The purpose of this calculation is to estimate the probability of criticality in a pressurized water reactor (PWR) uncanistered fuel waste package during the postclosure phase of the repository as a function of various waste package material, loading, and environmental parameters. Parameterization on the upper subcritical limit that is used to define the threshold for criticality will also be performed. The possibility of waste package misload due to human or equipment error during preclosure is also considered in estimating the postclosure criticality probability.

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 objective of this analysis is to characterize the criticality safety aspects of a degraded Department of Energy spent nuclear fuel (DOE-SNF) canister containing Massachusetts Institute of Technology (MIT) or Oak Ridge Research (ORR) fuel in the Five-Pack Defense High-Level Waste (DHLW) waste package to demonstrate concept viability related to use in the Mined Geologic Disposal System (MGDS) environment for the postclosure time frame.

The objective of this analysis is to evaluate accumulations within the thermally altered tuff surrounding a drift. The evaluation examines accumulation of Uranium minerals (sddyite), Plutonium oxide (Pu2O), and combinations of these materials. A hypothetical model of the tuff is used to provide insight into the factors that affect criticality for this near-field scenario. The factors examined include: the size of the accumulation, the fissile composition of the accumulation, the water of clayey material in the accumulation and the water fraction in the tuff.

The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall.
The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared ·according to Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]).

The Disposal Criticality Analysis Methodology Topical Report (Reference 1) states that the accuracy of the criticality analysis methodology (MCNP Monte Carlo code and cross-section data) designated to assess the potential for criticality of various configurations in the Yucca Mountain proposed repository is established by evaluating appropriately selected benchmark critical experiments.

The purpose of this calculation is to characterize the criticality safety concerns for the codisposal of Shippingport pressurized water reactor (SP PWR) spent nuclear fuel (SNF) contained in a standardized Department of Energy (DOE) SNF canister, and high-level waste (HLW) glass in a waste package (WP) placed in a Monitored Geologic Repository (MGR). The result of this calculation will be used to evaluate criticality issues and provide input for the DOE SNF canister design, referred to as "the canister" in this document.

The Disposal Criticality Analysis Methodology Topical Report prescribes an approach to the methodology for performing postclosure criticality analyses within the monitored geologic repository at Yucca Mountain, Nevada. An essential component of the methodology is the Configuration Generator Model for In-Package Criticality that provides a tool to evaluate the probabilities of degraded configurations achieving a critical state.

The purpose of this calculation is to perform partially and fully degraded mode criticality evaluations of plutonium disposed of in a ceramic waste form and emplaced in a Monitored Geologic Repository. The partially degraded mode is represented by the immobilized plutonium ceramic discs piled in the bottom of the waste package (WP) while neutron absorbers begin to leach out of the discs.

The intended purpose of the multiscale thermohydrologic model (MSTHM) is to predict the possible range of thermal-hydrologic conditions, resulting from uncertainty and variability, in the repository emplacement drifts, including the invert, and in the adjoining host rock for the repository at Yucca Mountain. The goal of the MSTHM is to predict a reasonable range of possible thermal-hydrologic conditions within the emplacement drift.