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.

The objective of this calculation is to document the Grand Gulf Unit 1 (GG1) reactivity calculations for sixteen critical statepoints in cycles 4 through 8. The GG1 reactor is a boiling water reactor (BWR) owned and operated by Entergy Operations Inc. The Commercial Reactor Criticality (CRC) evaluations support the development and validation of the neutronic models used for criticality analyses involving commercial spent nuclear fuel to be placed in a geologic repository. This calculation is performed as part of the evaluation in the CRC program.

This calculation file provides comparisons of one- and two-dimensional methods for calculating the isotopic content of spent nuclear fuel. The one-dimensional methods use the SAS2H sequence of SCALE 4.4a (Reference 7.1) and the SAS2 sequence of SCALE 5.0 (Reference 7.2). The two-dimensional method uses the TRITON control module along with the T-DEPL sequence of SCALE 5.0 (Reference 7.3). The SAS2H results for SCALE 4.4a are taken from Reference 7.4. Data from previous two-dimensional calculations (Reference 7.5) using CASM03 will also be used for comparisons with TRITON.

The objective of this calculation is to evaluate the probability of flooding a waste package with seepage water. Disruptive events can affect the Engineered Barrier System (EBS) components and have the potential to allow an advective flow of seepage water to reach the waste package. The advective and diffusive flow paths into the waste package have the potential to result in water accumulation inside the waste package, which in turn can lead to a potentially critical configuration. This calculation will evaluate the following:

The purpose of this calculation is to document the LaSalle Unit 1 boiling water reactor (BWR) fuel depletion calculations performed as part of the commercial reactor critical (CRC) evaluation program. The CRC evaluations constitute benchmark calculations that support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel in a geologic repository. This calculation incorporates control blade effects and minor variations in the SAS2H assembly modeling.

This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the Project Design Criteria (PDC) Document (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in Project Requirements Document (Canori and Leitner 2003 [DIRS 166275], p.

This analysis documents the screening analysis for postclosure criticality features, events, and
processes (FEPs). It addresses the probability of criticality events resulting from degradation
processes as well as disruptive events (i.e., seismic, igneous, and rockfall). Probability
evaluations are performed utilizing the configuration generator model described in Configuration
Generator Model for In-Package Criticality1, a component of the methodology from Disposal

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 Emico Fermi Atomic Power Plant (Ref. 1). The Fermi fuel has been considered for disposal at the potential Yucca Mountain site.

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design.

The objective of the Limerick Unit 1 Radiochemical Assay Comparisons to SAS2H Calculations is to determine the accuracy of the SAS2H control module of the baselined modular code system SCALE, Version 4.4A (STN: 10129-4.4A-00), in predicting the isotopic concentrations of spent fuel, and to quantify the overall effect that the differences between the calculated and measured isotopic concentrations have on the system reactivity. The scope of this calculation covers eight different spent fuel samples from a fuel assembly that was irradiated in the Limerick Unit 1 boiling water reactor (BWR).