Cross Sections Influence on Monte Carlo Based Burnup Codes Applied to a GFR-like Configuration

This paper reports the results of a comparison among JEFF and ENDF/B data sets when used by SERPENT and MONTEBURNS codes on a gas-cooled fast reactor (GFR)-like configuration. Particularly, it shows a comparison between the two Monte Carlo-based codes, each one adopting three different cross-section data sets, namely, JEFF-3.1, JEFF-3.1.2, and ENDF/B-VII.1. Calculations have been carried out on the Allegro reactor, i.e., an experimental GFR-like facility that could be built in the European Union as a GFR demonstration. Results include nuclear parameters, such as the effective multiplication factor and fluxes, as well as the atomic densities for some important nuclides versus burn-up