A guide to the AUS modular neutronics code system.

Abstract

A general description is given of the AUS modular neutronics code system which may be used for calculations of a very wide range of fission reactors fusion blankets and other neutron applications. The present system has cross-section libraries derived from ENDF/B-IV and includes modules which provide for lattice calculations one-dimensional transport calculations and one two and three-dimensional diffusion calculations burnup calculations and the flexible editing of results. Details of all system aspects of AUS are provided but the major individual modules are only outlined. Sufficient information is given to enable other modules to be added to the system.