INTEGRATED DATA AND ANALYSIS SYSTEM FOR COMMERCIAL USED NUCLEAR FUEL SAFETY ASSESSMENTS

Uncertainties related to meeting packaging and transportation regulatory criteria are increasing as commercial used nuclear fuel (UNF) is being stored at reactor sites for longer time intervals than originally foreseen. As the storage times continue to increase issues associated with aging management and the potential consequences of the deleterious effects of aging will need to be assessed. The assessments demonstrate the continued efficacy of the storage system over extended storage (ES) periods, and also evaluate safety for subsequent transportation following ES. Licensed storage and transportation cask systems have well-defined assembly-loading criteria (e.g., specifications for “approved contents” in a storage cask system’s Certificate of Compliance). These specifications are typically used to define limiting loading conditions and characteristics for which the cask system’s safety analysis report has demonstrated compliance with the applicable regulatory requirements. In practice, due to the diversity in the discharged UNF available for loading (e.g., variations in UNF assembly burnup values, initial enrichments, discharge date, etc.), it is not possible to load a cask system with UNF assemblies that correspond exactly to the limiting licensing conditions. Hence, cask systems are loaded with assemblies that satisfy the limiting loading conditions with some amount of unquantified, uncredited margin. This reality in storage and transportation cask loading provides additional conservatism with respect to the regulatory safety requirements. These potentially large (depending on the specific loading conditions) safety margins may be quantified and potentially credited in the future to offset uncertainties in safety margins associated with ES and high-burnup fuel issues. To investigate this possibility, an integrated data and analysis capability is being developed to estimate safety margins in actual as-loaded storage and transportation casks. This paper describes how UNF inventory data, fuel assembly design data, site-specific cask loading data, and reactor operating data are coupled with fuel depletion, criticality safety, and thermal computational analysis capabilities to provide out-of-reactor nuclear safety evaluations in an autonomous manner. Comparisons with limiting loading specifications are provided to quantify the available margin present in a variety of cask systems being used at different nuclear sites.