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Sun, Haomin; Kunugi, Tomoaki*; Yokomine, Takehiko*; Shen, X.*; Hibiki, Takashi*
Experimental Thermal and Fluid Science, 154, p.111171_1 - 111171_24, 2024/05
Times Cited Count：0 Percentile：0.00(Thermodynamics)Sun, Haomin; Kunugi, Tomoaki*; Yokomine, Takehiko*; Shen, X.*; Hibiki, Takashi*
International Journal of Heat and Mass Transfer, 211, p.124214_1 - 124214_17, 2023/09
Times Cited Count：2 Percentile：46.28(Thermodynamics)Okazaki, Soichiro*; Ezure, Toshiki; Ohshima, Hiroyuki; Kawara, Zensaku*; Yokomine, Takehiko*; Kunugi, Tomoaki*
Proceedings of 10th Pacific Symposium on Flow Visualization and Image Processing (PSFVIP-10), 8 Pages, 2015/06
A visualization study is performed under suction vortex geometry in water. In the experiment, the shear-sensitive liquid crystal coating (SSLCC) is applied to grasp the distribution of wall shear stress under the suction votex flow. As the result, it was found that the peak value of wall shear stress is appeared at the center and edge of the projected area of suction pipe. The non-dimensional profile of wall shear stress obtained by suction vortex flow experiment agrees well with that of numerical simulation.
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
C in a vacuum of 10 
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.
Yamamoto, Michiyoshi; Arbeiter, F.*; Yokomine, Takehiko*; Wakai, Eiichi; Theile, J.*; Garcia, A. S.*; Rapisarda, D. S.*; Casal, N. I.*; Mas, A. S.*; Gouat, P.*; et al.
Fusion Engineering and Design, 88(6-8), p.746 - 750, 2013/10
Times Cited Count：14 Percentile：70.64(Nuclear Science & Technology)Under Broader Approach (BA) Agreement between EURATOM and Japan, IFMIF/EVEDA (International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities) has been performed since the middle of 2007. The present EVEDA phase aims at producing a detailed, compete and fully integrated engineering design of IFMIF. The deivery of the Intermediate IFMIF engineering design report is foreseen mid-2013. The main function of the IFMIF is to provide the materials database for the design and licensing of DEMO reactor and further fusion rectors from the material sets irradiated in High Flux Test Modules (HFTM, Startup Monitoring Module), Medium Flux Test Modules (Creep Fatigues Test Module, Tritium Release Test Module, Liquid Breeder Validation Module) and Low Flux Test Modules of IFMIF. This paper is summarizing the current progress of the engineering design of the test modules within the IFMIF/EVEDA project.
Wakai, Eiichi; Kim, B. J.; Nozawa, Takashi; Kikuchi, Takayuki; Hirano, Michiko*; Kimura, Akihiko*; Kasada, Ryuta*; Yokomine, Takehiko*; Yoshida, Takahide*; Nogami, Shuhei*; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 6 Pages, 2013/03
Wakai, Eiichi; Kikuchi, Takayuki; Yokomine, Takehiko*; Yamamoto, Michiyoshi; Soldaini, M.*; Polato, A.*
Fusion Science and Technology, 62(1), p.246 - 251, 2012/07
Times Cited Count：6 Percentile：42.10(Nuclear Science & Technology)Wakai, Eiichi; Yamamoto, Michiyoshi; Molla, J.*; Yokomine, Takehiko*; Nogami, Shuhei*
Fusion Engineering and Design, 86(6-8), p.712 - 715, 2011/10
Times Cited Count：6 Percentile：43.25(Nuclear Science & Technology)The main purpose of present phase of IFMIF/EVEDA (International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities) is to produce a detailed report of IFMIF engineering design with the integrated design of all facilities in IFMIF. The main function of the IFMIF is to give the demanded design database for the licensing of DEMO reactors and further reactors, and it is achieved from the materials data set obtained from the high, medium, and low flux test modules (HFTM, MFTM and LFTM) of IFMIF. In the evaluation using small specimens, developments and guidelines of small specimen test technique or technology (SSTT) are also demanded for the achievements. This paper is summarized about the design plan and requirements in these test modules and testing items in IFMIF.
Wakai, Eiichi; Kikuchi, Takayuki; Kogawara, Takafumi; Kimura, Haruyuki; Yokomine, Takehiko*; Kimura, Akihiko*; Nogami, Shuhei*; Kurishita, Hiroaki*; Saito, Masahiro*; Nishimura, Arata*; et al.
