快堆伪裂变产物截面和角分布的计算
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摘要
为满足我国示范快堆研究的需要并解决以往伪裂变产物截面数据偏小的问题,需重新研制一种制作伪裂变产物数据的方法,为制作多个裂变核的伪裂变产物全套中子数据提供基础。本文用浓度加权求和的方法计算伪裂变产物截面、微分截面和双微分截面。在挑选核素的过程中提出贡献法,即利用裂变率加权产额和吸收截面(反应道MT=27)得到产物核对反应堆的贡献值,从而量化了挑选核素的过程,提高了计算的准确性。最后以CENDL_NP库为主要数据来源,TENDL库数据为补充,制作出了一套~(235) U的伪裂变产物截面数据,通过与以往计算结果比较证明了上述方法的优越性和实用性。
In order to meet the need of Chinese Demonstration Fast Reactor and solve the problem that former calculated cross sections being smaller than the real values,it is necessary to develop a new method of calculating pseudo-fission products data,thus providing a foundation for generating complete neutron data of pseudo-fission products.The cross sections,angular distributions and double differential cross sections of pseudofission products were calculated by concentration-weighted su mmation.In the process of selecting nuclides,contribution method was proposed.The averaged fission yields and the absorption cross sections(reaction channel MT=27)were used to calculate the contributions of fission products to reactor,which quantifies the process of selecting nuclides and improves the accuracy.Finally,a complete pseudo-fission products data set for ~(235) U was generated,using CENDL_NP library as main data source and TENDL library as supplementary.By comparing these data with former calculated results,the superiority and practicability of above method are validated.
In order to meet the need of Chinese Demonstration Fast Reactor and solve the problem that former calculated cross sections being smaller than the real values,it is necessary to develop a new method of calculating pseudo-fission products data,thus providing a foundation for generating complete neutron data of pseudo-fission products.The cross sections,angular distributions and double differential cross sections of pseudofission products were calculated by concentration-weighted su mmation.In the process of selecting nuclides,contribution method was proposed.The averaged fission yields and the absorption cross sections(reaction channel MT=27)were used to calculate the contributions of fission products to reactor,which quantifies the process of selecting nuclides and improves the accuracy.Finally,a complete pseudo-fission products data set for ~(235) U was generated,using CENDL_NP library as main data source and TENDL library as supplementary.By comparing these data with former calculated results,the superiority and practicability of above method are validated.
引文
[1]徐銤.我国快堆技术发展的现状与前景[J].中国工程科学,2008,10(1):70-77.XU Mi.The status and prospects of fast reactor technology development in China[J].China Engineering Science,2008,10(1):70-76(in Chinese).
[2]刘廷进,刘萍.快堆伪裂变产物数据的计算[J].核科学与工程,2002,22(1):71-78.LIU Tingjin,LIU Ping.Calculation of pseudo fission-product neutron data for fast reactor[J].Nuclear Science and Engineering,2002,22(1):71-78(in Chinese).
[3]GRUPPELAAR H,PIJLGROMS B J,KLOOSTEMAN J L,et al.Status of pseudo fission-product cross-sections for fast reactors,NEA/WPEC-17[R].[S.l.]:[s.n.],1998.
[4]胡文超.基于快堆嬗变的伪裂变产物截面制作[J].核动力工程,2013,34(6):31-34.HU Wenchao.Production of pseudo fissionproduct for transmutation of fast reactor[J].Nuclear Power Engineering,2013,34(6):31-34(in Chinese).
[5]DEVAN K,GOPALAKRISHNAN V,MOHANAKRISHNAN P,et al.Preparation of multigroup lumped fission product cross-sections from ENDF/B-Ⅵfor FBRs[J].Annals of Nuclear Energy,1998,25(1-3):161-179.
[6]WILSON W B,ENGLAND T R,VAN K A.Class on CINDER’90code and associated programs and data for transmutation calculations[R].US:Los Alamos National Laboratory,1998.
[7]MACFARLANE R E,KAHLER A C.Methods for processing ENDF/B-Ⅶwith NJOY[J].Nuclear Date Sheets,2010,111:2 700-2 739.
[8]HERMAN M,TRKOV A.ENDF-6 format manual,data format and procedures for the evaluated nuclear data file ENDF/B-Ⅵand ENDF/B-Ⅶ[R].New York:Brookhaven National Laboratory,2009.
[2]刘廷进,刘萍.快堆伪裂变产物数据的计算[J].核科学与工程,2002,22(1):71-78.LIU Tingjin,LIU Ping.Calculation of pseudo fission-product neutron data for fast reactor[J].Nuclear Science and Engineering,2002,22(1):71-78(in Chinese).
[3]GRUPPELAAR H,PIJLGROMS B J,KLOOSTEMAN J L,et al.Status of pseudo fission-product cross-sections for fast reactors,NEA/WPEC-17[R].[S.l.]:[s.n.],1998.
[4]胡文超.基于快堆嬗变的伪裂变产物截面制作[J].核动力工程,2013,34(6):31-34.HU Wenchao.Production of pseudo fissionproduct for transmutation of fast reactor[J].Nuclear Power Engineering,2013,34(6):31-34(in Chinese).
[5]DEVAN K,GOPALAKRISHNAN V,MOHANAKRISHNAN P,et al.Preparation of multigroup lumped fission product cross-sections from ENDF/B-Ⅵfor FBRs[J].Annals of Nuclear Energy,1998,25(1-3):161-179.
[6]WILSON W B,ENGLAND T R,VAN K A.Class on CINDER’90code and associated programs and data for transmutation calculations[R].US:Los Alamos National Laboratory,1998.
[7]MACFARLANE R E,KAHLER A C.Methods for processing ENDF/B-Ⅶwith NJOY[J].Nuclear Date Sheets,2010,111:2 700-2 739.
[8]HERMAN M,TRKOV A.ENDF-6 format manual,data format and procedures for the evaluated nuclear data file ENDF/B-Ⅵand ENDF/B-Ⅶ[R].New York:Brookhaven National Laboratory,2009.