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

Abstract

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.

Neutronic analysis of sodium‐cooled fast reactor design with different fuel types using modified CANDLE shuffling strategy in a radial direction

“…Neutronic review is the most crucial aspect in nuclear reactor design and simulates the actual neutron behavior inside the reactor core. The behavior of the neutron, as they react with materials, can be expressed as the neutron flux distribution, whose mathematical representation is given by the neutron transport equation 7,8 . Neutron transport involves the distribution, movement, and interaction of the neutron with the nucleus in the reactor core.…”