The objective of this calculation is to document the Grand Gulf Unit 1 (GG1) reactivity calculations for sixteen critical statepoints in cycles 4 through 8. The GG1 reactor is a boiling water reactor (BWR) owned and operated by Entergy Operations Inc. The Commercial Reactor Criticality (CRC) evaluations support the development and validation of the neutronic models used for criticality analyses involving commercial spent nuclear fuel to be placed in a geologic repository. This calculation is performed as part of the evaluation in the CRC program.

The objective of this calculation is to determine the structural response of the 5-DHLW/DOE (Defense High Level Waste/Department of Energy) SNF (Spent Nuclear Fuel) Short Co-disposal Waste Package (WP) when subjected (while in the horizontal orientation emplaced in the drift) to a collision by a loaded (with WP) Transport and Emplacement Vehicle (TEV) due to an over-run. The scope of this calculation is limited to reporting the calculation results in terms of maximum total stress intensities (SIs) in the outer corrosion barrier (OCB).

The Nuclear Waste Policy Act of 1982, as amended (the Act), assigns the Federal Government the responsibility for the disposal of spent nuclear fuel and high-level waste. Section 302a of the Act authorized the Secretary to enter into contracts with the owners and generators of commercial spent nuclear fuel and or high level waste. The Standard Contract for Disposal of Spent Nuclear Fuel and or High Level Radioactive Waste (Standard Contract) established the contractual mechanism for the Department's acceptance and disposal of spent nuclear fuel and high level waste.

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.

Spent Fuel Project Office, Interim Staff Guidance - 8, Revision 2 - Burnup Credit in the Criticality Safety Analyses of PWR Spent Fuel in Transport
and Storage Casks

The purpose of this report, and the information contained in the associated computerized data bases, is to establish the DOE/OCRWM reference characteristics of the radioactive waste materials that may be accepted by DOE for emplacement in the mined geologic disposal system as developed under the Nuclear Waste Policy Act of 1982. This report provides relevant technical data for use by DOE and its supporting contractors and is not intended to be a policy document.

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.

DOE decided to terminate the Yucca Mountain repository program because, according to DOE officials, it is not a workable option and there are better solutions that can achieve a broader national consensus. DOE did not cite technical or safety issues. DOE also did not identify alternatives, but it did create a Blue Ribbon Commission to evaluate and recommend alternatives.

Spent fuel transportation and storage cask designs based on a burnup credit approach must
consider issues that are not relevant in casks designed under a fresh-fuel loading assumption. For
example, the spent fuel composition must be adequately characterized and the criticality analysis
model can be complicated by the need to consider axial burnup variations. Parametric analyses are
needed to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel

Several rod consolidation systems have been demonstrated in the United States with simulated boiling water reactor (BWR) and pressurized water reactor (PWR) fuel. The first U.S. consolidation of irradiated fuel was successfully demonstrated with four PWR fuel assemblies at the Oconee Nuclear Station in October-November 1982 (1-3) and with one PWR fuel assembly at Maine Yankee in August 1983(4). Maine Yankee has received approval from the U.S. Nuclear Regulatory Commission (NRC) to consolidate up to 20 fuel assemblies(5).