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The SORGENTINA-RF project aims at developing a 14 MeV fusion neutron source featuring an emission rate in the order of 5–7 × 1013 s−1. The plant relies on a metallic water-cooled rotating target and a deuterium (50%) and tritium (50%) ion beam. Beyond the main focus of medical radioisotope production, the source may represent a multi-purpose neutron facility by implementing a series of neutron-based techniques. Among the different engineering and technological issues to be addressed, the production of incondensable gases and corrosion product into the rotating target deserves a dedicated investigation. In this study, a preliminary analysis is carried out, considering the general layout of the target and the present choice of the target material.

Liquid metal breeding blankets are extensively studied in nuclear fusion. In the main proposed systems, the Water Cooled Lithium Lead (WCLL) and the Dual Coolant Lithium Lead (DCLL), the liquid metal flows under an intense transverse magnetic field, for which a magnetohydrodynamic (MHD) effect is produced. The result is the alteration of all the flow features and the increase in the pressure drops. Although the latter issue can be evaluated with system models, 3D MHD codes are of extreme importance both in the design phase and for safety analyses. To test the reliability of COMSOL Multiphysics for the development of MHD models, a method for verification and validation of magnetohydrodynamic codes is followed. The benchmark problems solved regard steady state, fully developed flows in rectangular ducts, non-isothermal flows, flow in a spatially varying transverse magnetic field and two different unsteady turbulent problems, quasi-two-dimensional MHD turbulent flow and 3D turbulent MHD flow entering a magnetic obstacle. The computed results show good agreement with the reference solutions for all the addressed problems, suggesting that COMSOL can be used as software to study liquid metal MHD problems under the flow regimes typical of fusion power reactors.

The experimental qualification of the Tritium Extraction Unit (TEU) from the LiPb eutectic alloy (15.7 at. % Li), the breeder material of the Water-Cooled Lithium-Lead (WCLL) breeding blanket concept, is one of the fundamental items for the demonstration of tritium balance sustainability for ITER and DEMO fusion reactors. Several technologies have been proposed as TEU, but the selection of the reference technology can be carried out only after the experimental measurement of the tritium extraction efficiency. For this purpose, a dedicated facility, called TRIEX-II, was designed and installed at ENEA Brasimone research centre, Italy. The facility is able to qualify Gas-Liquid Contactor (GLC), Permeator Against Vacuum (PAV) and Liquid-Vacuum Contactor (LVC) technologies at different temperatures, lithium-lead mass flow rates and hydrogen isotopes concentrations. In TRIEX-II, the hydrogen or deuterium, used to simulate tritium, are solubilised inside the LiPb with a dedicated saturator and are then extracted from the liquid metal in the GLC mock-up, which uses pure helium or a mixture constituted by helium and hydrogen as stripping gas and works in the temperature range between 300 and 500 °C.