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 analysis is to provide input on the criticality potential of various degraded configurations to an analysis on the probability of a criticality event in a Pressurized Water Reactor (PWR) Advanced Uncanistered Fuel (AUCF) Waste Package (WP).

The preclosure criticality analysis process described in this technical report provides a systematic approach for determining the need for criticality controls and for evaluating their effectiveness during the preclosure period of the Monitored Geologic Repository at Yucca Mountain, Nevada.

The criticality consequence analysis for pressurized water reactor (PWR) waste packages (WP)
(Civilian Radioactive Waste Management System [CRWMS] Management and Operating
Contractor [M&O] 1997) focused on results obtained by maximizing postulated rates of
reactivity insertion to assure no synergistic reactions could occur among waste packages from
hypothetical criticality events. Other variables potentially influencing the criticality
consequences were held constant during the above referenced analysis. One of those variables

This model report documents the drift scale coupled thermal-hydrological-mechanical (THM) processes model development and presents simulations of the THM behavior in fractured rock close to emplacement drifts. The modeling and analyses are used to evaluate the impact of THM processes on permeability and flow in the near-field of the emplacement drifts.

The purpose of this analysis is to provide input on the criticality potential of various degraded configurations to an analysis on the probability of a criticality event in a Pressurized Water Reactor (PWR) Advanced Uncanistered Fuel (AUCF) Waste Package (WP).

The purpose of this engineering calculation is to estimate the frequency of misloading spent nuclear fuel (SNF) assemblies that would result in exceeding the criticality design basis of a waste package (WP). This type of misload - a reactivity misload - results from the incorrect placement of one or more fuel assemblies into a waste package such that the criticality controls do not match the required controls for the fuel assemblies. An actual criticality event can not occur in a WP unless a moderator (e.g. water) is present.

This report documents the development and validation of the in-drift precipitates/salts (IDPS) process model. The IDPS process model is a geochemical model designed to predict the postclosure effects of evaporation and deliquescence on the chemical composition of water within the Engineered Barrier System (EBS) in support of the total system performance assessment (TSPA). Application of the model in support of TSPA is documented in Engineered Barrier System: Physical and Chemical Environment (BSC 2005 [DIRS 175083]).

The purpose of these calculations is to characterize the criticality safety concerns for the storage of Fast Flux Test Facility (FFTF) nuclear fuel in a Department of Energy spent nuclear fuel (DOE SNF) canister in a co-disposal waste package. These results will be used to support the analysis that will be done to demonstrate concept viability related to use in the Monitored Geologic Repository (MGR) environment.