Neutron Mean Free Path in the Slab Reactor Core using One-Dimensional Multi-group Diffusion Equation
Analysis of the neutron mean free path in the slab reactor core has been carried out using one-dimensional multi-group diffusion equation. This study aims to determine the neutron mean free path in the slab reactor core with the neutron diffusion coefficient calculation using macroscopic cross-section data in the nuclear fuel cell level and the neutron flux distribution. The type of reactor used in this research is a fast reactor with nuclear fuel is uranium-plutonium nitride (U-PuN). The neutron mean free path is calculated for 70 energy groups of neutron by dividing the energy groups, namely the fast energy group, the intermediate energy group and the thermal energy group. The results showed that the neutron mean free path value for U-235 and Pu-239 fuels were obtained almost the same in all energy groups, namely in the fast energy group ranging from 0.11.10-2 to 0.17.10-2 cm, in the intermediate energy group 0.16.10-2 to 1.78.10-2 cm, and in the thermal energy group 0.4.0-2 to 8.04.10-2 cm. The neutron mean free path value for U-238 fuel is much smaller than that for U-235 and Pu-239 fuel, ranging from 0.03.10-2 to 0.36.10-2 cm. These results can be confirmed, because U-238 fuel is a fertile material.
Keywords: Neutron mean free path, diffusion equation, neutron flux, slab reactor core