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Journal Articles

Evaluation of tritium confinement performance of alumina and zirconium for tritium production in a high-temperature gas-cooled reactor for fusion reactors

Katayama, Kazunari*; Ushida, Hiroki*; Matsuura, Hideaki*; Fukada, Satoshi*; Goto, Minoru; Nakagawa, Shigeaki

Fusion Science and Technology, 68(3), p.662 - 668, 2015/10

 Times Cited Count:7 Percentile:33.59(Nuclear Science & Technology)

Tritium production utilizing nuclear reactions by neutron and lithium in a high-temperature gas-cooled reactor is attractive for development of a fusion reactor. From viewpoints of tritium safety and production efficiency, tritium confinement technique is an important issue. It is known that alumina has high resistance for gas permeation. In this study, hydrogen permeation experiments in commercial alumina tubes were conducted and hydrogen permeability, diffusivity and solubility was evaluated. By using obtained data, tritium permeation behavior from an Al$$_{2}$$O$$_{3}$$-coated Li-compound particle was simulated. Additionally, by using literature data for hydrogen behavior in zirconium, an effect of Zr incorporation into an Al$$_{2}$$O$$_{3}$$ coating on tritium permeation was discussed. It was indicated that the majority of produced tritium was released through the Al$$_{2}$$O$$_{3}$$ coating above 500$$^{circ}$$C. However, it is expected that total tritium leak is suppressed to below 0.67% of total tritium produced at 500$$^{circ}$$C by incorporating Zr fine particles into the inside of Al$$_{2}$$O$$_{3}$$ coating.

Journal Articles

Measurement of tritium penetration through concrete material covered by various paints coating

Edao, Yuki; Kawamura, Yoshinori; Kurata, Rie; Fukada, Satoshi*; Takeishi, Toshiharu*; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 67(2), p.320 - 323, 2015/03

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

The present study aims at obtaining fundamental knowledge for tritium transfer behavior and interaction between tritium and paint coated on concrete walls. The amounts of tritium penetration and release in cement paste with epoxy and urethane paint coatings were measured. The tritium penetration amounts were increased with the HTO exposure time. Time to achieve each saturate tritium value was more than 60 days for cement paste coated with epoxy paint and with urethane paint, while cement paste without paint took 2 days to achieve it. Tritium penetration rates were estimated by an analysis of diffusion model. Although their paint coatings were effective for reduction of tritium penetration through the cement paste exposed to HTO for a short period, the amount of tritium trapped in the paints became large for a long time. This work has been performed under the collaboration research between JAEA and Kyushu University.

Journal Articles

Correlation of rates of tritium migration through porous concrete

Fukada, Satoshi*; Katayama, Kazunari*; Takeishi, Toshiharu*; Edao, Yuki; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko

Fusion Science and Technology, 67(2), p.99 - 102, 2015/03

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Status of development of Lithium Target Facility in IFMIF/EVEDA project

Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Hirakawa, Yasushi; Furukawa, Tomohiro; Hoashi, Eiji*; Fukada, Satoshi*; Suzuki, Akihiro*; Yagi, Juro*; Tsuji, Yoshiyuki*; et al.

Proceedings of Plasma Conference 2014 (PLASMA 2014) (CD-ROM), 2 Pages, 2014/11

In the IFMIF/EVEDA (International Fusion Materials Irradiation Facility/ Engineering Validation and Engineering Design Activity), the validation tests of the EVEDA lithium test loop with the world's highest flow rate of 3000 L/min was succeeded in generating a 100 mm-wide and 25 mm-thick free-surface lithium flow steadily under the IFMIF operation condition of a high-speed of 15 m/s at 250$$^{circ}$$C in a vacuum of 10 $$^{-3}$$ Pa. Some excellent results of the recent engineering validations including lithium purification, lithium safety, and remote handling technique were obtained, and the engineering design of lithium facility was also evaluated. These results will advance greatly the development of an accelerator-based neutron source to simulate the fusion reactor materials irradiation environment as an important key technology for the development of fusion reactor materials.

