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).

This calculation compares results from criticality evaluations for a 21-assembly pressurized water reactor (PWR) waste package based on 12 axial burnup profile representations for commercial spent nuclear fuel (SNF) assemblies. The burnup profiles encompass the axial variations caused by different fuel assembly irradiation histories in a commercial PWR, including end effects, and the concomitant effect on reactivity in the waste package. The bounding axial burnup profiles in Table T of reference 6.3 are used for this analysis.

The purpose of this engineering calculation is to document the benchmark range, over a variety of parameters, for the validation of the criticality calculations supporting the Monitored Geologic Repository (MGR). This engineering calculation accomplishes this by characterizing the Laboratory Critical Experiments (LCE) and the Pressurized Water Reactor (PWR) Commercial Reactor Criticals (CRC), and summarizing the significant parameters. This engineering calculation supports the Disposal Criticality Analysis Methodology program.

This report was developed in accordance with the requirements in Technical Work Plan for Postclosure Waste Form Modeling (BSC 2005 [DIRS 173246]). The purpose of the in-package chemistry model is to predict the bulk chemistry inside of a breached waste package and to provide simplified expressions of that chemistry as a function of time after breach to Total Systems Performance Assessment for the License Application (TSPA-LA).

This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide an assessment of the present waste package design from a criticality risk standpoint. The specific objectives of this initial analysis are to:
1. Establish a process for determining the probability of waste package criticality as a function of time (in terms of a cumulative distribution function, probability distribution function, or expected number of criticalities in a specified time interval) for various waste package concepts;

The objective of this calculation is to evaluate the criticality potential for co-disposal waste packages affected by an igneous intrusion disruptive event in the emplacement drifts. The scope of this calculation is limited to U.S. Department of Energy (DOE) Spent Nuclear Fuel (SNF) types in DOE standardized SNF canisters or Multi-Canister Overpack (MCO) Canisters.

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 the transportation, aging and disposal (TAD) canister-based systems.

The purpose of this report is to demonstrate that postclosure temperature limits can be met, and certain thermal characteristics of the postclosure thermal reference case can be preserved, with alternative thermal loading schemes. The analysis considers certain variations from the base case.waste stream, the predicted postclosure temperatures that develop within the rock mass due to these waste stream variations, and then compares these temperatures to postclosure temperature limits.

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 analytical model employed for this analysis was the SAS2H module of the SCALE sequence.

The purpose of this activity is to develop a representative “limiting” axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the “end-effect”. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package.