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.

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 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 Department of Energy (DOE) standardized Spent Nuclear Fuel (SNF) canisters.

The purpose of this analysis is to document Waste Package Development Department (WPPD) MCNP evaluations of benchmark solution Laboratory Critical Experiments (LCE's). The objective of this analysis is to quantify the ability of the MCNP 4A (Reference 5.4) code system to accurately calculate the effective neutron multiplication factor (keff) for various measured critical (i.e., keff=1.0) configurations.

There are more than 250 forms of U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program (NSNFP) has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. Fast Flux Test Facility (FFTF) fuel has been designated as the representative fuel for the mixed-oxide (MOX) fuel group which is a mixture of uranium and plutonium oxides.

The purpose of this analysis is to evaluate the types of defects or imperfections that could occur in a waste package or a drip shield and potentially lead to its early failure, and to estimate a probability of undetected occurrence for each type. An early failure is defined as the through-wall penetration of a waste package or drip shield due to manufacturing or handling-induced defects, at a time earlier than would be predicted by mechanistic degradation models for a defect-free waste package or drip shield.

The purpose of this calculation is to perform criticality evaluations for mixed oxide spent nuclear fuel (MOX SNF) in 12 and 21 Pressurized Water Reactor (PWR) waste packages (WPs) for both intact and degraded configurations.
The MOX assembly design considered in previous studies on Pu disposition in commercial reactors is based on the Westinghouse (W) 17x17 Vantage 5 assembly (Ref. 7.2). Depletion analyses of four Pu enrichment and burnup (expressed as gigawatt days/metric ton heavy metal; GWd/MTHM) combinations were performed in Reference 7.4. These are: