On the influence of spatial discretization in LWR-burnup calculations with HELIOS 1.9 – part II: mixed oxide (MOX) fuel

Cell- and burnup calculations are the fundament for all deterministic static and transient 3D full core calculations for different operational states of the reactor. The arising differences in the integral transport solution (neutron flux and kinf) for different discretization strategies over the burnup of mixed oxide (MOX) fuel due to different spatial discretization are demonstrated. The influence of different discretization strategies on the calculation of homogenized few group cross sections which are forwarded to the 3D full core calculations is investigated and on the calculation time is evaluated. The differences between UOX and MOX fuel are discussed.