Tritium control and mitigation is one of the most significant issues in Fluoride Salt-cooled High-temperature Reactors (FHRs). Tritium is primarily generated from neutron activation of the primary coolant, FLiBe, a eutectic mixture of LiF and BeF2 and has the tendency of permeating into the surrounding environment especially under the high operating temperature of the reactor system. To address this issue, a cross-flow tritium removal facility has been proposed to collect and remove molecular tritium, T2, from the primary coolant. Previous simulations and calculations have shown the effectiveness of the cross-flow design. To validate the simulation results, a lab-scale experiment has been set up with a cross-flow tritium removal facility fabricated. For the validation experiment, instead of using a molten salt, a carrier gas is planned to be used and will be premixed with hydrogen before entering the cross-flow facility. A sweep gas will be flowing in the tubes and remove the permeated hydrogen. Samples will be collected from the gas inlets and outlets for component analysis using a gas chromatography. By obtaining the hydrogen concentration change with experiment time in the carrier and sweep gases, the efficiency of the cross-flow tritium removal facility can be derived.
Computational simulations have been carried out for the experiment setup using COMSOL to inform the experiment design. The COMSOL model is validated against a static hydrogen permeation experiment. The tritium removal rate under the planned experimental conditions is predicted. In the experiment, the operation temperature, initial tritium concentration, as well as the gas flow rates will be varied to study their effects on tritium removal. The simulation results will be compared to the experiment results once available.