Portable reactor data library interpolation via self-describing ORIGEN libraries

A prominent feature of the ORIGEN nuclear fuel depletion code (part of SCALE) is in the ability to use pre-generated reactor data libraries for common fuel types which can be interpolated to problem-specific conditions, thereby allowing for rapid depletion calculations without the need to repeat computationally-expensive lattice physics transport calculations. However, a limitation of the existing interpolation capability provided through the ARP module is that interpolation is limited to a set of fixed parameters, such as initial enrichment, moderator density, and burnup for UO₂ fuel assemblies. In this paper, we present recent developments toward a more generalized approach to reactor data interpolation that allows arbitrary specification of interpolation parameters at problem runtime, thereby affording an extension of interpolated reactor data library capabilities for ORIGEN to new problem types, including advanced reactors and fuel types. This generalized approach is premised upon the development of new, "self-describing" ORIGEN reactor data libraries which contain information on interpolable dimensions (including traditional parameters like enrichment, burnup, and moderator density, but also admitting parameters such as fuel temperature, etc.). Further, the new, generalized approach is designed to be extended to an arbitrary number of simultaneous interpolation dimensions, thus affording far greater flexibility in interpolating reactor data libraries to problem-specific conditions. Finally, we present the results of a benchmarking assessment of the new generalized interpolation approach against both interpolated cross-sections produced using the older ARP methodology as well as results from lattice physics transport calculations.