The purpose of this calculation is to document the LaSalle Unit 1 boiling water reactor (BWR) fuel depletion calculations performed as part of the commercial reactor critical (CRC) evaluation program. The CRC evaluations constitute benchmark calculations that support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel in a geologic repository. This calculation incorporates control blade effects and minor variations in the SAS2H assembly modeling.

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.

The purpose of this calculation is to perform waste-form specific nuclear criticality safety calculations to aid in establishing criticality safety design criteria, and to identify design and process parameters that are potentially important to the criticality safety of Department of Energy (DOE) standardized Spent Nuclear Fuel (SNF) canisters.

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.

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.

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.

The waste forms under consideration for disposal in the repository at Yucca Mountain contain scores of radionuclides. It would be impractical and highly inefficient to model all of these radionuclides in a total system performance assessment (TSPA). Thus, the purpose of this radionuclide screening analysis is to remove from further consideration (screen out) radionuclides that are unlikely to significantly contribute to radiation dose to the public from a nuclear waste repository at Yucca Mountain.

The objective of the Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations is to determine the accuracy of the SAS2H control module of the baselined modular code system SCALE, Version 4.4A (STN: 10129-4.4A-00), in predicting the isotopic concentrations of spent fuel, and to quantify the overall effect that the differences between the calculated and measured isotopic concentrations have on the system reactivity. The scope of this calculation covers eight different spent fuel samples from a fuel assembly that was irradiated in the Limerick Unit 1 boiling water reactor (BWR).

The purpose of this calculation is to demonstrate that the preclosure performance objectives specified in 10 CFR 63.111(a) and 10 CFR 63.111(b) (Reference 2.2.1) have been met for the proposed design and operations in the geologic repository operations area (GROA) during normal operations and Category 1 event sequences, and following Category 2 event sequences. Category 1 event sequences are those natural and human-induced event sequences that are expected to occur one or more times before permanent closure of the repository.

This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC (Bechtel SAIC Company) 2004 (DIRS 167614).

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 analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide pressurized water reactor (PWR) isotopic composition data as a function of time for use in criticality analyses. The objectives of this evaluation are to generate burnup and decay dependant isotopic inventories and to provide these inventories in a form which can easily be utilized in subsequent criticality calculations.

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.