VERIFICATION OF THE DRAGON5 DETERMINISTIC CODE FOR NEUTRONIC AND BURNUP ANALYSIS OF OECD-NEA MOX FUEL BENCHMARK

抄録

This work presents comprehensive and elaborate analysis of the OECD-NEA MOX fuel benchmark based on different nuclear data libraries to investigate the reliability and accuracy of the Dragon5 lattice code developed by École Polytechnique de Montréal for neutronic analysis of mixed uranium-plutonium oxide fuel (MOX) fuel. The neutronics and burn-up calculations for rectangular pin and assembly geometry filling with different compositions of MOX fuel are computed and investigated. Performance of different nuclear data library are evaluated and compared. Parameters such as infinite multiplication factor, reactivity change agree very well with the averaged reference values provided by the other institutions if the JEF2.2 library is used. Inventories of important actinides and fission product nuclides at different burn-up depth are also compared with the published values at the MOX pin cell and assembly level, results of Dragon5 lattice code are consistent with averaged values provided by other codes. Further, the deviation between newer libraries and solutions of benchmark reference should be attributed to the differences of neutron data in these libraries. Therefore, the Dragon5 lattice code is reliable for neutronics and burn-up analysis of MOX fuel and can be applied to the neutronic analysis of mixed uranium-plutonium oxide fuel.