The purpose of this engineering calculation is to provide the chemical composition for the Department of Energy (DOE) Savannah River Site (SRS) High-Level Waste (HLW) glass. Since the glass is to be co-disposed with other DOE spent nuclear fuels (SNFs) in the Monitored Geologic Repository (MGR), its chemical composition is needed for the design of the co-disposal canisters and waste packages in term of criticality and degradation.

The purpose of this calculation is to perform criticality evaluations for mixed oxide spent nuclear fuel in 12 and 21 pressurized water reactor waste packages 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 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 Ref. 7.4.

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 these calculations is to characterize the criticality safety concerns for the storage of Fast Flux Test Facility (FFTF) nuclear fuel in a Department of Energy spent nuclear fuel (DOE SNF) canister in a co-disposal waste package. These results will be used to support the analysis that will be done to demonstrate concept viability related to use in the Monitored Geologic Repository (MGR) environment.

The purpose of this engineering calculation is to estimate the frequency of misloading spent nuclear fuel (SNF) assemblies that would result in exceeding the criticality design basis of a waste package (WP). This type of misload - a reactivity misload - results from the incorrect placement of one or more fuel assemblies into a waste package such that the criticality controls do not match the required controls for the fuel assemblies. An actual criticality event can not occur in a WP unless a moderator (e.g. water) is present.

The objective of this analysis is to characterize the criticality safety aspects of a degraded Department of Energy spent nuclear fuel (DOESNF) canister containing Masachusetts 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 Minded Geologic Disposal System (MGDS) environment for the postclosure time frame.

The purpose of this calculation is to document the validity of the commercial reactor criticals (CRC) as a source for a spent nuclear fuel benchmark, and to characterize the neutronic similarities between a CRC and a waste package (WP). This report illustrates comparisons of neutron spectrum and the effects on criticality arising from physical differences between a WP and a CRC. This report is an engineering calculation supporting the development of the disposal criticality analysis methodology, performed under Quality Administrative Procedure (QAP)-3-15 Revision 0.

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 report is to evaluate the potential for penetration of the Alloy 22 (UNS N06022) waste package outer barrier by localized corrosion due to the deliquescence of soluble constituents in dust present on waste package surfaces. The results support a recommendation to exclude deliquescence-induced localized corrosion (pitting or crevice corrosion) of the outer barrier from the total system performance assessment for the license application (TSPA-LA).

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package.