Two point-source, infinite-medium Monte Carlo codes were written for the IBM 704 computer. The codes simulate respectively gamma-ray and neutron transport from a monoenergetic source. A detailed description, including importance sampling methods, is given. Results showed good agreement with moments-method calculations for the same geometry. Calculations for 6-Mev gamma rays in tin provide data on the contribution of annihilation radiation to both the buildup factors and energy spectra. Calculations for 1-Mev gamma rays in water are also presented. The results of calculations for 6-Mev source neutrons in water show a peak in the energy spectrum which is found to be due to single scattering of source particles from oxygen. (Author).

The time-dependent behavior of the energy spectrum in neutron transport was investigated with a formulation, based on continuous-time Markov processes, for computing [alpha] eigenvalues and eigenvectors in an infinite medium. In this study, a research Monte Carlo code called "TORTE" (To Obtain Real Time Eigenvalues) was created and used to estimate elements of a transition rate matrix. TORTE is capable of using both multigroup and continuous-energy nuclear data, and verification was performed. Eigenvalue spectra for infinite homogeneous mixtures were obtained, and an eigenfunction expansion was used to investigate transient behavior of the neutron energy spectrum.