摘要
从ENDF/B-VII库提取数据,通过NJOY程序对快堆中生成的裂变产物核素进行模块加工,利用Matlab进行编程对NJOY程序计算得到的数据进行再次加工处理,得到235U核素快堆嬗变的多群伪裂变产物截面数据,然后用MCNP程序对设计的快堆进行计算得到中子能谱图,并用中子能谱对MCNP程序生成的多群截面进行并群。把生成的数据与NJOY程序生成的数据进行对比验证表明,经过处理的截面数据可以用于快堆的燃耗计算。
Firstly, the NJOY nuclear Date Processing System is used to process the fissional nuclides from ENDF/B-VII for sections of fast reactors cross, We use the Matlab software to process the cross section date from N JOY and get multigroup pseudo fission-product for fast reactor transmutation. The fissile parent nuclides are U-235.The cross section data include total cross section, elastic scattering cross section, radioactive capture cross section, the cross section of producing an alpha particle （n, a）, the cross section of producing a proton （n, p）, the cross section of producing one neutron （n, n） and the cross section of producing two neutrons （n, 2n）. The paper uses the neutron energy spectrum to deal with the multigroup cross sections from a fast reactor and validates the multigroup cross section from MCNP and the date form NJOY. It is concluded that the processed cross section date can be used for the accurate calculations of fast reactor trmasmutation.
出处
《核动力工程》
EI
CAS
CSCD
北大核心
2013年第6期1-4,共4页
Nuclear Power Engineering
基金
国家"863"计划资助项目(2009AA050704)
关键词
快堆
嬗变
NJOY
MCNP
伪裂变产物
Fast reactor, Transmutaticn, NJOY, MNCP, Pseudo fission-product
