Neutron flux distribution of slab reactor core using one-dimensional multi-group diffusion equation in the thermal energy region

Neutron flux distribution is an essential parameter in the neutronic analysis of the neutron transport in the nuclear reactor. The multi-group diffusion method is the approach commonly used to solve the neutron transport equation. Based on reactor types, the neutron energy range is classified into two regions, namely fast and thermal energy regions. This study presents the neutron flux distribution in the thermal region of the slab reactor core using a one-dimensional multi-group diffusion equation with the Gauss-Seidel iteration method. The reactor is designed in the form of a fast reactor, and it uses the macroscopic cross-sections of the calculation results in the U-PuN fuel cell level as initial input for 70 energy groups. The library data used is JFS-3-J33 70 energy groups, which is the library data of SLAROM computer codes from JAEA Japan. The computational program calculates the neutron flux distribution only in the thermal group of energy: The calculation of neutron flux distributions is executed using the Gauss-Seidel iteration method. The results indicated that the increase of group energy does not reflect the order of magnitude of a multi-group neutron flux distribution in the thermal region. In the case of neutron interaction with U-235 and Pu-239 nuclide, the pattern of neutron flux distribution for each energy group is not sequential, because both are fissile material. Contrary to the U-238 isotope, almost all group energies coincide with one another in the thermal region. The phenomenon occurs because of U-238 is a fertile material that cannot be directly fissioned.