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Yamauchi, Kunihito; Okano, Jun; Shimada, Katsuhiro; Omori, Yoshikazu; Terakado, Tsunehisa; Matsukawa, Makoto; Koide, Yoshihiko; Kobayashi, Kazuhiro; Ikeda, Yoshitaka; Fukumoto, Masahiro; et al.
JAEA-Technology 2015-053, 36 Pages, 2016/03
JAEA-Technology-2015-053.pdf:8.33MB
The superconducting Satellite Tokamak machine "JT-60SA" under construction in Naka Fusion Institute is an international collaborative project between Japan (JA) and Europe (EU). The contributions for this project are based on the supply of components, and thus European manufacturer shall conduct the installation, commissioning and tests on Naka site. This means that Japan Atomic Energy Agency (JAEA) had a quite difficult issue to manage the works by European workers and their safety although there is no direct contract. This report describes the approaches for the work and safety managements, which were agreed with EU after the tough negotiation, and then the completed on-site works for Quench Protection Circuits (QPC) as the first experience for EU in JT-60SA project. With the help of these approaches by JAEA, the EU works for QPC were successfully completed with no accident, and a great achievement was made for both EU and JA.
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.
Ikeda, Yoshitaka; Okano, Fuminori; Sakasai, Akira; Hanada, Masaya; Akino, Noboru; Ichige, Hisashi; Kaminaga, Atsushi; Kiyono, Kimihiro; Kubo, Hirotaka; Kobayashi, Kazuhiro; et al.
Nihon Genshiryoku Gakkai Wabun Rombunshi, 13(4), p.167 - 178, 2014/12
The JT-60U torus was disassembled so as to newly install the superconducting tokamak JT-60SA torus. The JT-60U used the deuterium for 18 years, so the disassembly project of the JT-60U was the first disassembly experience of a fusion device with radioactivation in Japan. All disassembly components were stored with recording the data such as dose rate, weight and kind of material, so as to apply the clearance level regulation in future. The lessons learned from the disassembly project indicated that the cutting technologies and storage management of disassembly components were the key factors to conduct the disassembly project in an efficient way. After completing the disassembly project, efforts have been made to analyze the data for characterizing disassembly activities, so as to contribute the estimation of manpower needs and the radioactivation of the disassembly components on other fusion devices.
Ikeda, Yoshitaka; Okano, Fuminori; Hanada, Masaya; Sakasai, Akira; Kubo, Hirotaka; Akino, Noboru; Chiba, Shinichi; Ichige, Hisashi; Kaminaga, Atsushi; Kiyono, Kimihiro; et al.
Fusion Engineering and Design, 89(9-10), p.2018 - 2023, 2014/10
Times Cited Count：2 Percentile：16.44(Nuclear Science & Technology)Disassembly of the JT-60U torus was started in 2009 after 18-years D
operations, and was completed in October 2012. The JT-60U torus was featured by the complicated and welded structure against the strong electromagnetic force, and by the radioactivation due to D-D reactions. Since this work is the first experience of disassembling a large radioactive fusion device in Japan, careful disassembly activities have been made. About 13,000 components cut into pieces with measuring the dose rates were removed from the torus hall and stored safely in storage facilities by using a total wokers of 41,000 person-days during 3 years. The total weight of the disassembly components reached up to 5,400 tons. Most of the disassembly components will be treated as non-radioactive ones after the clearance verification under the Japanese regulation in future. The assembly of JT-60SA has started in January 2013 after this disassembly of JT-60U torus.
Hayashi, Takumi; Isobe, Kanetsugu; Nakamura, Hirofumi; Kobayashi, Kazuhiro; Oya, Yasuhisa*; Okuno, Kenji*; Oyaizu, Makoto; Edao, Yuki; Yamanishi, Toshihiko
Fusion Engineering and Design, 89(7-8), p.1520 - 1523, 2014/10
Times Cited Count：1 Percentile：8.88(Nuclear Science & Technology)Tritium confinement is the most important safety issue in the fusion reactor. Tritium behavior on the water metal boundary is very important to design tritium plant with breading blanket system using cooling water. A series of tritium permeation experiment into pressurized water or water vapor jacket with He or Ar have been performed through pure iron piping with/without 7 micro-meter gold plating, which contained about 1 kPa of pure tritium gas at 423 K, with monitoring the chemical forms of tritium. Also, deuterium permeation experiments from heavy water vessel through various metal piping, such as pure iron (Fe), nickel (Ni), stainless steel (SS304), and pure iron with 10 micro-meter gold plating, were performed at 573 K and at 15 MPa. Recently, using the above heavy water system, we have succeeded to detect simultaneous hydrogen isotopes transfer from and to the metal surface by introducing H gas to the metal piping after stabilized deuterium permeation was detected.
