Accession Number : AD0875371

Title :   A One-Dimensional Few-Group Diffusion Code Package for Fast and Intermediate Reactors.

Descriptive Note : Master's thesis,

Corporate Author : AIR FORCE INST OF TECH WRIGHT-PATTERSON AFB OH SCHOOL OF ENGINEERING

Personal Author(s) : Shosa, Dale W.

Report Date : JUN 1970

Pagination or Media Count : 252

Abstract : Three computer codes have been written to provide a complete package with which one-dimensional few group diffusion calculations may be performed for fast and intermediate reactors. Two sixteen-group microscopic cross-section decks (ANL and LASL) are maintained on punched cards withe the codes. IMX collapses the sixteen-group microscopic cross-sections over zero-moment calculated spectra. FMX collapses the sixteen-group cross-sections over multigroup diffusion calculated spectra. ONEDIF performs eigenvalue searches on geometry and composition. (Author)

Descriptors :   (*NEUTRON TRANSPORT THEORY, COMPUTER PROGRAMS), NEUTRON FLUX, NEUTRON SPECTRUM, CRITICAL ASSEMBLIES, NEUTRON CROSS SECTIONS, DIFFERENCE EQUATIONS, INTEGRAL EQUATIONS, MATRICES(MATHEMATICS), CONVERGENCE, FAST REACTORS, GROUPS(MATHEMATICS), TABLES(DATA).

Subject Categories : Computer Programming and Software
      Fission Reactor Physics

Distribution Statement : APPROVED FOR PUBLIC RELEASE