Features of Hybrid Core Calculations at BFS-62-3A Experiments

Abstract

Different uranium-plutonium fuel compositions are considered for sodium fast reactors of the next generation. Considerable discrepancies in axial and radial neutron spectra for hybrid reactor systems compared to uranium oxide fuel cores increase uncertainties in the key calculated neutronic characteristics of hybrid systems. The calculation results of a BFS-62-3A critical assembly considered as a full-scale model of BN-600 hybrid core with steel reflector specify quite different spectra in local areas. In such systems the MCNP5 calculations demonstrate a noticeable sensitivity of the key neutronic characteristics (effective multiplication factor keff, spectral indices) to nuclear data libraries and extra steel such as dowels placed in the core. Uncertainties in the location of stainless steel dowels and in their quantity cause uncertainties in the fuel-to-steel mass ratio in the core. For 235U, 238U, and 239Pu, the calculated radial fission rate distributions against the reconstructed ones are analyzed. A comparative analysis of spectral indices, neutron spectra, and radial fission rate distributions is performed using nuclear data libraries generated from ENDF/B-VII.0, JEFF-3.1.1, JENDL-3.3, and BROND-3 for Fe and Cr isotopes. When performing analysis of the fission-rate sensitivity to the presence of plutonium in fuel, 239Pu is replaced by 235U in local areas containing plutonium. For radial fission rate distributions, peak discrepancies may be due to possible underestimation of some features of experimental data processing and reconstruction methods (Westcott factors, temperature dependence, local core features). A more-sophisticated impact analysis of spatially different neutron spectra on neutron characteristics of the core is also required. To confirm the results of BFS-62-3A analysis, radial fission-rate distributions are calculated for BFS-62-4 with UO2 blanket instead of steel reflector.