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：73.47(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, Takumi; Nakamura, Hirofumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Oyaizu, Makoto; Yamanishi, Toshihiko; Oya, Yasuhisa*; Okuno, Kenji*
Fusion Engineering and Design, 87(7-8), p.1333 - 1337, 2012/08
Times Cited Count：8 Percentile：52.49(Nuclear Science & Technology)In order to investigate the behavior of hydrogen isotope on the water-metal boundary, deuterium permeation experiments from heavy water vessel through various metal piping, such as pure iron (Fe), nickel (Ni), stainless steel (SS304), and pure iron with 10 
m gold plating, were performed at 573 K and at 15 MPa. During the experiment, surfaces of metal piping except gold plating one were oxidized at the heavy water boundary and then deuterium would generate by the oxidation reactions. This deuterium could be detected by mass spectrometer, which monitored the inside gases of the piping under vacuum. The result showed clearly that the deuterium permeated through Fe, Ni, and SS304 piping was detected as deuterium gas (D) under vacuum, though that through gold plating one could not be detected effectively. The D permeation rate through Fe, Ni, and SS304 piping reached some stabilized value. This paper summarizes the above experimental results and discusses the mechanism of deuterium behavior on the water metal boundary.
Fukada, Satoshi*; Edao, Yuki*; Sato, Koichi*; Takeishi, Toshiharu*; Katayama, Kazunari*; Kobayashi, Kazuhiro; Hayashi, Takumi; Yamanishi, Toshihiko; Hatano, Yuji*; Taguchi, Akira*; et al.
Fusion Engineering and Design, 87(1), p.54 - 60, 2012/01
Times Cited Count：4 Percentile：31.96(Nuclear Science & Technology)An experimental study on tritium (T) transfer in porous concrete for the tertiary T safety containment is performed to investigate (1) how fast HTO penetrates through concrete walls, (2) how well concrete walls contaminated with water-soluble T are decontaminated by a solution-in-water technique, and (3) how well hydrophobic paint coating works as a protecting film against HTO migrating through concrete walls. The epoxy paint coating can work as a HTO diffusion barrier and the PRF value is around 1/10. The silicon paint coating cannot work as the anti-T permeation barrier, because water deteriorates contact between the paint and cement or mortar.
Kobayashi, Kazuhiro; Nakamura, Hirofumi; Hayashi, Takumi; Yamanishi, Toshihiko
Fusion Science and Technology, 60(4), p.1335 - 1338, 2011/11
Times Cited Count：1 Percentile：10.73(Nuclear Science & Technology)In a fusion reactor of high safety and acceptability, safety 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 and the organic materials. Transport properties of tritiated water vapor (HTO) in the epoxy paint has been evaluated by the HTO exposure and removal behavior from the epoxy paint in order to obtain the data base of tritium behavior in the confinement facilities such as the hot cell or the tritium plant building of ITER.
Iwai, Yasunori; Kobayashi, Kazuhiro; Yamanishi, Toshihiko
Fusion Science and Technology, 60(3), p.1025 - 1028, 2011/10
Times Cited Count：2 Percentile：18.29(Nuclear Science & Technology)We have tested a number of polymeric materials used as gasket, insulator, glove and casing panel in the solid-polymer-electrolyte (SPE) tritiated water electrolyzer to evaluate the contamination by tritiated water and the change in contamination by irradiation. HTO contamination on polymeric materials both being exposed to 740-1110 Bq/cm
of HTO vapor with a 1kPa of HO pressure and being immersed in 70000 Bq/cm of HTO water was considered in the test. The exposed time affected negligibly the total amount of leached HTO from the rubber samples exposed to HTO vapor. The immersed time in contrast affected strongly the total amount of leached HTO from the rubber samples. The total amount of leached HTO from radiation-crosslinkable butyl rubber and radiation-degradable perfluoro Karlez rubber immersed in HTO was considerably increased as the integrated dose was increased. However, we found that the total amount of leached HTO from the irradiated rubber can maintain the similar amount from unirradiated by setting the hydrogen/fluoride ratio of the polymeric component to the suitable number.
