Neutronic and burnup analysis of OECD-NEA MOX fuel benchmark with Dragon5 lattice code

This paper presents a comprehensive and detailed 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 the neutronic analysis of mixed uranium-plutonium oxide (MOX) fuel. The neutronics and burn-up calculations for rectangular pin and assembly geometries filled with different compositions of MOX fuel are computed. The performance of different nuclear data libraries is evaluated. Parameters such as infinite multiplication factor, reactivity change agree very well with the averaged reference values provided by the other institutions when the JEF2.2 library is used. Inventories of major actinides and fission product nuclides at different burn-up depths are also compared with published values at the MOX pin cell and assembly level, results of the Dragon5 lattice code are in agreement with averaged values provided by other codes. Furthermore, the deviation between newer libraries and reference solutions of the benchmark should be attributed to the differences in neutron data. 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. •Dragon5 deterministic code for neutronic and burnup analysis of MOX fuel is evaluated.•Weapons-grade plutonium MOX and thermal nuclear reactor spent fuel reprocessing MOX are computed.•Effect of different neutron data libraries are evaluated.•Reliability of Dragon5 code is verified by compared with published results and OpenMC.