Neutronic and thermo-fluid diffusion coupling and simulation in the ELFR reactor

This work shows the safety analysis of the European Lead-cooled Fast Reactor (ELFR) in a refrigerant loss scenario with a coupling of the neutron diffusion equation and the thermal effects of the reactor fuel rods, the which has 8 core zones or rings and is cooled by liquid lead. To obtain the feedback between the effects of neutron diffusion with the thermal effects by Doppler Effect, the effective sections neutron velocity and diffusion coefficients were used depending on the average fuel temperature, which are extracted from the Serpent code. The coupling is carried out by calculating the axial average power density profile by means of the neutron diffusion equation. Said average profile is used to calculate the temperature of an average fuel bar, which is a simplification though an average channel. The neutron diffusion equation model considers a two-dimensional and transient model for a homogeneous core and a single energy group, while the heat transfer equations of the fuel rod are considered one-dimensional, as is the transport equation of heat, these last two also considered in transitory regime. To enrich the analysis, the equations are programmed in Python and with the use of OpenGL libraries a graphical interface of the real-time simulation behavior is developed. (Author)