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

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-6 and Endf/B-7 neutron cross sections are used for the computations. The results are compared to verify group-wise calculations. The calculated keff values tend to be lower than the benchmark values for configurations with weapons-grade Pu for all the libraries except Endf/B-6.6, and substantially higher for reactor-grade Pu, irrespective of the library used. Comparison of the multigroup APOLLO2-MORET 4 and continuous energy MCNPX results based on the same evaluated nuclear data files, are in a good agreement for both criticality and reaction rates computation with JEF-2.2 for all the configurations and for majority of the cases calculated with JEFF-3.1