A method based on finite Fourier transform technique has been developed to solve the steady state multigroup neutron diffusion 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 equations 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 analysing a few benchmark problems.