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We present a method for solving the equations of neutron transport with discretized energetic dependence and angular dependence approximated by the diffusion theory. We are interested in the stationary solution that characterizes neutron fluxes within the nuclear reactor core in an equilibrium state. We work with the VVER-1000 type core with hexagonal fuel assembly lattice and use a nodal method for numerical solution. The method effectively combines a whole-core coarse mesh calculation with a more detailed computation of fluxes based on the transverse integrated diffusion equations. By this approach, it achieves a good balance between accuracy and speed.
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