The P_ij matrix and flux calculation of one-dimensional neutron transport in the slab geometry of nuclear fuel cell using collision probability method

¹ Andalas University, Department of Physics, Padang, West Sumatera, Indonesia
² Sam Ratulangi University, Department of Physics, Manado, North Sulawesi, Indonesia
³ Universitas Pendidikan Indonesia, Department of Electrical Engineering Education, Bandung, Indonesia

^* Corresponding author: mashafii@fmipa.unand.ac.id

Abstract

Calculation of P_ij matrix of one-dimensional neutron transport in the slab geometry of the nuclear fuel cell using Collision Probability (CP) method has been done. P_ij 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 P_ij 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 P_ij matrix has been calculated then neutron flux can be determined. The results show that total value of P_ij 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 P_ij matrix, neutron flux is also appropriate with the reference.