Application of MCNPX 2.7.D for Reactor Core Management at the Research Reactor BR2

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK•CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full-scale core model. The second method represents fully automatic whole core depletion and criticality calculations in the full-scale 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared with the MCNPX & ORIGEN-S combined method developed at the BR2 reactor department. Testing on criticality measurements at the BR2 reactor is presented.