One-Group Cross Sections Generation for Monte Carlo Burnup Codes

Performing depletion calculations in coupled MC codes requires the knowledge of 1-g cross section for all relevant isotopes. The new subroutine implemented in BGCore code for taking into account the spatial shielding effects typically represented by the Bell and Dancoff factors produces high accuracy results without requiring any additional calculation time. The modified MG method relies on automatically generated MCNP statistical data that allows reconstruction of the Bell and Dancoff factors. The method was tested on a typical Sodium cooled fast reactor assembly design. The results show that using the original MG approach(7), e.g. in ALEPH code, leads to prohibitively large systematic errors. The main conclusion of this work is that the correction for calculating the effective chord length introduced into BGCore is essential in order for the code to be universally applicable to the analysis of a wide range of reactor systems. The presented method takes into account the fuel lattice geometry effects commonly described by the Bell and Dancoff factors and yields correct values of the 1-group cross sections for fast as well as thermal spectrum systems