Research On The Calculation Method Of High-precision Multi-group Cross-sections For Fast Reactors Based On Continuous Point-wise Cross-sections

Fast reactor is the main reactor type chosen by the fourth generation reactor forum and also the important reactor type in the current development of China.Due to the need to consider the complex resonance self-shielding effect between heavy and medium mass nuclides,anisotropic scattering and hetergeneity of the whole reactor,the traditional multi-group cross section calculation method can not meet the requirements of high precision fast reactor physical design calculation.It is of great significance for the reactor physical design and safety analysis of fast reactor to study the calculation method of multi-group cross sections.There are many nuclides in fast reactors and the resonance phenomena are complex.In general,ultrafine group structure is used to describe the resonance phenomena in detail in fast reactors.In this paper,equal lethargy energy width 1/120 is used to divide energy group structure,and 2082 ultra-fine groups are divided in the energy range of fast reactor to calculate multi-group cross sections.Continuous point cross section data is the basic data of multi-group cross section calculation.This study uses the open source code NJOY to make the continuous point-wise cross section library.Based on the temperature dependent continuous point-wise cross section data,the calculation method of multi-group cross section of fast reactor is studied in view of the physical characteristics of fast reactor and the characteristics of ultra-fine group energy structure.The research content mainly includes resonance calculation of multi-group cross section of fast reactor,calculation of ultra-fine group high-order scattering matrix,calculation of ultra-fine group efficient fission matrix and fission spectrum,calculation of few-group cross section,program development and engineering verification.(1)For the resonance calculation of multi-group cross sections of fast reactors,the neutron energy spectrum calculation of problem dependent nuclides is adopted in this study,which can directly consider the resonance interference problem of multi-nuclides.For the overlapping effect between the resolved and unresolved resonance energy regions of resonance nuclides,a resonance self-shielding factor iteration method based on continuous point cross section is proposed in this study,which can effectively calculate the resonance of resolved and unresolved resonance energy regions.(2)For the calculation of ultra-fine group elastic scattering matrix,This research studies and implements the T-function method,which can directly and quickly calculate the elastic scattering matrix online compared with the traditional method of prefabricated elastic scattering transfer probability database.By comparing with the keff calculation of the elastic scattering matrix calculated by NJOY method in the critical benchmarks,the correctness of the T-function method is verified.For inelastic scattering and threshold energy reactions such as(n,2n),there is basically no resonance,and the correlation with the neutron energy spectrum of the actual calculation system is weak.Therefore,this study adopts the method of prefabrication in advance,and the relevant data can be directly read during calculation,so as to save calculation time.(3)For the calculation of the ultra-fine group fission matrix and fission spectrum,the idea of fracture energy was introduced in this study,and the calculation of the fission matrix was divided into two parts.The fission matrix was calculated in the high-energy region and the fission spectrum related to a single energy was calculated in the low-energy region,which could greatly improve the calculation efficiency without losing the calculation accuracy of the fission matrix.By verifying the fission spectra of the main fission nuclides and comparing with the calculation results of NJOY/MGGC,the correctness of the calculation process of the ultra-fine group fission matrix and fission spectra is verified.(4)For the calculation of few-group cross sections,the equivalent R-Z model transport calculation method is used to obtain accurate neutron flux moments in different regions,so as to carry out condensation calculation and obtain accurate few-group cross sections.Through UO2,MOX,U-TRU-Zr fuel problems,the calculation process of ultra-fine group and small group section is verified.Compared with OpenMC,the relative error of macro cross sections of four resonant reaction types calculated in this study is basically less than 1%.Based on the realization of the above methods,a high precision multi-group section processing code is developed.Before checking the calculation performance of the code,it is also important to select an appropriate evaluated nuclear library.In this paper,the critical and shielding calculation applications of evaluated nuclear library are also studied in detail.The results of shielding and critical verification show that the computational performance of ENDF/B-Ⅶ.1 is reliable.In this paper,the ENDF/B-Ⅶ.1 version was used for subsequent benchmark test work.In this paper,the key technical problems of fast reactor multi-group cross section are studied and the high precision fast reactor multi-group cross section processing program MAGIC is developed.The program can be used in the production of multi-group cross section and few-group cross section nuclear libraries,and can produce the multi-group cross section libraries of ANISN,ISOTXS and other data formats for the determination codes.In terms of fast reactor engineering benchmark verification,45 fast energy spectrum critical benchmarks in ICSBEP critical benchmark manual are used to verify the ANISN format ultra-fine group database generated by MAGIC program.The MAGIC+ANISN program calculates the effective multiplication factor with little difference from the Monte Carlo code.Compared with Monte Carlo code,91%of the benchmark questions had absolute error less than 300pcm.By using ZPR-6/7 and ZPR-9 fast reactor full reactor benchmarks,the performance of ISOTXS format multi-group and few-group cross section libraries developed by MAGIC program in the neutron computation of the determination theory program is verified.The difference between the neutron fluxes of 33 groups in the inner and outer cores calculated by MAGIC+TWODANT program and that calculated by Monte Carlo program is small,and the relative error is basically within 5%.The RBEC-M is used to verify the power distribution of the fast reactor core for the few-group cross section in ISOTXS format generated by MAGIC program.The relative error of MAGIC+DIF3D program is less than 3%compared with Monte Carlo code.The effectiveness of MAGIC,a multi-group cross section processing program developed for fast reactor with high precision,is verified for the calculation of power distribution of fast reactor cores.