Abstract
Cell and burnup calculations are fundamental 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 significantly the results for the used integral transport solutions. The arising differences in the neutron flux distribution are demonstrated of different spatial discretization strategies during the burnup of Uranium oxide (UOX) fuel. These differences in the flux distribution cause significant changes in the isotopic densities and the k inf value. The influence of different discretization strategies on the calculation of homogenized few group cross-sections is investigated. The influence of the discretization strategies on the calculation time is evaluated.
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