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Abstract
A new nodal transport method has been developed so as to perform accurate transport calculations in three-dimensional (3-D) hexagonal geometry. In the present method the neutron angular distributions of intra-node fluxes and transverse-leakage are represented using the SN quadrature set, and the spatial distribution of neutron source is approximated by a quadratic polynomial expansion. Additionally, the authors have applied the nodal-equivalent finite difference algorithm to 3-D hexagonal geometry in order to establish a stable and efficient iterative scheme. The present method has been applied to a fast reactor benchmark problem. The results of the present method agree well with those of a Monte Carlo method, the difference in kea being less than 0.05%Δk. This shows high accuracy of the present method.
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