JT-60U重水素実験におけるトリチウム処理

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Tritium retention analysis and tritium concentration measurement have been made during the large Tokamak JT-60U deuterium operations. This work has been carried out to evaluate the tritium retention for graphite tiles inside the vacuum vessel and tritium release characteristics in the tritium cleanup operations. JT-60U has carried out D-D experiments since July 1991. In the deuterium operations during the first two years, about 1.7x1019 D-D fusion neutrons were produced by D(d,p)T reactions in plasma, which are expected to produce ～31GBq of tritium. The tritium produced is evacuated by a pumping system. A part of tritium is, however, trapped in the graphite tiles. Several sample tiles were removed from the vessel and the retained tritium distribution in the tiles was measured using a liquid scintillator. The results of poloidal distribution showed that the tritium concentration in the divertor tiles was higher than that in the first wall tiles and it peaked in the tiles between two strike points of divertor magnetic lines. Tritium concentration in the exhaust gas from the vessel have also been measured with an ion chamber during the tritium cleanup operations with hydrogen divertor discharges and He-GDC. Total of recovered tritium during the cleanup operations was ～7% of that generated. The results of these measurements showed that the tritium of 16-23 GBq still remained in the graphite tiles, which corresponded to about 50-70% of the tritium generated in plasma. The vessel is ventilated during the in-vessel maintenance works, then the atmosphere is always kept lower than the legal concentration guide level of 0.7 Bq/cm3 for radiation work permit requirements

JT-60U重水素実験におけるトリチウム処理

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