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Cross Sections Influence on Monte Carlo Based Burnup Codes

[+] Author Affiliations
Davide Chersola, Guglielmo Lomonaco

University of Genova & INFN, Genova, Italy

Guido Mazzini

Centrum výzkumu Řež, Husinec-Rez, Czech Republic

Paper No. ICONE22-31049, pp. V004T11A015; 9 pages
  • 2014 22nd International Conference on Nuclear Engineering
  • Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory
  • Prague, Czech Republic, July 7–11, 2014
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-4594-3
  • Copyright © 2014 by ASME


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/B-VII.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.

Copyright © 2014 by ASME



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