Journal Articles

Penetration of tritiated water vapor through hydrophobic paints for concrete materials

Edao, Yuki; Kawamura, Yoshinori; Yamanishi, Toshihiko; Fukada, Satoshi*

Fusion Engineering and Design, 89(9-10), p.2062 - 2065, 2014/10

 Times Cited Count:1 Percentile:88.23(Nuclear Science & Technology)

Tritium transfer behavior through hydrophobic paints, epoxy resin and acrylic-silicon resin, was investigated experimentally. The authors measured the amount of tritium permeated through the paint membranes which exposed in HTO atmosphere of 2$$sim$$100 Bq/cm$$^{3}$$. The most of tritium permeated through the paints in the form of HTO at room temperature. Tritium permeation through the acrylic-silicon paint was explained a linear sorption/release model and that through the epoxy paint was suggested to be controlled by a one-dimensional diffusion model. While effective diffusivity was 1.0$$times$$10$$^{-13}$$$$sim$$1.8$$times$$10$$^{-13}$$ m$$^{2}$$/s at 21$$^{circ}$$C$$sim$$26$$^{circ}$$C for epoxy membrane, the diffusivity was found to be hundreds times larger than that for cement-paste coated with epoxy paint. Hence, tritium diffusivity through interface between cement-paste and the epoxy paint was considered to be most effective in the overall tritium transfer process. Tritium transfer behavior in the interface is important to explain the mechanism of tritium transfer behavior in concrete walls.

Journal Articles

Engineering validation and engineering design of lithium target facility in IFMIF/EVEDA project

Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Furukawa, Tomohiro; Hirakawa, Yasushi; Watanabe, Kazuyoshi; Ida, Mizuho*; Ito, Yuzuru; Niitsuma, Shigeto; Edao, Yuki; et al.

Fusion Science and Technology, 66(1), p.46 - 56, 2014/07

 Times Cited Count:4 Percentile:59.64(Nuclear Science & Technology)

Journal Articles

Development of lithium target system in engineering validation and engineering design activity of the International Fusion Materials Irradiation Facility (IFMIF/EVEDA)

Wakai, Eiichi; Kondo, Hiroo; Sugimoto, Masayoshi; Fukada, Satoshi*; Yagi, Juro*; Ida, Mizuho; Kanemura, Takuji; Furukawa, Tomohiro; Hirakawa, Yasushi; Watanabe, Kazuyoshi; et al.

Purazuma, Kaku Yugo Gakkai-Shi, 88(12), p.691 - 705, 2012/12

no abstracts in English

Journal Articles

Transfer of tritium in concrete coated with hydrophobic paints

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:61.15(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.

Journal Articles

IFMIF/EVEDA lithium test loop; Design and fabrication technology of target assembly as a key component

Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Nakamura, Kazuyuki; Ida, Mizuho; Watanabe, Kazuyoshi; Kanemura, Takuji; Wakai, Eiichi; Horiike, Hiroshi*; Yamaoka, Nobuo*; et al.

Nuclear Fusion, 51(12), p.123008_1 - 123008_12, 2011/12

 Times Cited Count:35 Percentile:13.81(Physics, Fluids & Plasmas)

The Engineering Validation and Engineering Design Activity (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) is proceeded as one of the ITER Broader Approach (BA) activities. The EVEDA Li test loop (ELTL) is aimed at validating stability of the Li target and feasibility of a Li purification system as the key issues. In this paper, the design of the ELTL especially of a target assembly in which the Li target is produced by the contraction nozzle is presented.

Journal Articles

Present status of Japanese tasks for lithium target facility under IFMIF/EVEDA

Nakamura, Kazuyuki; Furukawa, Tomohiro; Hirakawa, Yasushi; Kanemura, Takuji; Kondo, Hiroo; Ida, Mizuho; Niitsuma, Shigeto; Otaka, Masahiko; Watanabe, Kazuyoshi; Horiike, Hiroshi*; et al.

Fusion Engineering and Design, 86(9-11), p.2491 - 2494, 2011/10

 Times Cited Count:9 Percentile:35.08(Nuclear Science & Technology)

In IFMIF/EVEDA, tasks for lithium target system are shared to 5 validation tasks (LF1-5) and a design task (LF6). The purpose of LF1 task is to construct and operate the EVEDA lithium test loop, and JAEA has a main responsibility to the performance of the Li test loop. LF2 is a task for the diagnostics of the Li test loop and IFMIF design. Basic research for the diagnostics equipment has been completed, and the construction for the Li test loop will be finished before March in 2011. LF4 is a task for the purification systems with nitrogen and hydrogen. Basic research for the purification equipment has been completed, and the construction of the nitrogen system for the Li test loop will be finished before March in 2011. LF5 is a task for the remote handling system with the target assembly. JAEA has an idea to use the laser beam for cutting and welding of the lip part of the flanges. LF6 is a task for the design of the IFMIF based on the validation experiments of LF1-5.

