On the calculation of angular neutron flux in MCNP

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• 作者：Jure Beričič ; Luka Snoj ; ^{luka.snoj@ijs.si" class="auth_mail" title="E-mail the corresponding author}
• 关键词：Angular flux ; Neutron flux ; Neutron transport ; Monte Carlo ; MCNP ; Computer code ; Solid angle partitioning
• 刊名：Annals of Nuclear Energy
• 出版年：2017
• 出版时间：February 2017
• 年：2017
• 卷：100
• 期：part_P2
• 页码：128-149
• 全文大小：5146 K

文摘

Modern Monte Carlo neutron transport codes offer many options for neutron flux and spectra calculations, however, they often lack the option to obtain the angular neutron flux in a region of the problem. The angular flux can also be obtained from deterministic programs, however, it includes biases due to discretization and other physical approximations. Therefore, a novel method for determining the angular neutron flux from the standard output of the MCNP is proposed in this paper. The method was also implemented as a set of Python libraries and tested in several examples. The results were then used to investigate the self-shielding effect in a realistic angular profile of the flux, i.e., the TRIGA research reactor.