Recent fuel anomalies at a Swiss power plant have motivated the development at PSI of the "Cycle Check-Up" (CHUP) methodology which consists in performing Monte Carlo transport calculations using nuclide compositions and thermal-hydraulic boundary conditions coming from reference validated core simulator models. The paper documents a successful effort to verify the proper transfer of information between the core simulator and two Monte Carlo codes involved in CHUP. Two calculation schemes based on both Serpent and MCNP produce consistent results. Considering 2D lattice models of modern BWR fuel assemblies designs, discrepancies lower than 500pcm for k-inf or 0.5% root mean squared for fission rate distributions are obtained with respect to CASMO-5, for the range of exposures expected in a Swiss BWR reactor.The capability of CHUP to extract more spatially resolved information (fuel composition at the sub pin level) is demonstrated. It leads to substantial impact at the rod periphery with an over prediction of the fission rates by up to 25% at 45 MWd/kg compared to using pin-wise information. The limitations of CASMO-5 to accurately predict the radial fission rate profiles within a pellet are also illustrated.For the modelling of larger models with CHUP and MCNP, considering local fuel composition lead to un-manageable runtimes and memory footprints. Using On-The-Fly Doppler broadening and a memory efficient mode, it is possible to carry out such calculations with Serpent. The memory footprint of the Serpent calculations is mainly driven by the storage of material data which appears unnecessarily large and could be reduced.