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) leads to the following observations: 1. (a) the P₆ or P₇ approximations seem to be adequate enough for flux calculations, with less than 3% error, in spite of the high degree of the source and the cross-section's anisotropy; 2. (b) the calculation error is reduced significantly by increasing the order of scattering from P₄ to P₇, mainly in mesh points close to the axis and of those energies in which the anisotropy of the elastic and discrete level inelastic scattering processes is most pronounced. Finally, the dot 4.2 calculations are compared with Monte Carlo mcnp calculations; both calculated spectra are in a good agreement.