Calculation of Pij matrix of one-dimensional neutron transport in the slab geometry of the nuclear fuel cell using Collision Probability (CP) method has been done. Pij matrix is one of important parameters within the distribution of neutron flux in the nuclear fuel cell. The CP method is the most efficient methods to solve the neutron transport equation in the reactor core. The study is focused on neutron interaction with nuclear fuel cell of U-235 and U-238 for homogeneous condition. The parameters to calculate the Pij matrix are the cross section of nuclear fuel, width of the region and number of regions. A lattice of slabs have been constructed using void boundary conditions for model of finite system to calculate the collision probabilities. If the Pij matrix has been calculated then neutron flux can be determined. The results show that total value of Pij matrix using CP method for U-235 and U-238 is less than one, respectively. This is in accordance with the definition of void boundary conditions for finite slab geometry. Along with Pij matrix, neutron flux is also appropriate with the reference.
Neutron flux distribution calculation for various spatial mesh of finite slab geometry using one-dimensional diffusion equation
