OFELIA: An OpenMC-FEniCSx coupling for neutronic calculation with temperature feedback

The state of an operating nuclear reactor depends on several physical phenomena that coexist and are interdependent: they can be taken into account simultaneously by adopting a multi-physics approach, allowing a higher level of detail of the system's properties. Neutron physics and thermal hydraulics are of great importance in this framework, their interdependence being the most fundamental coupling effect in nuclear reactors, as their interaction determines the power and temperature profiles, being quantities of interest during the design and safety analysis phases. This work focuses on developing a fully open-source multi-physics and multi-scale tool capable of determining the temperature profile of a characteristic fuel pin of a PWR when the power generated by the system is known. This methodology is implemented in a Python environment coupling the open source library FEniCSx for the thermal-hydraulic analysis with the OpenMC Monte Carlo code for the description of the fissionable system: regarding the former, FEniCSx handles the thermal calculations, whereas a 1D model is used to predict the axial coolant temperature distribution. The coupling applies an explicit method, whose convergence is based on a Picard scheme, using an adaptive relaxation scheme. This coupling strategy is compared with literature data, providing a good agreement with a fully Multi-Physics solver.