On the influence of spatial discretization in LWR-burnup calculations with HELIOS 1.9 – part I: Uranium oxide (UOX) 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 spatial discretization used for the cell- and burnup calculations influences the results for the used integral transport solutions significantly. The arising differences in the neutron flux distribution for different discretization strategies over the burnup of Uranium oxide (UOX) fuel due to different spatial discretization are demonstrated. These differences in the flux distribution cause significant changes in the kinf value. 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 The influence of the discretization strategies on the calculation time is evaluated.