Analysis of SPECTROX method of multigroup spectra calculation in unitary reactor cells

The thermal neutron spectrum in a lattice cell is strongly space-dependent. In addition, in many situations, as for example in core design calculations, a more precise energetic and spatial representation of the flux is needed, which cannot be provided by few group diffusion theory. In such cases, the well-known SPECTROX method, employing diffusion theory in the moderator, where it is supposed sufficiently accurate, and collision probability theory in the fuel, together with appropriate interface current relations for assuring neutron conservation, has been widely used by WIMS as well as other codes. In this work, the approximations leading to the SPECTROX equations are reviewed and the calculated average fluxes in the fuel are compared with accurate values obtained from the solution of the transport equation by the FN method. (author)