Multi-group cross section for new group structure using weightiung spectrum for sodium cooled fast reactor

The cross section data is the primary requirement to carry-out multi-group calculations for nuclear reactor design and analysis. The multi-group cross section data library is generated for a Multipurpose Experimental Sodium Cooled Fast Reactor (MESOF). A specific flux weighting spectra is used for point averaging the cross section obtained from Evaluated Nuclear Data Files (ENDF/B-VII.0). A new group structure is formed and well-known established procedure for nuclear data processing is adopted. The multi-group calculations are performed using diffusion and transport theory based codes for sodium cooled fast reactor with new and existing multi-group data libraries. The calculations have shown improved results for multiplication factor.