TRIEX-II: an experimental facility for the characterization of the tritium extraction unit of the WCLL blanket of ITER and DEMO fusion reactors

Abstract

Abstract 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 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 and liquid-vacuum contactor 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.