Neutronics Codes and Methodologies Applied for the 3-D Whole Core Evaluations of the BR2 Fuel Cycle

This paper presents an overview of the neutronic codes and methodologies which are used for the management of the current BR2 fuel cycle. The application and comparison of three depletion methodologies are discussed. In the first two methodologies MCNPX is coupled with a 1-D burn up code (ORIGEN-S, or CINDER90 through MCNPX 2.7.D2), which evaluates the evolution of the fuel composition in an infinite lattice. The method in these methodologies is based on preliminary preparation of databases, containing depleted isotopic fuel compositions with depletion step 2% between 0% and 80% fuel burn up, and power peaking factors, which are calculated following the standard irradiation history of the fuel element in the BR2 fuel cycle. The approach taken is to calculate by MCNPX (any version) the total power and the mean burn up in each fuel element at different time depletion steps and then along with the databases to evaluate the 3-D power and 3-D isotopic fuel distributions in the core. The third methodology represents fully automatic 3-D whole core depletion calculations by the Monte Carlo burn up code MCNPX 2.7.D2.