Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Edao, Yuki; Sato, Katsumi; Iwai, Yasunori; Hayashi, Takumi
Journal of Nuclear Science and Technology, 53(11), p.1831 - 1838, 2016/11
Times Cited Count：6 Percentile：61.4(Nuclear Science & Technology)Isobe, Kanetsugu; Kawamura, Yoshinori; Iwai, Yasunori; Oyaizu, Makoto; Nakamura, Hirofumi; Suzuki, Takumi; Yamada, Masayuki; Edao, Yuki; Kurata, Rie; Hayashi, Takumi; et al.
Fusion Engineering and Design, 98-99, p.1792 - 1795, 2015/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Activities on Broader Approach (BA) were started in 2007 on the basis of the Agreement between the Government of Japan and the EURATOM. The period of BA activities consist of Phase1 and Phase2 dividing into Phase 2-1 (2010-2011), Phase 2-2 (2012-2013) and Phase 2-3 (2014-2016). Tritium technology was chosen as one of important R&D issues to develop DEMO plant. R&D activities of tritium technology on BA consist of four tasks. Task-1 is to prepare and maintain the tritium handling facility in Rokkasho BA site in Japan. Task 2, 3 and 4 are main R&D activities for tritium and these are focused on: Task-2) Development of tritium accountancy technology, Task-3) Development of basic tritium safety research, Task-4) Tritium durability test. R&D activities of tritium technology in Phase 2-2 were underway successfully and closed in 2013.
Hayashi, Takumi; Nakamura, Hirofumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Yamada, Masayuki; Suzuki, Takumi; Kurata, Rie; Oyaizu, Makoto; Edao, Yuki; et al.
Fusion Science and Technology, 67(2), p.365 - 370, 2015/03
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Kawamura, Yoshinori; Edao, Yuki; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko
Fusion Engineering and Design, 89(7-8), p.1539 - 1543, 2014/10
Times Cited Count：6 Percentile：50.1(Nuclear Science & Technology)Tritium recovery system using adsorption or catalytic isotope exchange has already been proposed for a solid breeding blanket system of a nuclear fusion reactor. Synthetic zeolite is often used as an adsorbent or a substrate of chemical exchange catalyst. And, it is well known that its properties are changed easily by exchanging their cations. So, in this work, adsorption capacities of hydrogen isotope and water vapor on cation-exchanged mordenite with transition metal ion were investigated. Ag ion-exchanged mordenite (Ag-MOR) has indicated considerably large hydrogen adsorption capacity in lower pressure range at 77 K. And, adsorption capacity of water vapor did not so vary with exchaned cation in comparison with hydrogen adsorption. The discussion from the viewpoint of adsorption rate is still remaining, but more compact cryosorption column for tritium recovery system is possible to design if Ag-MOR is adopted.
Oda, Yasuhisa; Oshima, Katsumi; Nakamoto, Takashi*; Hashimoto, Yasunori*; Yamamoto, Tsuyoshi; Hayashi, Kazuo*; Ikeda, Yukiharu; Ikeda, Ryosuke; Kajiwara, Ken; Takahashi, Koji; et al.
Purazuma, Kaku Yugo Gakkai-Shi, 90(7), p.365 - 373, 2014/07
no abstracts in English
Yamanishi, Toshihiko; Nakamura, Hirofumi; Kawamura, Yoshinori; Iwai, Yasunori; Isobe, Kanetsugu; Oyaizu, Makoto; Yamada, Masayuki; Suzuki, Takumi; Hayashi, Takumi
Fusion Engineering and Design, 87(5-6), p.890 - 895, 2012/08
Times Cited Count：1 Percentile：11.68(Nuclear Science & Technology)In JAEA, the tritium processing and handling technologies have been studied at TPL. The main basic R&D activities in this field are: the tritium processing technology for the blanket recovery system; the tritium behavior in a confinement; and detritiation and decontamination. The R&D for tritium processing and handling technologies to a demonstration reactor (DEMO) are also planned to be carried out in the Broader Approach (BA) program in Japan by JAEA with Japanese universities. The ceramic electrolysis cell has been studied as a tritium processing method for the blanket system. The permeation behavior of tritium through pure iron into the gas containing water vapor has been studied. As for the behavior of high concentration tritium water, it was observed that the formation of the oxidized layer was prevented by the presence of tritium in water. Tritium durability tests were also carried out for the electrolysis cell of the chemical exchange column.
