Effects of a multi-physics coupling approach in monte carlo burnup calculations

In nuclear reactor analysis, a relevant challenge is to achieve a suitable global description of nuclear systems through the coupling between neutronics and thermal-hydraulics. Indeed, a multi-physics approach improves the reactor safety analysis and the design of different types of nuclear systems; in addition, it allows the investigation of physical effects at different scales of time and space. In this context, a challenging task is the development of multi-physics tools to study the fuel burnup: these tools could improve the fuel management and estimate the amount of long-lived radionuclides in spent nuclear fuel for current and innovative nuclear reactors. This paper presents the study of a burnup analysis with the Serpent Monte Carlo code, that implements an external interface for the coupling with OpenFOAM, in order to import material temperatures and density field calculated by a thermal-hydraulics solver. In particular, we carried out a burnup analysis for the entire fuel cycle of a simplified fuel cell, composed by an UO2 pin surrounded by water. We evaluated the effects of the multi-physics coupling by comparing the results from simulations that adopt uniform distributions of material temperatures and densities, to those obtained with the multi-physics coupled approach. Particularly, we will show the differences in nuclide densities and the results from the transport calculation (neutron fluxes, reaction rates and criticality).