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This work investigates a method for generating medium fidelity reactor modelcross sections. The methodology used here couples a neutron transport code witha depletion calculator. Together the two can be used to generate time dependentgroup cross sections for medium fidelity reactor models in fuel cyclesimulation. This work analyzes XSgen, a software that performs the functionsmentioned above. Addtionally, it is shown that XSgen is capable of creatingdatasets for a medium fidelity reactor model known as Bright-lite. The resultsof this work show that XSgen produces datasets that match NEA benchmarks forboth uranium based LWR, and MOX reactors.

XSgen: A Cross Section Generation Methodology for Fuel Cycle Simulation