Development of a computer code for neutronic modeling of reactor core with rectangular fuel assembly

In this study, we have reformulated the AFEN method for diffusion equations and implemented in a version of the AFEN method code MGHANSP3 for rectangular-z geometry. For increasing the speed of calculations, the coarse group rebalancing (CGR) method has been utilized. Finally, we developed a computer code, named SDANM, using C# programming language. This code takes few group macroscopic cross sections and calculates the effective multiplication factor, reactivity, multi-groups fluxes, power density distributions and power peaking factor in reactor cores with square geometry.The solution accuracy is examined for IAEA benchmark problem. The numerical results illustrate that the SDANM code is an accurate code for calculating multiplication factor and power density distributions

کلیدواژه ها:

Diffusion Equation ، Nodal Method ، AFEN ، Numerical method ، Nuclear computer code