研究報告16 - 1 解説 - Review
JT-60U重水素実験におけるトリチウム処理
Tritium Processing in JT-60U
宮 直之、正木 圭
日本原子力研究所那珂研究所
〒311-01 茨城県那珂郡那珂町向山801-1
Naoyuki MIYA and Kei MASAKI
Naka Fusion Research Establishment Japan Atomic Research Institute,Naka-machi 801-1, Naka-gun, Ibaraki 311-01, JAPAN
(Received September 4. 1996; accepted January 11. 1997)
Abstract
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.