Use of the local Pu-239 concentration as an indicator of burnup spectral history in DYN3D

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density.This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS.