Neutron Mean Free Path in the Slab Reactor Core using One-Dimensional Multi-group Diffusion Equation

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