Calculation of Cross Section and Angle Distribution for Pseudo-fission Product in Fast Reactor

ZHANG Chong; WU Hai-cheng; GE Zhi-gang; LIU Ping; WU Xiao-fei

doi:10.7538/yzk.2017.51.09.1550

Atomic Energy Science and Technology > 2017 > 51(9): 1550-1556. > DOI: 10.7538/yzk.2017.51.09.1550

ZHANG Chong, WU Hai-cheng, GE Zhi-gang, LIU Ping, WU Xiao-fei. Calculation of Cross Section and Angle Distribution for Pseudo-fission Product in Fast Reactor[J]. Atomic Energy Science and Technology, 2017, 51(9): 1550-1556. DOI: 10.7538/yzk.2017.51.09.1550

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Calculation of Cross Section and Angle Distribution for Pseudo-fission Product in Fast Reactor

China Nuclear Data Center, China Institute of Atomic Energy, Beijing 102413, China

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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 pseudo-fission products were calculated by concentration-weighted summation. 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 ²³⁵U 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.
- pseudo-fission product,
- selecting nuclide,
- concentration,
- weighted summation

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Calculation of Cross Section and Angle Distribution for Pseudo-fission Product in Fast Reactor

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