Publications Repository - Helmholtz-Zentrum Dresden-Rossendorf

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 c burnup calculations influences the results for the used integral transport solutions significantly. The arising differences in the neutron flux distribution are demonstrated for different discretization strategies . These differences in the flux distribution cause significant changes in the infinite multiplication factor and in the localized burnup ditribution in the fuel rod for identical averaged burnup. The influence of different discretization strategies on the calculation of homogenized two group cross sections which are forwarded to the 3D full core calculations is investigated and the different calculation times are analyzed.