Journal Articles

Design of purification loop and traps for the IFMIF/EVEDA Li test loop; Design of cold trap

Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Iuchi, Hiroshi; Ida, Mizuho; Yagi, Juro*; Suzuki, Akihiro*; Fukada, Satoshi*; Matsushita, Izuru*; Nakamura, Kazuyuki

Fusion Engineering and Design, 86(9-11), p.2437 - 2441, 2011/10

 Times Cited Count:17 Percentile:16.68(Nuclear Science & Technology)

Engineering Validation and Engineering Design Activities (EVEDA) for The International Fusion Materials Irradiation Facility (IFMIF) were started from July 2007 under an international agreement called ITER Broader Approach. As a major Japanese activity, EVEDA Li test loop (ELTL) to simulate hydraulic and impurity conditions of IFMIF has already designed and is under construction, in which feasibility of hydraulic stability of the liquid Li target, the purification systems of hot traps are major key issues to be validated in this loop. This paper focuses on the purification systems of the ELTL. Design of a cold trap and hot traps are discussed in this paper.

Journal Articles

Target system of IFMIF-EVEDA in Japanese activities

Ida, Mizuho; Fukada, Satoshi*; Furukawa, Tomohiro; Hirakawa, Yasushi; Horiike, Hiroshi*; Kanemura, Takuji*; Kondo, Hiroo; Miyashita, Makoto; Nakamura, Hiroo; Sugiura, Hirokazu*; et al.

Journal of Nuclear Materials, 417(1-3), p.1294 - 1298, 2011/10

 Times Cited Count:3 Percentile:69.54(Materials Science, Multidisciplinary)

Engineering Validation and Engineering Design Activities (EVEDA) of the International Fusion Materials Irradiation Facility (IFMIF) was started. As a Japanese activity for the target system, EVEDA Lithium Test Loop simulating hydraulic and impurity conditions of IFMIF is under design and preparation for fabrication. Feasibility of thermo-mechanical structure of the target assembly and the replaceable back-plate made of F82H (a RAFM) and 316L (a stainless steel) is a key issue. Toward final validation on the EVEDA loop, diagnostics applicable to a high-speed free-surface Li flow and hot traps to control nitrogen and hydrogen in Li are under tests. For remote handling of target assemblies and the replaceable back-plates activated up to 50 dpa/y, lip weld on 316L-316L by laser and dissimilar weld on F82H-316L are under investigation. As engineering design of the IFMIF target system, water experiments and hydraulic/thermo-mechanical analyses of the back-plate are going.

Journal Articles

Design and construction of IFMIF/EVEDA lithium test loop

Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Ida, Mizuho; Matsushita, Izuru*; Horiike, Hiroshi*; Kanemura, Takuji; Sugiura, Hirokazu*; Yagi, Juro*; Suzuki, Akihiro*; et al.

Journal of Engineering for Gas Turbines and Power, 133(5), p.052910_1 - 052910_6, 2010/12

 Times Cited Count:7 Percentile:54.74(Engineering, Mechanical)

As a major Japanese activity for the IFMIF/EVEDA, EVEDA Li Test Loop (ELTL) to simulate hydraulic and impurity conditions of IFMIF is under design and preparation for fabrication. Feasibility of hydraulic stability of the liquid Li target and the purification systems of hot traps are major key issues to be validated. This paper presents the current status of the design and construction of the EVEDA Li Test Loop. Detail designs of the loop components such as the target assembly, tanks, an electro-magnetic pump and flow meter and a cold trap for purification system are described in addition to the flow diagnostics system and the hot traps.

Journal Articles

Engineering design and construction of IFMIF/EVEDA lithium test loop; Design and fabrication of integrated target assembly

Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Nakamura, Hiroo*; Ida, Mizuho; Watanabe, Kazuyoshi; Miyashita, Makoto*; Horiike, Hiroshi*; Yamaoka, Nobuo*; Kanemura, Takuji; et al.

Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 8 Pages, 2010/10

The Engineering Validation and Engineering Design Activity (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) is proceeded as one of the ITER Broader Approach (BA) activities. The EVEDA Li test loop (ELTL) is aimed at validating stability of the Li target and feasibility of a Li purification system as the key issues. In this paper, the design of the ELTL especially of a target assembly in which the Li target is produced by the contraction nozzle is presented.

Journal Articles

Current status of design and construction of IFMIF/EVEDA lithium test loop

Kondo, Hiroo; Furukawa, Tomohiro; Hirakawa, Yasushi; Matsushita, Izuru*; Ida, Mizuho; Horiike, Hiroshi*; Kanemura, Takuji; Sugiura, Hirokazu*; Yagi, Juro*; Suzuki, Akihiro*; et al.

Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05

IFMIF is a neutron source aimed at producing an intense high energy neutron flux for testing candidate fusion reactor materials. Under Broader Approach activities, Engineering Validation and Engineering Design Activities (EVEDA) of IFMIF started on July 2007. Regarding to the lithium (Li) target facility, design and construction of EVEDA Li Test Loop is a major activity and is in progress. The detail design was started at the early 2009. Fabrication of the loop was started at middle of 2009, and completion is planned at the end of Feb. 2011.

Journal Articles

Tritium removal by Y hot trap for purification of IFMIF Li target

Edao, Yuki*; Fukada, Satoshi*; Yamaguchi, Sho*; Wu, Y.*; Nakamura, Hiroo

Fusion Engineering and Design, 85(1), p.53 - 57, 2010/01

 Times Cited Count:15 Percentile:23.46(Nuclear Science & Technology)

Journal Articles

Status of engineering design of liquid lithium target in IFMIF-EVEDA

Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; Giusti, D.*; Groeschel, F.*; Horiike, Hiroshi*; et al.

Fusion Engineering and Design, 84(2-6), p.252 - 258, 2009/06

 Times Cited Count:25 Percentile:12.45(Nuclear Science & Technology)

Journal Articles

Latest design of liquid lithium target in IFMIF

Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Cevolani, S.*; Chida, Teruo*; Ciotti, M.*; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; et al.

Fusion Engineering and Design, 83(7-9), p.1007 - 1014, 2008/12

 Times Cited Count:14 Percentile:27.98(Nuclear Science & Technology)

This paper describes the latest design of liquid lithium target system in IFMIF. Design requirement of the Li target is to provide a stable Li jet with a speed of 20 m/s to handle an averaged heat flux of 1 GW/m$$^{2}$$. A double reducer nozzle and a concaved flow are applied to the target design. On Li purification, a cold trap and two kinds of hot trap are applied to control impurities below permissible levels. Nitrogen concentration shall be controlled below 10 wppm by one of the hot trap. Tritium concentration shall be controlled below 1 wppm by an yttrium hot trap. To maintain reliable continuous operation, various diagnostics are attached to the target assembly. Among the target assembly, a back-plate made of RAFM is located in the most severe region of neutron irradiation (50 dpa/y). Therefore, two design options of replaceable back wall and their remote handling systems are under investigation.

Journal Articles

Research and development on Water-Cooled Solid Breeder Test Blanket Module in JAEA

Enoeda, Mikio; Tanigawa, Hisashi; Tsuru, Daigo; Hirose, Takanori; Ezato, Koichiro; Yokoyama, Kenji; Dairaku, Masayuki; Seki, Yohji; Suzuki, Satoshi; Mori, Kensuke*; et al.

Proceedings of 16th Pacific Basin Nuclear Conference (PBNC-16) (CD-ROM), 6 Pages, 2008/10

Journal Articles

Concentration profiles of tritium penetrated into concrete

Takata, Hiroki*; Furuichi, Kazuya*; Nishikawa, Masabumi*; Fukada, Satoshi*; Katayama, Kazunari*; Takeishi, Toshiharu*; Kobayashi, Kazuhiro; Hayashi, Takumi; Namba, Haruyuki*

Fusion Science and Technology, 54(1), p.223 - 226, 2008/07

 Times Cited Count:8 Percentile:46.21(Nuclear Science & Technology)

Concentration profiles of tritium penetrated into cement paste, mortar and concrete were measured by using samples with a shape of column. Tritium penetrated until a location of about 5 cm from the exposed surface after 6 months' exposure. The amount of tritium penetrated into mortar and concrete were less than 70% and half that into cement paste.

39 (Records 1-20 displayed on this page)