Kawamura, Yoshinori; Ochiai, Kentaro; Hoshino, Tsuyoshi; Kondo, Keitaro*; Iwai, Yasunori; Kobayashi, Kazuhiro; Nakamichi, Masaru; Konno, Chikara; Yamanishi, Toshihiko; Hayashi, Takumi; et al.
Fusion Engineering and Design, 87(7-8), p.1253 - 1257, 2012/08
Times Cited Count：15 Percentile：78.3(Nuclear Science & Technology)Tritium generation and recovery study on lithium ceramic packed bed was started by use of FNS in JAEA. Lithium titanate was selected as tritium breeding material. In this work, the effect of sweep gas species on tritium release behavior was investigated. In case of sweep by helium with 1% of hydrogen, tritium in water form was released sensitively corresponding to the irradiation. This is due to existence of the water vapor in the sweep gas. On the other hand, in case of sweep by dry helium, tritium in gaseous form was released first, and release of tritium in water form was delayed and was gradually increased.
Hayashi, Nobuhiko; Ide, Shunsuke; Suzuki, Takahiro; Itami, Kiyoshi; Kawano, Yasunori; Sasao, Mamiko*; Kusama, Yoshinori
Purazuma, Kaku Yugo Gakkai-Shi, 88(7), p.392 - 393, 2012/07
no abstracts in English
Yoshida, Maiko; Isayama, Akihiko; Urano, Hajime; Kawano, Yasunori; Hamamatsu, Kiyotaka; Hayashi, Nobuhiko
Purazuma, Kaku Yugo Gakkai-Shi, 87(12), p.849 - 852, 2011/12
Some ITPA Topical Group Meetings were held individually. We discussed international joint experiments, issues for ITER and plan for next year in each group.
Yamanishi, Toshihiko; Hayashi, Takumi; Iwai, Yasunori; Isobe, Kanetsugu; Hara, Masanori*; Sugiyama, Takahiko*; Okuno, Kenji*
Fusion Engineering and Design, 86(9-11), p.2152 - 2155, 2011/10
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)It is quite significant subject how to confine the tritium in a fusion reactor. Especially, it is strongly desired to get the data for tritiated water. This is because tritiated water is much hazardous than the hydrogen form of tritium. As for the behavior of high concentration tritium water, we could get a series of valuable data for the corrosion of the tritiated water against metal materials. In the case where a metal material is in water, an oxidized layer is formed at the surface of the metal. The oxidized layer functions as a passive layer for the corrosion. However, it has been observed that the formation of the oxidized layer was prevented by the presence of tritium in water (0.23 GBq/cc). The chemical exchange column has been applied in ITER as the tritium recovery system from tritiated water. A set of data for an advanced chemical exchange column has been obtained.
Yamanishi, Toshihiko; Yamada, Masayuki; Suzuki, Takumi; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Inomiya, Hiroshi; Hayashi, Takumi
Fusion Science and Technology, 60(3), p.1083 - 1087, 2011/10
Times Cited Count：2 Percentile：20.43(Nuclear Science & Technology)Tritium Process Laboratory (TPL) in Japan Atomic Energy Agency has been established as the only test facilities to handle over 1 gram of in Japan. From March 1988, TPL has been operated with tritium, and no tritium release accident has been observed. The average tritium concentration in a stream from a stack of the TPL to environment was 71 Bq/m
, and was 1/70 of the Japanese regulation value for HTO. The failure data have been analyzed for several main components of the safety systems such as pumps, valves, and monitors. The data on the tritium waste and accountancy has also been accumulated. As a study of the Grants-in-Aid for Scientific Research, these data are analysed and are reported.
Kobayashi, Kazuhiro; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko
Journal of Nuclear Materials, 417(1-3), p.1183 - 1186, 2011/10
Times Cited Count：2 Percentile：20.43(Materials Science, Multidisciplinary)In a fusion reactor of high safety and acceptability, safe confinement of tritium is one of key issues for the fusion reactor. Tritium should be well-controlled and not excessively released to environment and to prevent workers from excess exposure. Especially, the hot cell and tritium facility of ITER will be used various construction materials such as the concrete, the organic materials. As the results, the concrete materials were almost saturated with HTO vapor within about 1month except for cement paste and it was larger in the order of cement paste 
mortar concrete. Even if one month passes from the exposure beginning, the amount of sorbed tritium to cement paste did not reach saturation. The chemical form of desorbed tritium from the sample was almost HTO. In addition, the tritium behavior that adsorbs the surface of concrete materials will be discussed by using FT-IR.
