  Neutron diffusion length in reactor grade graphite is measured both
experimentally and theoretically. The experimental work includes Monte Carlo
(MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL
Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal
neutron diffusion length in a reactor moderator of 2m x 2m with slightly
enriched uranium ($^{235}U$), accompanied with a model designed for thermal
hydraulic analysis using point kinetic equations, based on partial and ordinary
differential equation. The theoretical work includes numerical approximation
methods including transcendental technique to illustrate the iteration process
with the FEA method. Finally collision density of thermal neutron in graphite
is measured, also specific heat relation dependability of collision density is
also calculated theoretically, the thermal neutron diffusion length in graphite
is evaluated at $50.85 \pm 0.3cm$ using COMSOL Multiphysics and $50.95 \pm
0.5cm$ using MCNP. Finally the total neutron cross-section is derived using FEA
in an inverse iteration form.
