Studies of intra-pin power distributions in operated BWR fuel assemblies using MCNP with a cycle check-up methodology

Abstract At PSI, the reference approach for core safety analyses of all the Swiss operating reactors is based on a deterministic 2-D lattice depletion/3-D two-group nodal diffusion code platform. As a complement to this, a cycle check-up (CHUP) methodology is being established where, for instance, the continuous energy Monte Carlo transport solver MCNP can be plugged-in at any selected operating point of a given reactor/cycle to conduct in-depth examinations of the 3-D flux/power distributions. After briefly outlining the concepts of the CHUP methodology, its application with MCNP to obtain additional 3-D pin power information than provided by current state-of-the-art nodal core simulators is illustrated. More specifically, the pin power and local intra-pin power distributions in operated assemblies, representative of modern BWR fuel designs, are examined. Focus is given to fuel rods characterised by strongly heterogeneous spectral environments. The CHUP/MCNP methodology is applied to evaluate and compare pin and intra-pin power behaviour between different cycles and/or burnup steps, for normal conditions as well as when assuming local perturbations of the coolant and/or mechanical structures. Although only pin-average nuclide compositions are at this stage used, it is shown that already in this case, strong to very-strong azimuthal gradients take place both radially as well as axially and that these can become substantially more pronounced by very small local perturbations, in particular from geometrical changes related to assembly bowing.