Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

Release behavior of tritium from the graphite tiles used at dome top and inner dome wing in JT-60U was investigated by the thermal desorption method in dry argon, argon with oxygen and water vapor, or argon with hydrogen. It was found that approximately 20–40% of total tritium is left in graphite even after heating to the high temperature above 1000 °C in dry argon. The residual tritium could be removed by exposing the graphite tile to oxygen with water vapor or hydrogen at the high temperature above 1000 °C. The tritium retention of the dome top tile was quantified as 84–30 kBq/cm2. The inner dome wing tile had a steep tritium distribution from 8 to 0.1 kBq/cm2. It is observed that a measurable amount of tritium existed in the deep site of the graphite tile.