Tritium distributions in castellated structures of Be limiter tiles from JET-ITER-like wall experiments

Abstract

Tritium (T) retention in the castellated structure of beryllium limiters used in JET with the ITER-like wall (ILW) during the first (ILW1), third (ILW3) and all three (ILW1-3) campaigns were examined and evaluated. Tritium was deposited on the surfaces inside the castellation grooves together with deuterium, beryllium, oxygen, carbon and small amounts of metallic impurities such as nickel, copper and tungsten. The T content after ILW1 was greater than after ILW3. This is attributed to the steadily decreasing amount of carbon impurities in JET from campaign to campaign. The majority of T was retained in shallow regions in the grooves, up to 2 mm from the entrance to the gap. It was comparable on all sides of the castellation, i.e. no difference has been detected between the toroidal and poloidal gaps. Secondly, the T retention in the gaps was similar on all specimens independently on their position in the tokamak while the retention on the plasma-facing surfaces was clearly depended on the tile position. The tritium deposition patterns in the castellation were also compared with the deuterium distribution determined in earlier studies.

Mechanism of hydrogen isotope exchange for tritium removal in plasma-facing materials: a multi-scale investigation

“…Its retention in materials leads not only to fuel loss but also to radioactive contamination. The safety of future fusion devices is greatly dependent on the amount of T retention in materials [3][4][5][6][7]. Although tungsten (W), one of the preferred PFMs, has the advantage of low HIs retention [8][9][10][11][12], there will still be a large amount of T retained in W due to a significant number of trapping sites (TS) generated under the irradiation of 14.1 MeV fusion neutrons.…”