Kobayashi, Kazuhiro; Hayashi, Takumi; Yamanishi, Toshihiko
Fusion Science and Technology, 60(3), p.1041 - 1044, 2011/10
Times Cited Count：2 Percentile：18.29(Nuclear Science & Technology)In a fusion reactor of high safety and acceptability, safety 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 and the organic materials. Since the organic and the concrete materials will be contaminated by tritium compared with the metal materials such as SS, it is very important to study the tritium behavior on the materials from viewpoint of protection the excess exposure to workers. Therefore, in order to understand for tritium behavior on the concrete materials, the sorption and desorption experiment was carried out as a function of the exposure time, temperature and tritiated water concentration. From the results, the behavior of tritium sorption and desorption in the concrete materials will be discussed.
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：18.29(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.
Iwai, Yasunori; Kobayashi, Kazuhiro; Yamanishi, Toshihiko
Journal of Nuclear Materials, 417(1-3), p.1187 - 1190, 2011/10
Times Cited Count：2 Percentile：18.29(Materials Science, Multidisciplinary)The concept of tritium containment and confinement is the root of fusion safety. Hence, HTO contamination on concretes and epoxy paint should be reduced as low as possible. Several kinds of hydrophobic coatings, a commercial silicic paint, a commercial acrylic paint, a commercial fluoric paint, methoxytrimethylsilane paint or metallic stick-sheets, were tested on concrete and epoxy samples. These samples were exposed to 740-1110Bq/cm of HTO vapor at room temperature for a given period from 1 to 60 weeks. Static leaching tests were carried out for every HTO absorbed sample in distilled water, and the amount of leached HTO was evaluated. The hydrophobic barriers were effective to reduce HTO penetration into concrete. After exposure to HTO for 1 week, the HTO amount penetrated into concrete was reduced to 54.2% of non-paint sample for methoxytrimethylsilane paint, 56.0% for a commercial fluoric paint, 66.8% for a commercial silicic paint, respectively. Effectiveness of these hydrophobic barriers became less as the samples were exposed to HTO for a longer period.
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：18.29(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.
Hayashi, Takumi; Nakamura, Hirofumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Oyaizu, Makoto; Oya, Yasuhisa*; Okuno, Kenji*; Yamanishi, Toshihiko
Fusion Science and Technology, 60(1), p.369 - 372, 2011/07
Times Cited Count：2 Percentile：18.29(Nuclear Science & Technology)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：11 Percentile：59.22(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.
Nakamura, Hirofumi; Kobayashi, Kazuhiro; Yokoyama, Sumi*; Saito, Shigeru; Yamanishi, Toshihiko; Kikuchi, Kenji*
Journal of Plasma and Fusion Research SERIES, Vol.9, p.326 - 331, 2010/08
Based on results of tritium measurement in the SS316 specimens irradiated up to 5.9 dpa in the SINQ target (580 MeV proton) using a thermal desorption (TDS) method, trap site density and trap energy in the materials induced by the high-energy proton irradiation have been evaluated by means of the numerical tritium transport analysis. The results indicate that almost residual tritium in the SS316 specimen exists in the trap site, whose trap density is maximum 238 appm (5.9 dpa) and trap energy is 1.4 eV, and that tritium release by the TDS is mainly attributed to the disappearance of the trap sites by the specimen heating, whose activation energy is about 0.7 eV. The trap site density seems to be almost proportional to the irradiation dose (dpa). Additionally, irradiation conditions such as the dose or irradiation temperature do not affect on the trapping mechanism.
Hayashi, Takumi; Nakamura, Hirofumi; Isobe, Kanetsugu; Kobayashi, Kazuhiro; Oyaizu, Makoto; Yamanishi, Toshihiko; Ishikawa, Hirotada*; Oya, Yasuhisa*; Okuno, Kenji*
Fusion Science and Technology, 56(2), p.836 - 840, 2009/08
Times Cited Count：11 Percentile：59.85(Nuclear Science & Technology)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：10 Percentile：37.12(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.
