This analysis is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) in response to a request received via a QAP-3-12 Design Input Data Request (Reference 5.1) from Waste Acceptance, Storage, & Transportation (WAST) Design (formerly MRSMPC Design). This design analysis is an answer to the Design Input Data Request to provide: Specific requirements for long-term criticality control.

This report presents the analysis and conclusions with respect to disposal criticality for canisters containing aluminum-based fuels from research reactors. The analysis has been divided into three phases. Phase I, dealt with breached and flooded waste packages containing relatively intact canisters and intact internal (basket) structures; Phase II, the subject of this report, covers the degradation of the spent nuclear fuel (SNF) and structures internal to the codisposal waste package including high level waste (HLW), canisters, and critically control material.

The Monitored Geologic Repository Waste Package Operations of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Navy (Refs. 1 and , 2). The Navy SNF has been considered for disposal at the potential Yucca Mountain site. For some waste packages, the containment may breach (Ref. 3), allowing the influx of water. Water in the waste package may moderate neutrons, increasing the likelihood of a criticality event within the waste package.

The purpose of this document is to evaluate the potential for criticality for the fissile material that could accumulate in the near-field (invert) and in the far-field (host rock) beneath the U.S. Department of Energy (DOE) spent nuclear fuel (SNF) codisposal waste packages (WPs) as they degrade in the proposed monitored geologic repository at Yucca Mountain. The scope of this calculation is limited to the following DOE SNF types: Shippingport Pressurized Water Reactor (PWR), Enrico Fermi, Fast Flux Test Facility (FFTF), Fort St.

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 (dCB).

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 model report is to describe the evolution of the physical and chemical environmental conditions within the waste emplacement drifts of the repository, including the drip shield and waste package surfaces. This report documents the development of a new process-level model, the near-field chemistry (NFC) model, and develops two abstraction models.

The purpose of this calculation is to perform a parametric study to determine the effects of fission product leaching, assembly collapse, and iron oxide loss (Me203) on the reactivity of a waste package (WP) containing mixed oxide (MOX) spent nuclear fuel (SNF). Previous calculations (CRWMS M&O 1998a) have shown that the criticality control features of the WP are adequate to prevent criticality of a flooded WP for all the enrichment/ burnup pairs expected for the MOX SNF.

This analysis is prepared by the Mined Geologic Disposal System (MODS) Waste Package Development (WPD) department to provide an evaluation of the criticality potential within a waste package having some or all of its contents degraded by corrosion and removal of neutron absorbers. This analysis is also intended to provide an estimate of the consequences of any internal criticality, particularly in terms of any increase in radionuclide inventory. These consequence estimates will be used as part of the WPD input to the Total System Performance Assessment.

The purpose of this report is to <,locument the sensitivity of the drift-scale thermal-hydrologic- chemical (THC) seepage model (SNL 2007 [DIRS 177404]) to heterogeneities in permeability and capillarity, which could affect predicted fluxes and chemistries of water and gases seeping into the emplacement drifts. This report has been developed following Technical Work Plan for: Revision of Model Reports for Near-Field and In-Drift Water Chemistry (SNL 2007 [DIRS 179287]).