to those evaluated within this analysis and to establish the consistency of SAS2H predictions. The results of this analysis may then be applied to future depletion calculations using SAS2H in which no measurements are available.

The purpose of this design analysis is to determine the accuracy of the SAS2H module of SCALE 4.3 in predicting isotopic concentrations of spent fuel assemblies. The objective is to develop a methodology for modeling assemblies similar to those evaluated within this analysis and to establish the consistency of SAS2H predictions. The results of this analysis may then be applied to future depletion calculations using SAS2H in which no measurements are available.

The purpose of this study is to provide insights into the neutronic similarities that may exist between a
generic cask containing typical spent nuclear fuel assemblies and commercial reactor critical (CRC) state-
points. Forty CRC state-points from five pressurized-water reactors were selected for the study and the
type of CRC state-points that may be applicable for validation of burnup credit criticality safety
calculations for spent fuel transport/storage/disposal systems are identified. The study employed cross-

Isotopic characterization of spent fuel via depletion and decay calculations is necessary for
determination of source terms for subsequent system analyses involving heat transfer, radiation
shielding, isotopic migration, etc. Unlike fresh fuel assumptions typically employed in the criticality
safety analysis of spent fuel configurations, burnup credit applications also rely on depletion and
decay calculations to predict the isotopic composition of spent fuel. These isotopics are used in

The purpose of this design analysis is to determine the accuracy of the SAS2H module of SCALE 4.3 in predicting isotopic concentrations of spent fuel assemblies. The objective is to develop a methodology for modeling assemblies similar to those evaluated within this analysis and to establish the consistency of SAS2H predictions. The results of this analysis may then be applied to future depletion calculations using SAS2H in which no measurements are available.