TRANSPORT CALCULATIONS OF NEUTRON ENERGY-SPECTRA IN A GRAPHITE CYLINDER WITH A D-T SOURCE

Neutron energy spectra resulting from the transport of 14. 7 MeV neutrons from a collimated D-T source through a graphite cylinder, have been calculated with the discrete-ordinates 2-D transport DOT 4. 2 code, with multigroup cross-sections generated using the NJOY code from the ENDF/B (IV&V) libraries. The results confirm the conclusion of Goldfeld et al. (1985), that energy spectra at mesh points close to the axis of the system, in front of the collimated beam, consist mainly of one-collision contributions of elastically or inelastically scattered neutrons. Investigation of the dependence of the calculated spectra on the order of truncation of the Legendre polynomials expansion of the flux and of the cross-sections (i. e. the order of scattering) is reported.