Neutronics Calculations in Large Symbiotic Spallation-Fission Reactor-Assemblies

Abstract

Monte-Carlo-Methods and deterministic standard fission reactor calculation methods have been combined to describe the neutron transport in large spallation-breeding-facilities. A model has been developed which consists of a high energy transport part, several interface codes and standard-fission reactor-programs using SN-transport and diffusion theory. This calculation model allows to analyze neutron transport and fissile fuel breeding in large spallation-fission-assemblies in an effective and cost and time saving way. Running-in periods and equilibrium cycles, fuel shuffling and recycling can be modelled using the same standard procedures as for normal fission reactors. Time consuming Monte-Carlo techniques are reduced to the minimum number necessary and parameter studies in the low energy neutron domain can be done using the same high energy results if changes are not too strong.