Discrete nodal method for neutron transport calculations based on unstructured-meshes

An efficient numerical solution of discrete nodal transport method in the triangular node was developed for two-dimensional transport equation based on the discrete nodal transport method, where arbitrary triangles were transformed into regular triangles via a coordinate transformation, and the spatial distributions of intra-node flux and source were approximated by an orthogonal quadratic polynomial. A second-order expansion of transverse-leakage was achieved when the spatial distribution of neutron flux was simulated by a binary quadratic polynomial in a regular triangle. An improved nodal-equivalent finite difference algorithm was adopted to eliminate the divergency due to the small mesh width. The numerical results demonstrate the higher efficiency and accuracy of this method for unstructured neutron transport problem.