A study of tritium decontamination of deposits by UV irradiation

Tritium Decontamination from Co-deposited Layer on Tungsten Substrate by Ultra Violet Lamp and Laser

“…However, they were not sufficiently high to cut C=C bond (about 7.2 eV 10) ). In the previous study, 7) it has been reported that only C-H bond on the surface of the co-deposited layer is dissociated by the UV lamp irradiation. The result of this study is consistent with the previous one.…”

Pulsed-laser ablation of co-deposits on JT-60 graphite tile

“…One of the JT-60 open-divertor tiles, which were exposed to 1800 hydrogen discharges from June 1988 to October 1988, with the limiter configuration including 300 lower X-point divertor configura- [17] ps laser [7,19,20] JT-60 [10,12] Xe lamp (172 nm) [5,6,15] TORE-SUPRA [14] tions, was used as a sample. Hirohata et al have revealed that hydrogen was retained homogeneously in the co-deposits of the JT-60 open-divertor tiles with the nearly constant concentration of H/C $0.03 or 1.4 · 10 21 atoms m À2 lm À1 using thermal desorption spectroscopy [23].…”

“…Some advantages of tritium removal using lasers are that it is possible to remove tritium from plasma-shadow regions, the production of hazardous tritiated water is small, and it is not necessary to introduce gases. Many experiments on the removal of hydrogen isotopes from carbon materials have been conducted using several types of light sources with different wavelength, pulse duration, and intensity [3][4][5][6][7][8][9][10][11][12][13][14][15][16][17][18][19][20]. See Table 1 for a summary.…”

“…In the region where I L is less than the threshold intensity I ablation for ablation or vapourization (non-ablation region; NAR), thermal desorption of hydrogen isotopes occurs [3,5,6,9,11,13,17]. When I L is larger than I ablation (ablation region; AR), sublimation of both carbon materials and hydrogen isotopes occurs, and ablative removal starts [3,7,8,10,12,[14][15][16][18][19][20].…”

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

“…It is important to understand the release behavior of tritium trapped to graphite from viewpoints of fuel recycling and radiation safety. Many studies on tritium removal from graphite tile have been performed so far and various techniques such as laser heating method [4][5][6], glow discharge method with hydrogen isotope or oxygen [7][8][9], and oxidation method with pure oxygen gas, air or water vapor [10][11][12][13][14][15][16][17], have been proposed and tried for the graphite tiles used during D-T or D-D operation of tokamak machines such as JET, TFTR or TEXTOR. Most previous experiments have focused on removal of tritium from the co-deposition layer or removal of the co-deposition layer itself at the temperature range from 200°C to 500°C targeting on in situ cleaning.…”