Calculation of Collision Probability Matrix of Nuclear Fuel Cell as a Function of Neutron Energy Group Using Flat Flux Model

Abstract

One of the methods that widely used in solving neutron transport equations in the nuclear fuel cell is the collision probability (CP) method. The neutron transport is very important to solve because the neutron distribution is related to the reactor power distribution. The important thing in the CP method is the CP matrix calculation, better known as has an important role in determining the neutron flux distribution in the reactor core. This study uses a linear flat flux model in each cell region for each energy group with white boundary condition. Although the type of reactor used in this study is a fast reactor, the matrix calculation still carried out in fast and thermal group energy. The matrix depends on the number of mesh in each cell region. The matrix formed from the mesh distribution will produce a matrix for each energy group. Because the boundary condition of the system is assumed that there are no contributions neutron source from the outside, the sum of the matrix must be less than one. In general, the results of the calculations in this study are following the theory