Generation of Multi-group Cross Sections with Improved Self Shielding Calculations
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The multi group cross section data library is one of the major portion of nuclear reactor design and calculations. The generation of multi-group cross section from up to date nuclear data sheets (ENDF) is carried out for direct whole core neutron transport calculation code, nTRACER which is based on the method of characteristics (MOC). The subgroup method is used to conserve the effective cross section in the presence of resonance interference effect among resonance isotopes. The technique of handling the resonance interference among resonance isotopes is improved through the introduction of interference correction factors in subgroup method. A mathematical relationship based on subgroup formulism is derived to compute the interference correction factors. This new approach has been implemented in nTRACER code. Sample calculations have been performed for various pin and assembly problems with satisfactory performance. The calculation results for effective resonance cross sections show maximum error of less than 4% and an improvement in reactivity estimation by 24 pcm to 101 pcm for fuel pin problems. Whereas the computed results show improvement in reactivity estimation by 59 to 134 pcm for fuel assembly problems and 300 pcm for MOX fuel respectively. The study of fuel involving minor actinide for future application such as transmutation in power reactors, resonance self-shielding treatment for them becomes remarkable issue for criticality and isotope depletion. Resonance treatment for important minor actinides and other resonant absorbers has been carried out with subgroup method. The updated multi-group cross sections and subgroup data is replaced with existing multi-group library. The resonance interaction of uranium with important minor actinides has been accommodated through modified interference treatment by interference correction in subgroup method. The results for cross sections and multiplication factor for pin and assembly problems showed improvement from systematic resonance treatment for resonant absorber nuclides including 237Np and 243Am. The multi-group cross section data library has been generated for sodium cooled fast reactor. The well-known established procedure has been used to obtain multi-group cross section data set with new group structure using specific flux weighting spectra. The multi-group calculations for sodium cooled fast reactor have shown improved results for multiplication factor with new library compared to the existing multi-group library. The self-shielding was performed for uranium isotopes by the subgroup method with and without interference correction factors. The subgroup data was generated to obtain the effective cross section for fast reactor assembly case. The effective cross section obtained using subgroup data and interference correction factors has shown results similar to reference cross sections.