Nuclear data impact on multiplication factor and reactor physics parameters calculation for experiments simulating damp MOX powders

Abstract
The paper presents calculation results for experiments performed to address needs of the MOX fuel manufacturing process and particularly with low-moderated MOX fissile media. Computation of criticality for eleven configurations and reaction rates ratios for major and minor actinides are performed with CRISTAL french criticality code package, and the reference MCNPX and TRIPOLI-4.3 codes. The multigroup JEF-2.2, JEFF-3.1 and continuous energy JEF-2.2, JEFF-3.1, ENDF/B-VI and ENDF/B-VII neutron cross sections are used for the computations. The results are compared to verify group-wise calculations. Then the obtained calculated values will be used to predict eigenvalue accuracy for applications with damp MOX powder, and to validate ²³⁸U, ²³⁹Pu, and ²⁴⁰Pu cross sections.