Iwai, Yasunori; Sato, Katsumi; Hiroki, Akihiro; Tamada, Masao; Hayashi, Takumi; Yamanishi, Toshihiko
Fusion Engineering and Design, 85(7-9), p.1421 - 1425, 2010/12
Times Cited Count：2 Percentile：19.27(Nuclear Science & Technology)The deteriorations of polymeric materials for a SPE-type high-level tritiated water electrolyzer composed of the Water Detritiation System (WDS) against sulfonic acid environment and radiation environment were discussed. A long-term durability of VITON, AFLAS, denaturated polyphenylene ether, and Kapton polyimide immersed in a sulfonic acid was demonstrated. Negligible degradation in percent elongation at break of these polymeric materials was observed up to the immersing period of 2 years. The detectable radiation deterioration in ionic conductivity of Nafion N117CS ion exchange membrane irradiated with electron beams up to the integrated dose of 1500 kGy was measured. The ionic conductivity of Nafion N117CS ion exchange membrane irradiated at more than 1000 kGy was slightly deteriorated. As for the elastomers for its use as a seal, the radiation degradation in hardness of VITON, AFLAS was investigated. Negligible degradation in hardness of these rubbers was observed up to the integrated dose of 1500 kGy. The water uptake of rubbers was generally increased as the integrated dose was increased. However, irradiated VITON rubbers had constant water uptake up to the integrated dose of 1500 kGy.
Yamanishi, Toshihiko; Hayashi, Takumi; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Suzuki, Takumi; Yamada, Masayuki
Fusion Engineering and Design, 85(7-9), p.1002 - 1006, 2010/12
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The R&D for tritium technologies to a demonstration reactor (DEMO) is planned to be carried out in the Broader Approach (BA) program in Japan by JAEA with Japanese universities: (1) tritium analysis technology; (2) basic tritium safety research; and (3) tritium durability test. A multi-purpose RI facility is under construction at Rokkasho in Aomori to carry out the above R&D subjects. A preliminary safety study has been carried out for the amount of tritium released to the environment and for the radiation dose of workers. The main subjects of the R&D of tritium analysis are the technologies for real-time analysis for hydrogen isotopes, gas, liquid and solid. The materials of interest include F82H, SiC, ZrCo, solid and liquid advanced breeder and multipliers. In the tritium durability tests, organic materials and metals are studied for the radiation and the corrosion damage. A series of preliminary studies for the above subjects has been started.
Hayashi, Takumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Nakamura, Hirofumi; Yamanishi, Toshihiko; Perevezentsev, A.*
Fusion Engineering and Design, 85(7-9), p.1386 - 1390, 2010/12
Times Cited Count：6 Percentile：43.83(Nuclear Science & Technology)In order to establish effective ITER atmosphere detritiation system (DS), JAEA has investigated the performance and the durability at various incident/accident conditions, and supported to finalize the DS conceptual design through the ITER design review. The current DS at the safety important component has been discussed and mainly consists of catalytic reactors, wet scrubber column (SC) and blowers. The functional failure of the DS design with SC was evaluated using database of failure experiences of valves, controllers and components. Even in the tritium release into the biggest confinement sector of Tokamak gallery, it improved more than tow orders of magnitude comparing with that of original DS design using Molecular Sieve (MS) dryer beds in the 2001 design report. This improvement is achieved mainly by the minimization of valve operation like MS dryers and by the standardized module arrangement of DS with SC.
Osakabe, Masaki*; Shinohara, Koji; Toi, Kazuo*; Todo, Yasushi*; Hamamatsu, Kiyotaka; Murakami, Sadayoshi*; Yamamoto, Satoshi*; Idomura, Yasuhiro; Sakamoto, Yoshiteru; Tanaka, Kenji*; et al.
