Home / Store / Journals / Electronic Articles / Fusion Science and Technology / Volume 60 / Number 4 / Pages 1535-1538
Y. Nobuta et al.
Fusion Science and Technology / Volume 60 / Number 4 / Pages 1535-1538
Format:electronic copy (download)
Tritium retention in plasma facing materials is a primary issue for ITER and next step fusion devices, since it greatly affects its safety and operational schedule. In the ITER, carbon and tungsten are used as divertor materials. In the present study, co-deposited carbon film, tungsten and isotropic graphite were exposed to tritium gas, and then the amount of absorbed tritium was investigated. During the tritium exposure, the partial pressure of tritium gas was kept at 10 Pa. The sample temperature was kept a constant in the range from RT to 573 K. The amounts of absorbed tritium were evaluated by -ray-induced X-ray spectrometry (BIXS). The amounts of absorbed tritium in co-deposited carbon films were one or two orders of magnitude larger than that of polycrystalline tungsten and isotropic graphite. The amount of absorbed tritium for co-deposited carbon film with a high volume density (1.53 g/cm3) was several times larger than that of the film with a low volume density (1.13 g/cm3). The amount of absorbed tritium increased with the temperature. These results indicate that co-deposited carbon films can absorb much larger amount of tritium than tungsten and graphite, and carbon film density affects the amount of absorbed tritium.
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