Proceedings of 23rd IAEA Fusion Energy Conference (FEC 2010) (CD-ROM), 6 Pages, 2011/03
Japanese activities of test facilities in IFMIF-EVEDA (International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities) project have three subjects of engineering design of post irradiation examination (PIE) facilities, small specimen test technique (SSTT), and engineering design of high flux test module (HFTM), and this paper is summarized about present status. Functional analysis and design of 2-D and 3-D models of PIE facility were performed. In HFTM, as materials of heater, W-3Re alloy and/or SiC/SiC composite were selected in the points of high temperature materials, fabrication technology and some suitable properties such as resistance of thermal shock, high temperature re-crystallization, ductility, resistance of irradiation degradation, and low-activation. In SSTT, a test machine of fracture toughness was designed and developed for small specimens with 10 mm square, and it had high accuracy controllability for stress and displacement.
Wakai, Eiichi; Kikuchi, Takayuki; Hirano, Michiko; Ida, Mizuho; Niitsuma, Shigeto; Kimura, Haruyuki; Nishitani, Takeo; Yamamoto, Michiyoshi; Matsumoto, Hiroshi; Sugimoto, Masayoshi; et al.
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no abstracts in English
Wakai, Eiichi; Kogawara, Takafumi; Kikuchi, Takayuki; Yamamoto, Michiyoshi; Molla, J.*; Kimura, Akihiko*; Kasada, Ryuta*; Kim, B.*; Nogami, Shuhei*; Hasegawa, Akira*; et al.
no journal, ,
no abstracts in English
Wakai, Eiichi; Kanemura, Takuji; Furukawa, Tomohiro; Hirakawa, Yasushi; Kondo, Hiroo; Nakaniwa, Koichi; Tanaka, Hiroshi; Sugimoto, Masayoshi; Ohira, Shigeru; Yokomine, Takehiko*
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no abstracts in English
Wakai, Eiichi; Kogawara, Takafumi; Kikuchi, Takayuki; Yokomine, Takehiko*; Molla, J.*; Yamamoto, Michiyoshi
no journal, ,
no abstracts in English
Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Hirakawa, Yasushi; Furukawa, Tomohiro; Kikuchi, Takayuki; Ito, Yuzuru*; Hoashi, Eiji*; Yoshihashi, Sachiko*; Horiike, Hiroshi*; et al.
no journal, ,
no abstracts in English
Wakai, Eiichi; Kondo, Hiroo; Kanemura, Takuji; Furukawa, Tomohiro; Hirakawa, Yasushi; Nakaniwa, Koichi; Ito, Yuzuru; Tanaka, Hiroshi; Tsuji, Yoshiyuki*; Ito, Takahiro*; et al.
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no abstracts in English
Nakaniwa, Koichi; Tanaka, Hiroshi; Ito, Yuzuru; Wakai, Eiichi; Yokomine, Takehiko*
no journal, ,
In the validation test of the lithium (Li) target system facility under the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA), the evaluation test of the flow stability is performed using the 1/2.6 scale in the flow channel of the width of the Li target in the EVEDA Li Test Loop. In this study, to proceed the integrity assessment of the equipment based on the expertise obtained in the validation test of the EVEDA Li test loop, a small water experiment device which is the simulation of the target part, the downstream pipe and the quench tank arranged downstream thereof, or the modification of the flow channel of the downstream pipe, was fabricated. Along with the observation of the high-speed liquid flow by visualization experiments, to confirm the generation conditions of cavitation, an accelerometer was installed in the pipes, and flow properties evaluation was performed with the flow rate, vacuum and temperature as variables. In the observation results, the flow stability in the target up to the speed of 15m/s was confirmed and the spread of the flow in the downstream pipe with the increase in the velocity, was observed. Also, in the evaluation by an accelerometer, at the speed of 8m/s, under the pressure conditions from 20kPa to the atmospheric pressure, the noise values in the downstream pipes are small and the values hardly changed. On the other hand, the noise values were confirmed to increase as the flow velocity increases to 12m/s, 15m/s and vacuum. From now on, the intermittency of the noise phenomena will be examined in detail, and analysis of the flow stability and the optimization of the downstream pipe structure will be examined.
Watanabe, Kazuhito; Nakamura, Makoto; Someya, Yoji; Masui, Akihiro; Katayama, Kazunari*; Hayashi, Takumi; Yanagihara, Satoshi*; Konishi, Satoshi*; Yokomine, Takehiko*; Torikai, Yuji*; et al.
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In the DEMO design, the blanket primary cooling system involves high temperature pressurized water (~300C). 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.
Nakaniwa, Koichi; Tanaka, Hiroshi; Wakai, Eiichi; Yokomine, Takehiko*
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no abstracts in English
