Simplified transport models for neutron energy spectra evaluations

An important issue in the design of nuclear reactor cores is the determination of multigroup cross sections. This task is accomplished by means of a combined procedure involving a space homogenization and an energy averaging to obtain group-collapsed data that can well reproduce the neutronic interactions in the real multiplying system. For this purpose, it is necessary to determine accurate neutron spectra (distribution in the energy domain of the neutron flux) that are used as weights in the collapsing process. In this thesis the energy aspects are analysed. The neutron transport equation is solved through the use of some simplifications: first of all the asymptotic theory is used in order to eliminate the space dependence of the problem and reduce it to a single parameter; then some methods to approximate the collision density and, consequently, the scattering integral are used. In particular, the classic Fermi continuous slowing down approach, the Groeling-Goertzel and the Selengut-Goertzel methods are considered. Adopting these simplifications, the neutron spectra is evaluated for different moderator media, in slab geometry, using the Pn and the Bn methods. The lethargy variable is discretized for the solution of the transport integral equation. In order to reduce the error, a grid independence analysis is performed for each method considered. This thesis is important because it is a first step in order to determine, in a physical consistent way, the energy grid for multigroup models, on which deterministic solvers of the neutron transport equation are based. The results allow also to get a physical insight into the energy collapsing procedure for the generation of multigroup nuclear data.