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Wang, J.*; Hatano, Yuji*; Toyama, Takeshi*; Suzudo, Tomoaki; Hinoki, Tatsuya*; Alimov, V. Kh.*; Schwarz-Selinger, T.*
Journal of Nuclear Materials, 559, p.153449_1 - 153449_7, 2022/02
Times Cited Count：3 Percentile：68.71(Materials Science, Multidisciplinary)To study the effect of the content of chromium (Cr) in the tungsten (W) matrix on the vacancy formation and retention of hydrogen isotopes, the samples of the W-0.3Cr alloy were irradiated with 6.4 MeV Fe ions in the temperature range of 523-1273 K. These displacement-damaged samples were exposed to D
gas at a temperature of 673 K. The addition of 0.3% Cr into the W matrix resulted in a significant decrease in the retention of deuterium compared to pure W after irradiation especially at high temperature. Positron lifetime for W-0.3Cr alloy irradiated at 1073 K was almost similar to that for non-irradiated one. These facts indicate the suppression of the formation of vacancy-type defects by 0.3% Cr addition.
Kobayashi, Kazuhiro; Torikai, Yuji*; Saito, Makiko; Alimov, V. Kh.*; Miya, Naoyuki; Ikeda, Yoshitaka
Fusion Science and Technology, 67(2), p.428 - 431, 2015/03
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Disassembly of the JT-60U torus was started in 2010 after 18 years deuterium operations. In the disassembly of the JT-60U torus, tritium retention in the vacuum vessel of the JT-60U is one of the most important safety issues for the fusion reactor. It was very important to study the tritium behavior in Inconel 625 from viewpoint of the clearance procedure in the future plan. After the tritium release for about 1 year at 298 K, the residual tritium in the specimen was released by heating up to 1073 K, and then the residual tritium in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, most of tritium in the specimen was released during 1 year.
Isobe, Kanetsugu; Alimov, V. Kh.*; Taguchi, Akira*; Saito, Makiko; Torikai, Yuji*; Hatano, Yuji*; Yamanishi, Toshihiko
Journal of Plasma and Fusion Research SERIES, Vol.10, p.81 - 84, 2013/02
The distribution of hydrogen trapping sites on W surface exposed with D plasma was examined by the techniques of imaging plate and autoradiography. Recrystallized W specimens were exposed with D plasma at around 495 and 550 K to the same fluence of 10
D/m
. Then, tritium was introduced into specimen by the exposure to tritium gaseous at 473 K. After that, the tritium distribution on W surface was examined by the techniques of imaging plate and autoradiography. From the results of the imaging plate, tritium was found to be highly concentrated within the area exposed with D plasma and the concentration of tritium was slightly varied even in that area. In the autoradiograph of W surface, it was found that tritium concentrated on the grain boundary and blisters.
Isobe, Kanetsugu; Alimov, V. Kh.*; Yamanishi, Toshihiko; Torikai, Yuji*
Toyama Daigaku Suiso Doitai Kagaku Kenkyu Senta Kenkyu Hokoku, 31, p.49 - 57, 2011/00
The limits on tritium inventory in the vacuum vessel and the need for prevention of impurity ingress into plasma, make plasma-surface interaction on tungsten an important issue. It is well known that plasma exposure on tungsten makes some kinds of blisters on the surface and increases the hydrogen inventory. On the other hands, there is a possibility that plasma exposure would change the characteristic of surface and surface region in tungsten and cause the increase of tritium inventory. Tritium concentration in tungsten exposure by low-energy, high-flux D plasma with was examined with BIXS after thermal exposure of tritium gas. The tritium concentration was measured with BIXS. The tritium concentration in surface and surface region was found to be increased by plasma exposure. And its concentration of tungsten exposed at 495 K was estimated to be twice higher than that of as-received tungsten.
He, p)
He resonant nuclear reactionHayashi, Takao; Sugiyama, Kazuyoshi*; Krieger, K.*; Mayer, M.*; Alimov, V. Kh.*; Tanabe, Tetsuo*; Masaki, Kei; Miya, Naoyuki
Journal of Nuclear Materials, 363-365, p.904 - 909, 2007/06
Times Cited Count：11 Percentile：61.1(Materials Science, Multidisciplinary)The absolute concentrations and the depth profiles of deuterium in plasma-facing graphite tiles used in JT-60U were determined by means of the D(He, p)He resonant nuclear reaction. The highest deuterium concentration was found at a plasma-facing surface near the outer pumping slot on the outer dome wing tile, where redeposited layers with thicknesses in the micron or submicron range were observed, indicating deuterium codepostion with carbon. In addition, a high flux of high energy deuterium originating from NBI is expected on the outer dome wing as well as on the dome top tile, and could have some contribution to this area of highest deuterium retention. The deuterium content integrated up to about 16 
m was 
2.5
10
D/m. The depth profile has a broad peak in the atomic ratio of D/C0.05 at a depth of about 2.5 m. This is mainly because D retained in the top surface was replaced by H due to isotope exchange during H discharges, which were carried out to remove tritium from the plasma-facing wall before air ventilation. In erosion dominated areas such as the outer divertor tiles, the amount of deuterium was one order of magnitude lower than that on the outer dome wing tile. In the first wall area, the highest amount of deuterium with a content of 1.010 D/m was found in the upper region, nearest to the plasma. Again implantation of high energy deuterium due to NBI could have some contribution to the high deuterium retention.
