380. MOX燃料炉物理試験MISTRALの解析

Abstract

Nuclear Power Engineering Corporation (NUPEC) has been performing conceptual design studies of high moderation full MOX LWR cores that aim for increasing fissile Pu consumption rate and reducing residual Pu in discharged MOX fuel. As part of these studies, NUPEC, French Atomic Energy Commission (CEA) and their industrial partners implemented an experimental program, MISTRAL, that was devoted to measuring the core physics parameters of such advanced cores. The program consists of one reference UO2 core, two homogeneous full MOX cores and one full MOX PWR mock-up core that have higher moderation ratio than the conventional lattice.NUPEC has been analyzing the experimental results with the diffusion and the transport calculations by the SRAC code system and the continuous energy Monte Carlo calculations by the MVP code with the common nuclear data file, JENDL-3.2. The calculation results well reproduce the experimental data approximately within the same range of the experimental uncertainty. This indicates that these applied analysis methods give the same accuracy for the UO2 core and MOX cores, for the different moderation MOX cores, and for the homogeneous and the mock-up MOX cores.