On the calculation of angular neutron flux in MCNP

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.