Methodology to resolve the transport equation with the discrete ordinates code TORT into the IPEN/MB-01 reactor

Abstract

Resolution of the steady-state Neutron Transport Equation in a nuclear pool reactor is usually achieved by means of two different numerical methods: Monte Carlo (stochastic) and Discrete Ordinates (deterministic). The Discrete Ordinates method solves the Neutron Transport Equation for a set of selected directions, obtaining a set of directional equations and solutions for each equation which are the angular flux. In order to deal with the energy dependence, an energy multi-group approximation is commonly performed, obtaining a set of equations depending on the number of energy groups. In addition, spatial discretization is also required and the problem is solved by sweeping the geometry mesh. However, special cross-sections are required due to the energy and directional discretization, thus a methodology based on NJOY99 code capabilities has been used. Finally, in order to demonstrate the capability of this method, the 3D discrete ordinates code TORT has been applied to resolve the IPEN/MB-01 reactor.