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

Abstract

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 238 U, 239 Pu, and 240 Pu cross sections.

Use of integral experiments in support to the validation of JEFF-3.2 nuclear data evaluation

“…Concerning BFS experiments [6], one can notice that a quite good agreement is obtained, except for experiments with 10% of 240 Pu (99/1 and 99/2). The The results are reported in Table 1, which exhibits a refinement of k eff using the JEFF-3.2 library.…”

Benchmarking of the JEFF-3.1 library against critical integral experiments

“…Regarding the validation of 238 U, 239 Pu and 240 Pu in the intermediate energy range, the new experimental programme performed in BFS reactor by IPPE (Russia) has been investigated [9]. Most of these 11 experiments have been categorized as MIXED since there are less than 50% of fissions occurring in fast, intermediate and thermal range.…”