To assess the overall pin-by-pin fission rate distribution, the collective per cent error measures were investigated. absorption/fission reaction rates per isotope, and pin-by-pin group integrated fission rates on 1/8 fraction of the geometry are determined. The data that are used in this paper are based on the benchmark by the NEA Nuclear Science Committee (NSC).
The aim of the present paper is to assess the calculations of pin- by-pin group integrated fission rates within MOX/UO2 Fuel assemblies using the Monte Carlo code MCNPX2.7c with two sets of the available latest nuclear data libraries used for calculating MOX-fueled systems. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP is either command line or interactively driven. Processor Output files, and two user supplied tables. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells.
MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment.
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