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Determination of tritium inventory in carbon divertor tiles used in deuterium plasma experiment by induction heating method
The tritium inventory of the plasma-facing materials in a Large Helical Device (LHD) was analyzed using a thermal desorption system. In this study, we developed an induction heating system for the thermal desorption of tritium from carbon divertor tiles, which has several advantages, including a rapid temperature rise, internal heating via eddy currents, and heating with the original size and shape of the carbon tile. The apparatus was capable of heating a carbon divertor tile to a temperature greater than 1373 K over 20 min. Following a 12-h heating period at 1423 K, the tritium release rate from the carbon tile was 99.7 %. The chemical form of the tritium released from the divertor tiles was approximately 80 % molecular hydrogen, with the remaining tritium in the form of water vapor. The results of the tritium analysis of the divertor tiles suggest that the tritium distribution in the divertor region is not uniform, and that the flux varies from location to location in the LHD. To improve the accuracy of the tritium inventory analysis in an LHD vacuum vessel, it is essential to conduct a tritium inventory analysis of other plasma-facing materials.
Determination of tritium inventory in carbon divertor tiles used in deuterium plasma experiment by induction heating method
The tritium inventory of the plasma-facing materials in a Large Helical Device (LHD) was analyzed using a thermal desorption system. In this study, we developed an induction heating system for the thermal desorption of tritium from carbon divertor tiles, which has several advantages, including a rapid temperature rise, internal heating via eddy currents, and heating with the original size and shape of the carbon tile. The apparatus was capable of heating a carbon divertor tile to a temperature greater than 1373 K over 20 min. Following a 12-h heating period at 1423 K, the tritium release rate from the carbon tile was 99.7 %. The chemical form of the tritium released from the divertor tiles was approximately 80 % molecular hydrogen, with the remaining tritium in the form of water vapor. The results of the tritium analysis of the divertor tiles suggest that the tritium distribution in the divertor region is not uniform, and that the flux varies from location to location in the LHD. To improve the accuracy of the tritium inventory analysis in an LHD vacuum vessel, it is essential to conduct a tritium inventory analysis of other plasma-facing materials.
Determination of tritium inventory in carbon divertor tiles used in deuterium plasma experiment by induction heating method
Masahiro Tanaka (author) / Hiromi Kato (author) / Naoyuki Suzuki (author) / Hiroki Chimura (author) / Hiroaki Yonezu (author) / Suguru Masuzaki (author)
2025
Article (Journal)
Electronic Resource
Unknown
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Tritium distributions on W-coated divertor tiles used in the third JET ITER-like wall campaign
| DOAJ | 2019