Interface-Flux Nodal Transpon Method

Abstract

The development of the interface-flux nodal (ITN) method is presented to determine the flux distribution in reactor cells, cores and shielding. The method offers geometric flexibility, high order of spatial expansions of the node interior sources and the node surface quantities. The surface-integral formulation is reduced to response matrix-like global equations through coupling coefficients which axe generalized expressions for escape and transmission probabilities. The spatial distribution of the neutron flux may be represented by high order polynomials using geometric basis functions determined by a least-square minimalization technique. The angular dependency of the outgoing/incoming surface flux components is treated using a general DP, expansion and the spatial variation is handled with a boundary element technique. The scattering iterations are eliminated by using an explicit expansion of the scalar fluxes. Based on the IFN method a computer code has been developed capable of calculating fixed-source and eigenvalue problems. Test problems for 1-D and 2-D X-Y and hexagonal geometries are presented including comparison with other techniques to demonstrate the validity and accuracy of the IFN method.