Cross Sections Influence on Monte Carlo Based Burnup Codes

This paper reports the results of a comparison among JEFF and ENDF/B datasets when used by SERPENT and MONTEBURNS codes on a GFR-like configuration. Particularly, it shows a comparison between the two Monte Carlo based codes, each one adopting three different cross sections dataset, namely JEFF-3.1, JEFF-3.1.2 and ENDF/BVII.1. Calculations have been carried out on the Allegro reactor, i.e. an experimental GFR-like facility that should be built in EU as GFR demonstrator. Results concern nuclear parameters as effective multiplication factor and fluxes, as well as the atomic densities for some important nuclides versus burnup