Purazuma, Kaku Yugo Gakkai-Shi, 85(12), p.839 - 842, 2009/12
no abstracts in English
Yamanishi, Toshihiko; Hayashi, Takumi; Iwai, Yasunori; Isobe, Kanetsugu; Hara, Masanori*; Sugiyama, Takahiko*; Okuno, Kenji*
Purazuma, Kaku Yugo Gakkai-Shi, 85(10), p.716 - 725, 2009/10
In a fusion reactor, tritium must be handled in a vacuum vessel, a fuel cycle, and other systems. It is quite significant subject how to confine the tritium in these systems. ITER is the first machine in the world where the tritium confinement would be demonstrated. It is essential to establish a series of database for tritium handling technology to analyze the data obtained at ITER. Especially, it is strongly desired to get the data for tritium water. This is because tritium water is much hazardous than the hydrogen form of tritium. For these reasons, our attention is focused on the study on the behavior of tritium water in the fuel cycle system and structural materials of the system. As for the behavior of high concentration tritium water, we could get the first series of valuable data for the corrosion of the tritium water against metal materials in the world. The behavior of tritium in a surface area of metal and organic compounds was also discussed. As for tritium recovery from tritium water, we could get the data on an advanced chemical exchange column. The chemical exchange column has been applied in ITER as the tritium recovery system from tritium water. Our data showed that the column performance could be remarkably improved. Tritium durability and catalyst tests have also been carried out for the chemical exchange column. Some other possible method for the tritium recovery has also been studied such as an advance adsorption method.
Kawamura, Yoshinori; Iwai, Yasunori; Hayashi, Takumi; Yamanishi, Toshihiko; Munakata, Kenzo*
Fusion Science and Technology, 56(1), p.168 - 172, 2009/07
Times Cited Count：4 Percentile：32.89(Nuclear Science & Technology)Synthesis zeolite is the candidate material of the separation column of the gas chromatograph for the hydrogen isotope analysis. Mordenite (MOR) is one of the zeolite, and has been reported that the MOR column can separate hydrogen isotope at about 200 K. So, the present authors have investigated the adsorption capacities of H
and D on MOR at various temperatures, and have predicted the adsorption isotherms of HD, HT, DT and T. In this work, the adsorption capacities of tritium on MOR at 77 K and 87 K were investigated, and they were compared with the predicted isotherms. The observed isotherm at 87 K agreed with the predicted isotherm fairly well. However, at 77 K, the adsorption capacity at low presser region was smaller than the isotherm of D.
Kawamura, Yoshinori; Isobe, Kanetsugu; Iwai, Yasunori; Kobayashi, Kazuhiro; Nakamura, Hirofumi; Hayashi, Takumi; Yamanishi, Toshihiko
Nuclear Fusion, 49(5), p.055019_1 - 055019_8, 2009/05
Times Cited Count：9 Percentile：36.96(Physics, Fluids & Plasmas)Tritium technologies have reached the level where they allow us to design the main fuel cycle of ITER. On the other hand, for the blanket tritium recovery system, a series of fundamental studies have still been carried out even though the system is essential to realize the fusion reactor from the viewpoint of the fuel production. In the case of a water cooling solid breeder blanket, the blanket tritium recovery system will be composed of three processes: tritium recovery from the helium sweep gas as hydrogen, that as water vapor and tritium recovery from the coolant water. For these processes, the present authors have proposed a set of advanced systems, and have proved that the proposed systems would be feasible for a DEMO reactor.
Yamanishi, Toshihiko; Hayashi, Takumi; Shu, Wataru; Kawamura, Yoshinori; Nakamura, Hirofumi; Iwai, Yasunori; Kobayashi, Kazuhiro; Isobe, Kanetsugu; Arita, Tadaaki; Hoshi, Shuichi; et al.
Fusion Engineering and Design, 83(10-12), p.1359 - 1363, 2008/12
Times Cited Count：4 Percentile：31.93(Nuclear Science & Technology)At TPL (Tritium Process Laboratory) of JAEA, ITER relevant tritium technologies have been studied. The design studies of Air Detritiation System have been carried out in JAEA as a contribution of Japan to ITER. For the tritium processing technologies, our efforts have been focused on the research of the tritium recovery system of ITER test blanket system. A ceramic proton conductor has been studied as an advanced blanket system. A series of fundamental studies on tritium safety technologies not only for ITER but also for fusion DEMO plants has also been carried out at TPL of JAEA. The main research activities in this field are the tritium behavior in a confinement and its barrier materials; monitoring; accountancy; detritiation and decontamination etc. In this paper, the results of above recent activities at TPL of JAEA are summarized from viewpoint of ITER relevant and future fusion DEMO reactors.
