This model report documents the abstraction of drift seepage, conducted to provide seepage relevant parameters and their probability distributions for use in Total System Performance Assessment for License Application (TSPA-LA). Drift seepage refers to the flow of liquid water into waste emplacement drifts.

The purpose of the material degradation and release (MDR) model is to predict the fate of the waste package materials, specifically the retention or mobilization of the radionuclides and the neutron-absorbing material as a function of time after the breach of a waste package during the 10,000 years after repository closure. The output of this model is used directly to assess the potential for a criticality event inside the waste package due to the retention of the radionuclides combined with a loss of the neutron-absorbing material.

The purpose of this calculation is to simulate the processing of an incoming waste stream into waste packages, simulating the required aging as applicable, and the emplacement of the waste packages into the Yucca Mountain repository.

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.

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay.

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay.

The Monitored Geologic Repository (MGR) Waste Package Operations (WPO) 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 Shippingport Pressurized Water Reactor (PWR) (Ref. 1). The Shippingport PWR SNF has been considered for disposal at the proposed Yucca Mountain site.

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.

Presented at the NEI Used Fuel Management Conference, St. Petersburg, FL, May 7-9, 2013

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.

This calculation is prepared by the Monitored Geologic Repository Waste Package Requirements & Integration Department. The purpose of this calculation is to compile source term and commercial waste stream information for use in the analysis of waste package (WP) designs for commercial fuel. Information presented will consist of the number of WPs, source terms, metric tons of uranium, and the average characteristics of assemblies to be placed in each WP design. The source terms provide thermal output, radiation sources, and radionuclide inventories.

The objective of this calculation is to evaluate the peak temperatures due to thermal loading and boundary conditions of the TAD Waste Package design under nominal Monitored Geologic Repository conditions.

This report was prepared in accordance with Technical Work Plan for: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package (BSC 2004 [DIRS 171583]). This report provides information on the phase stability of Alloy 221, the current waste package outer barrier material. The goal of this model is to determine whether the single-phase solid solution is stable under repository conditions and, if not, how fast other phases may precipitate.

This study provides results supporting the conclusion that the repository can be operated over a varying range of thermal modes and therefore temperatures. In particular, this work focused on limiting the peak, postclosure waste package surface temperature to less than 85 degrees Celsius, a possible limit due to corrosion considerations. These operating modes were compared by varying the waste package in drift spacing (0.1-2.83 meters), drift pitch (drift spacing centerline to centerline of 40-120 meters), ventilation duration (75-300 years), and ventilation efficiency (50-80%).

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.

The Monitored Geologic Repository (MGR) Waste Package Project of the BSC Management and Operating Contractor for the Department of Energy's Office of Civilian Radioactive Waste Management performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Enrico Fermi Reactor owned by the DOE (Ref. 9). The Fermi SNF has been considered for disposal at the proposed Yucca Mountain site.

This evaluation is prepared by the Mined Geologic Disposal System (MGDS) Waste Package Development Department (WPDD) to provide analyses of disposal of aluminum (AI)-based Department of Energy-owned research reactor spent nuclear fuel (DOE-SNF) in a codisposal waste package with five canisters of high-level waste (HLW). The analysis was performed in sufficient detail to establish the technical viability of the Al-based DOE-SNF codisposal canister option.

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 Training, Research, Isotopes, General Atomics (TRIGA) reactor (Ref. 1). The TRIGA SNF has been considered for disposal at the potential Yucca Mountain site.

The Monitored Geologic Repository (MGR) Waste Package Department 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 Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site.

This report presents a two-phased approach to develop and analyze a “thermal envelope” to represent the postclosure response of the repository to the anticipated range of repository design thermal loadings. In Phase 1 an estimated limiting waste stream (ELWS) is identified and analyzed to determine the extremes of average and local thermal loading conditions. The coldest thermal loading condition is represented by an emplacement drift loaded exclusively with high-level radioactive waste (HLW) and/or defense spent nuclear fuel (DSNF).

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).

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<p><span style="font-size: 12.000000pt; font-family: 'TimesNewRomanPSMT'">This report fulfills the M1 milestone M11UF041401, &ldquo;Storage R&amp;D Opportunities Report&rdquo; under Work Package Number FTPN11UF0414. </span></p>