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Hirade, Tetsuya; Furuta, Hikaru*; Torikai, Yuji*; Fujimura, Yuki; Michishio, Koji*
JJAP Conference Proceedings (Internet), 9, p.011106_1 - 011106_7, 2023/00
Positron annihilation lifetime (PAL) measurements by use of a positron source of 
Na were performed for polycrystalline ITER-grade tungsten samples exposed to low-energy deuterium plasma. The energy of deuterium plasma was low and then it was expected that it would affect just near-surface region. However, we obtained the longer mean positron annihilation lifetime in the tungsten samples exposed to the low-energy deuterium plasma than the virgin tungsten samples. Moreover, almost same longer values were obtained even on the other (no exposed) side of the samples, although the thickness of the samples were about 2 mm. Although, there has been no report of observation of defect formation by existence of hydrogen or deuterium in tungsten, the results indicated that deuterium existence in tungsten can be one of reasons of defects formation.
Ueno, Yasuhiro*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Katsuhiko*; Ito, Takashi; Iwasaki, Masahiko*; et al.
Hyperfine Interactions, 238(1), p.14_1 - 14_6, 2017/11
Times Cited Count：3 Percentile：86.59(Physics, Atomic, Molecular & Chemical)Strasser, P.*; Aoki, Masaharu*; Fukao, Yoshinori*; Higashi, Yoshitaka*; Higuchi, Takashi*; Iinuma, Hiromi*; Ikedo, Yutaka*; Ishida, Katsuhiko*; Ito, Takashi; Iwasaki, Masahiko*; et al.
Hyperfine Interactions, 237(1), p.124_1 - 124_9, 2016/12
Times Cited Count：7 Percentile：90.97(Physics, Atomic, Molecular & Chemical)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.
Agui, Akane; Mizumaki, Masaichiro*; Saito, Yuji; Matsushita, Tomohiro*; Nakatani, Takeshi*; Fukaya, Atsuko*; Torikai, Eiko*
Journal of Synchrotron Radiation, 8(2), p.907 - 909, 2001/03
Times Cited Count：9 Percentile：49.71(Instruments & Instrumentation)no abstracts in English
Isobe, Kanetsugu; Alimov, V.; Yamanishi, Toshihiko; Torikai, Yuji*; Hatano, Yuji*
no journal, ,
no abstracts in English
Isobe, Kanetsugu; Alimov, V.; Yamanishi, Toshihiko; Torikai, Yuji*
no journal, ,
To understand these plasma surface interactions, tritium distribution of tungsten exposed with low energy (38 eV), high flux D plasma was examined with BIXS. D plasma exposures were carried out at around 495 and 550 K of specimen. After that, specimen was exposed with gaseous tritium diluted with deuterium at 473 K in 5 hours. Amount of tritium in surface layer was measured with BIXS and tritium distribution of surface. The amount of tritium in surface layer was different of each exposure condition and tungsten exposed at 495 K shows highest amount of tritium. This result quite agrees with D inventory examination with thermal desorption spectrometer.
Watanabe, Kazuhito; Nakamura, Makoto; Someya, Yoji; Masui, Akihiro; Katayama, Kazunari*; Hayashi, Takumi; Yanagihara, Satoshi*; Konishi, Satoshi*; Yokomine, Takehiko*; Torikai, Yuji*; et al.
no journal, ,
In the DEMO design, the blanket primary cooling system involves high temperature pressurized water (~300
C). This means the temperature of blanket structural material is higher than that of ITER. This increases tritium permeation ratio from the fusion plasma and blanket breeder to the primary cooling water. Therefore, we need to consider installation of a water detritiation system. In this study, we estimate the demand of water detritiation system from the view point of the amount of tritium permeated to primary cooling water that assumed conservatively. We also organize the issues for management of tritiated water from the other point of view based on the characteristic of the fusion DEMO reactor. The result shows that the existing facilities can be adopted to the DEMO if we can control the tritium ratio of primary cooling water as same as that of CANDU reactor.
Nakamura, Makoto; Tobita, Kenji; Tanigawa, Hisashi; Someya, Yoji; Masui, Akihiro; Watanabe, Kazuhito; Konishi, Satoshi*; Torikai, Yuji*
no journal, ,
Tritium is major radioactive material in a fusion reactor. Evaluation of the dose due to the tritium is essential to understand environmental consequences of incidental or accidental conditions postulated in the fusion reactor. A purpose of this study is to identify issues to apply UFOTRI, a code of tritium dose analysis being used for the ITER safety assessment, the Japanese environmental conditions. Extensive scans of UFOTRI calculation runs were performed in various meteorological and release conditions. The scans show that the contribution of the secondary tritium release is more significant in the cases of lower release height, lesser stable atmosphere or more distant conditions. The analysis, thus, suggests that it is important to take into account the contribution of the secondarily released tritium in evaluating the early dose to the public due to the tritium release.
Kinase, Sakae; Torikai, Yuji*; Mikami, Satoshi; Ando, Masaki; Saito, Kimiaki
no journal, ,
no abstracts in English
