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.