Abstract
Cell and burnup calculations are the basis 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 k inf) for different discretization strategies during 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 is investigated. The influence of the discretization strategies on the calculation time is evaluated.
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