Abstract
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.
Published Version
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a callDisclaimer: All third-party content on this website/platform is and will remain the property of their respective owners and is provided on "as is" basis without any warranties, express or implied. Use of third-party content does not indicate any affiliation, sponsorship with or endorsement by them. Any references to third-party content is to identify the corresponding services and shall be considered fair use under The CopyrightLaw.
