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

被引：0

作者：

Zhang C. ^{[1
]}

Wu H.-C. ^{[1
]}

Ge Z.-G. ^{[1
]}

Liu P. ^{[1
]}

Wu X.-F. ^{[1
]}

机构：

[1] China Nuclear Data Center, China Institute of Atomic Energy, Beijing

来源：

Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | 2017年 / 51卷 / 09期

关键词：

Concentration; Pseudo-fission product; Selecting nuclide; Weighted summation;

D O I：

10.7538/yzk.2017.51.09.1550

中图分类号：

学科分类号：

摘要：

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. © 2017, Editorial Board of Atomic Energy Science and Technology. All right reserved.

引用

页码：1550 / 1556

页数：6

共 8 条

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