LAUSR

Abstract For examination of tritium transport in a self-cooled liquid blanket system using fluoride molten salt breeder/coolants, a series of neutron irradiation experiments have been conducted under the low flux neutron environment with an AmBe neutron source at the OKTAVIAN facility of Osaka University in Japan. Tritium release from molten salt FLiNaK (LiF-NaF-KF) at 773 K is evaluated with the neutron source, energies of which are up to 10.7 MeV and the intensity is ∼2 × 106 n/s. To measure the amount of tritium released from FLiNaK and permeating through an Inconel crucible, TF and/or T2O, and T2 are recovered separately using water bubblers and CuO oxidation beds. The amount of tritium recovered for one week in a steady state, 12.0 Bq, has closed to the calculated value, 10.6 Bq, obtained with MCNP6. More than 98% of the recovered tritium from FLiNaK has been TF and/or T2O. The overall mass transfer coefficients for tritium released from a FLiNaK free surface have been evaluated from the transient changes in recovered tritium and those are comparable to the value for HF released from FLiBe previously reported in the literature.