Neutron Distribution in the Nuclear Fuel Cell using Collision Probability Method with Quadratic Flux Approach. (March 2017) - Shafii, M A; Fitriyani, D; Tongkukut, S H J; Abdullah, A G
Abstract: To solve the integral neutron transport equation using collision probability (CP) method usually requires flat flux (FF) approach. In this research, it has been carried out in the cylindrical nuclear fuel cell with the spatial of mesh with quadratic flux approach. This means that the neutron flux at any region of the nuclear fuel cell is forced to follow the pattern of a quadratic function. The mechanism may be referred to as the process of non-flat flux (NFF) approach. The parameters that calculated in this study are the k -eff and the distribution of neutron flux. The result shows that all parameters are in accordance with the result of SRAC.