| Deterministic methods in solving the steady-state neutron transport equation usually require first determining the coefficients of the terms in the transport equation.The coefficients are usually the total macroscopic cross section of the material,the macroscopic scattering cross section,etc.,and are generally referred to as multi-group cross sections.The determination of the multi-group cross section is often related to the neutron flux in the actual problem,which in turn is the solution of the transport equation.Therefore,the determination of the multi-group cross section and the solution of the neutron flux is an iterative process.In the calculation of practical problems,the process of making a multi-group cross section library that accompanies the deterministic approach introduces more approximations,which have a greater impact on the results of the transport calculations.In order to improve the accuracy and efficiency of multi-group cross section libraries when applied to engineering calculations,this paper investigates the computational methods of multi-group cross section libraries in the process of making and preprocessing,as well as the optimization study of the energy group structure based on the demand for multi-group cross section data when solving neutron transport equations by different deterministic methods.First of all,the input parameter processing methods in the production of the multi-group cross section library are studied for the deterministic approach based on the NJOY program and its methods.The neutron thermal scattering law calculation method is studied for the thermal energy region,and the light water of the pressurized water reactor moderator is used as an example to calculate the phonon spectrum of light water at several temperature points by using the first nature principle,and the phonon spectrum data at arbitrary temperature points are obtained by interpolation calculation,and then the thermal scattering law calculation is carried out to obtain the thermal scattering cross section at arbitrary temperature points,and the preliminary verification analysis is carried out,and the results show that this paper has a good understanding of the thermal neutron spectrum at arbitrary temperature points of arbitrary materials.The results show that the thermal neutron scattering cross section calculation method for arbitrary temperature points of arbitrary materials is reliable and more applicable than the traditional processing method.The method of processing the fuel burnup data needed to supplement the multi-group cross section library for the collision probability method used in reactor core physics calculations is investigated,and a typical pressurized water reactor is used as an example to produce the matching fuel burnup data.The analysis of the single lattice for light water moderated uranium metal,light water moderated uranium oxide,and light water moderated MOX fuels,and the calculation results of uranium oxide lattice burnup in the Akio YAMAMOTO benchmark problem show that the paper is reliable for the principles of selecting the main fission products required for typical pressurized water reactor burnup calculations and for the calculation of burnup chains,fission product yields,and pseudo-fission product reaction cross sections.Second,during the pre-processing of the application of the multi-group cross section library,transport corrections,resonance self-shielding calculations and absorption balance calculations are required,otherwise the transport procedure cannot be used directly.This is due to reproducing the evaluated nuclear data as completely as possible to ensure its generality.However,the above processing is required before providing it to the discrete ordinate method transport program for calculation.In this paper,the BHS method and diagonal transport correction method are used for transport correction,the Bondarenko method is used for resonance processing,and the method of subtracting the neutron-producing reaction cross sections such as(n,2n)and(n,3n)from the absorption balance,and then accumulating the cross sections of each reaction channel while performing balance calculations to correct the total cross section,and the INTERFAX multi-group constant library preprocessor is developed accordingly.The critical and shielding benchmark with simple geometry and material composition were selected for preliminary analysis and applied to the calculation of the shielding problem of CSNS target station.The results of the benchmark calculations show that the method adopted by INTERFAX is comparable to TRANSX in accuracy,and the results of the neutron shutter and biological shielding calculations of the CSNS target station are compared with those of the U.S.SNS shielding calculations with the same trend of order of magnitude decrease in the total dose equivalent rate,indicating that the multi-group cross section library preprocessing method and procedure adopted based on this paper can be applied to engineering calculations.In addition,the energy group structure of the multi-group cross section library matched with the deterministic method has a significant impact on the shielding calculation.When the number of energy groups is small,the computation time is less,but the accuracy may be poorer to meet the target requirements of the calculation;finer energy groups perform better in terms of accuracy,but consume more computation time.Therefore,a balance between these two factors needs to be found when selecting the energy group structure.In this paper,the resonance phenomena of Fe-56 nuclides in the energy range of 1.1 keV～3.1164 MeV are analyzed for the shielding problem of iron-based materials.And the energy group division of the resonance region is further refined on the basis of the existing VITAMIN 199 group structure by inserting 70 points in this energy range to obtain the 269 group structure based on the resonance peak characteristics of Fe-56 total cross section.Three iron spherical shell benchmarks from the SINBAD shielding benchmark library were selected for comparison with three common energy group structures of VITAMIN 199,ABBN 299 and XMAS 172,and these four energy group structures were applied to evaluate the irradiation supervisory tube detector specific activity and RPV fast neutron flux calculation based on a real reactor model and data of a domestic pressurized water reactor nuclear power plant.The results show that the calculation accuracy of the refined 269 group structure is closer to the measured value,and the results of 269 group are more realistic to the physical process of neutron transport calculation in the resonance region.In summary,this paper investigates the process of generating,pre-processing and applying the multi-group cross section library.In particular,the processing of thermal neutron scattering cross sections,the absorption balance calculation and the optimization of the energy group structure in the pre-processing of multi-group cross section library applications are studied,and the relevant processing methods are implemented through programming.Through a large number of benchmark validation and engineering calculations,it is shown that the method studied in this paper has higher accuracy and wider applicability in engineering calculations,especially shielding calculations,compared with the traditional multi-group cross section library production and pre-processing,and has certain academic research value and engineering reference significance for deterministic methods supporting multi-group cross section libraries applied to practical problems. |
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