This work investigates a method for generating medium fidelity reactor model cross sections. The methodology used here couples a neutron transport code with a depletion calculator. Together the two can be used to generate time dependent group cross sections for medium fidelity reactor models in fuel cycle simulation. This work analyzes XSgen, a software that performs the functions mentioned above. Addtionally, it is shown that XSgen is capable of creating datasets for a medium fidelity reactor model known as Bright-lite. The results of this work show that XSgen produces datasets that match NEA benchmarks for both uranium based LWR, and MOX reactors.
Cite
@article{arxiv.1712.02241,
title = {XSgen: A Cross Section Generation Methodology for Fuel Cycle Simulation},
author = {Robert Flanagan and Anthony Scopatz},
journal= {arXiv preprint arXiv:1712.02241},
year = {2017}
}