A finite fourier transform method for three dimensional steady state reactor core calculations

A method based on finite Fourier transform technique has been developed to solve the steady state multigroup neutron di€usion equation for reactor core calculations in X-Y/X-Y- Z geometry. In the present work, the neutron source in a node is approximated by a 5 point quadratic in X-Y plane and a quadratic in axial direction. The partial currents on the surfaces of the node have been assumed to be constant and equal to their average values. The equa- tions cast in response matrix form are solved by standard fission source iterative approach. The outer iterations are accelerated using a coarse mesh rebalancing scheme. The algorithm has been implemented in a computer code finfor-sqr. The code has been validated by ana- lysing a few benchmark problems