Criticality calculation of a homogenous cylindrical nuclear reactor core using four-group diffusion equations

In this study, we present a general equation for Finite Difference Method Multi-group Diffusion (FDMMD) equations of a cylindrical nuclear reactor core. In addition, we developed an algorithm which we called TUNTOB for solving the FDMMD equations, determined the fluxes at each of the mesh points and calculated the criticality of the four energy group. This was with a view to using the four-group diffusion equations to estimate the criticality of a cylindrical reactor core that will be accurate and locally accessible for nuclear reactor design in developing countries. The multi-group diffusion equations were solved numerically by discretization using the Finite Difference Method (FDM) to obtain a general equation for a cylindrical reactor core. The fluxes at each mesh point and the criticality of the four energy group were then determined. From the results obtained, we observed that an increment in iteration led to an increase in the effective multiplication factor ( ) with a corresponding increase in the computation time. A maximum effective multiplication factor was reached when the number of iteration was 1000 and above. Having established the optimal number of iterations, the effects of the mesh sizes on the computation examined revealed that the values of 'keff' increases as the mesh sizes becomes smaller until an optimal mesh size of 1 x 1 cm2 was reached and further decrease in mesh sizes gave no further improvement in the value of 'keff'. The Study concluded that the accuracy in the values of 'keff' and the smoothness of the neutron distribution curves in 3-D representations depend on the number of mesh points.