Analysis of Thermal-Hydraulic Behaviour of the Molten Salt Nuclear Fuel

This paper presents a preliminary approach to thermo-hydraulics of the molten salt, which plays the role of both heat generator and coolant in the Molten Salt Reactor (MSR). This kind of nuclear reactor represents one of the "Generation IV International Forum" concepts that can be used for actinides burning, production of electricity, production of hydrogen, and breeding of nuclear fuel. Physics of circulating nuclear fuels, as the molten salt, is featured by a strong coupling between neutronics and thermo-hydrodynamics. In the present study, analyses are performed assuming that the neutronic term is decoupled from fluid dynamics and appears like an energy source term, and taking into account the thermodynamic and transport properties of the molten salt as well as its local flow conditions and heat transfer. Even if this assumption simplifies the equations to be solved, the thermo-hydrodynamic behaviour of the molten salt remains complex. The graphite-moderated channel type molten salt breeder reactor based on a previous research at Oak Ridge National Laboratory (ORNL) is considered: a preliminary study of the heat transfer and pressure losses in a typical MSR core channel is proposed referring to a simple axial-symmetric cylindrical geometry with the aim to investigate the specific behaviour of such system as well as to test and compare two different commercial computer codes – namely, COMSOL® (Multiphysics finite elements software) and FLUENT® (Computational Fluid Dynamics finite volumes software) – on the basis of an analytic framework, in view of their adoption for more realistic, design-oriented and multi-physics simulations.