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Tsuru, Daigo; Sakurai, Shinji; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Suzuki, Satoshi
Fusion Engineering and Design, 98-99, p.1403 - 1406, 2015/10
Times Cited Count：3 Percentile：25.85(Nuclear Science & Technology)Ezato, Koichiro; Suzuki, Satoshi; Seki, Yohji; Mori, Kensuke; Yokoyama, Kenji; Escourbiac, F.*; Hirai, Takeshi*; Kuznetsov, V.*
Fusion Engineering and Design, 98-99, p.1281 - 1284, 2015/10
Times Cited Count：40 Percentile：95.98(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) is now devoting to development of Full-W ITER divertor outer vertical target (OVT), especially, PFU that needs to withstand the repetitive heat load as high as 20MW/m
. JAEA have succeeded in demonstrating that the soundness of a bonding technology is sufficient for the full-W ITER divertor. For the development of bonding technology, the load carrying capability test on the W monoblock with a leg attachment to an OVT support structure was carried out and shows that the attachment can withstand against the uniaxial load more than 20 kN which is three times higher than the IO requirement. JAEA manufactured 6 small-scale mock-ups and tested under the repetitive heat load of 10 and 20 MW/m to examine the durability of the divertor structure including W tile bonding and the cooling tube. All of the mock-ups could survived 5000 cycles at 10 MW/m and 1000 cycles 20 MW/m with no failure such as debonding of the W tile and water leak from the cooling tube. The number of cycles at 20 MW/m is three times longer than the requirement of ITER divertor.
Enoeda, Mikio; Tanigawa, Hisashi; Hirose, Takanori; Nakajima, Motoki; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara; Kawamura, Yoshinori; Hayashi, Takumi; Yamanishi, Toshihiko; et al.
Fusion Engineering and Design, 89(7-8), p.1131 - 1136, 2014/10
Times Cited Count：21 Percentile：84.18(Nuclear Science & Technology)The development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. Regarding the fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. Also the assembling of the complete box structure of the TBM mockup and planning of the pressurization testing was studied. The development of advanced breeder and multiplier pebbles for higher chemical stability was performed for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium simulation technology, investigation of the TBM neutronics measurement technology and the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed.
Nakamura, Shigetoshi; Sakurai, Shinji; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Sakasai, Akira; Tsuru, Daigo
Fusion Engineering and Design, 89(7-8), p.1024 - 1028, 2014/10
Times Cited Count：5 Percentile：36.96(Nuclear Science & Technology)Carbon Fiber Composite mono-block divertor target is required for power handling in JT-60SA. Heat removal capability of the target is degraded by joint defect which is induced in manufacturing process. For screening heat removal capability, infrared thermography inspection (IR inspection) is improved an accuracy for the target using threaded cooling tube. In IR inspection, the targets heated at 95
C by hot water in steady state condition are instantaneously cooled down by cold water flow of 5C in three channels of test section. The heat removal capability of the targets is evaluated with comparing the transient thermal response time between defect-free and tested targets. A construction of a database for a correlation between the known defects, maximum surface temperatures in the heat load test and the IR inspection are successfully completed. Screening criteria is set with finite element methods based on the database.
Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Mori, Kensuke; Hirai, Takeshi*; Escourbiac, F.*; Kuznetsov, V.*
Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10
Nagai, Yasuki; Hashimoto, Kazuyuki; Hatsukawa, Yuichi; Saeki, Hideya; Motoishi, Shoji; Sonoda, Nozomi; Kawabata, Masako; Harada, Hideo; Kin, Tadahiro*; Tsukada, Kazuaki; et al.
Journal of the Physical Society of Japan, 82(6), p.064201_1 - 064201_7, 2013/06
Times Cited Count：43 Percentile：85.02(Physics, Multidisciplinary)Sakasai, Akira; Masaki, Kei; Shibama, Yusuke; Sakurai, Shinji; Hayashi, Takao; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Yokoyama, Kenji; Seki, Yohji; Shibanuma, Kiyoshi; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
The JT-60SA vacuum vessel (VV) and divertor are key components for the performance requirements. Therefore the manufacturing and development of VV and divertor are in progress, inclusive of the superconducting magnets. The vacuum vessel has a double wall structure in high rigidity to withstand electromagnetic force at disruption and to keep high toroidal one-turn resistance. In addition, the double wall structure fulfills originally two functions. (1) The remarkable reduction of the nuclear heating in the superconducting magnets is made by boric-acid water circulated in the double wall. (2) The effective baking is enabled by nitrogen gas flow of 200C in the double wall after draining of water. Three welding types were chosen for the manufacturing of the double wall structure VV to minimize deformation by welding. Divertor cassettes with fully water cooled plasma facing components were designed to realize the JT-60SA lower single null closed divertor. The divertor cassettes in the radio-active VV have been developed to ensure compatibility with remote handling (RH) maintenance in order to allow long pulse high performance discharges with high neutron yield. The manufacturing of divertor cassettes with typical accuracy of *1 mm has been successfully completed. Brazed CFC (carbon fiber composite) monoblock targets for a divertor target have been manufactured by precise control of tolerances inside CFC blocks. The infrared thermography test of monoblock targets has been developed as new acceptance inspection.
Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Mori, Kensuke; Enoeda, Mikio
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
The outer vertical targets of the ITER divertor are procured by the Japan Domestic Agency, JADA. Manufacturing a full-scale prototype of the half cassette which consists of 11 plasma facing units and a steel support structure has been started in Japan. JADA has greatly improved the success rate of the joint between the plasma-facing materials and heat-sink materials, in consequence of R&D on joint technology and quality control. JADA solved problems of quality control of the joint interface by an improved system of infrared thermography inspection, which provides quick feedback during the manufacturing process about the presence of defect in the joint. This paper reports on the achievements and the clarifications of technical and quality issues for the manufacture of the divertor components to be supplied by Japan.
Seki, Yohji; Ezato, Koichiro; Yokoyama, Kenji; Enoeda, Mikio; Kubota, Jinichi*; Sakamoto, Kensaku
Proceedings of 8th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-8) (USB Flash Drive), 8 Pages, 2012/12
Japan Atomic Energy Agency has been performing R&D and design of a blanket module of a nuclear fusion reactor. Pebbles of a ceramic tritium breeder are packed in a container of the blanket. Helium purge gas is applied as a transport fluid in a tritium recovery system. Prediction of the flow phenomena with a tritium transfer is important for designs of the container. A purpose of our research is to establish and verify a method for a prediction of the flow in the pebble bed. In this study, pressure drops of the helium purge gas through the pebble bed were measured up to 100 L/min of flow rate. Reliability of prediction ability of the pressure drop was validated by this experiment within the flow rate which is less than 40 L/min. A numerical simulation for the flow field through the pebble bed also has been performed. Consequently, the velocity distributions are quantitatively and qualitatively obtained at near the wall and the center region in the pebble bed.
Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Mori, Kensuke; Yokoyama, Kenji; Enoeda, Mikio
Fusion Engineering and Design, 87(5-6), p.845 - 852, 2012/08
Times Cited Count：18 Percentile：78.41(Nuclear Science & Technology)After the successful completion of the prequalification activity for ITER divertor procurement, Japanese Domestic Agency (JADA) and ITER Organization (IO) have entered into the procurement arrangement of divertor outer vertical target (OVT) in June 2009. In accordance with the arrangement, JADA has started to manufacture an OVT full-scale prototype in order to pick out and solve technical and quality issues, then to establish a rational manufacturing process toward the start of the series of production of the OVT components to be installed in tokamak. This paper presents the overview of JADA's activity on the divertor outer target procurement and also procurement schedule will be presented.
Hirose, Takanori; Tanigawa, Hisashi; Yoshikawa, Akira; Seki, Yohji; Tsuru, Daigo; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato
Fusion Engineering and Design, 86(9-11), p.2265 - 2268, 2011/10
Times Cited Count：5 Percentile：38.65(Nuclear Science & Technology)As one of the most important fabrication technologies of the WCCB TBM, Hot Isostatic Pressing (HIP) joining technology was selected to fabricate the first wall with built-in cooling channel structure made of reduced activation martensitic/ferritic steel, F82H. By using developed HIP technology, a real scale TBM first wall mockup was successfully fabricated. High heat flux test of the fabricated mockup showed the feasibility to with the equivalent conditions of the WCCB TBM operation. The breeder pebble box was successfully fabricated with thin wall cooling pipes and thin plate sleds by Laser welding. With respect to the side walls with built in cooling channels were also fabricated using drilling technology. Assembling of the first wall and side walls is one of the critical fabrication processes of the fabrication of the TBM structure. By using a F82H first wall mockup and side wall mockups, assembling process was demonstrated successfully by Electron Beam welding.
Nobuta, Yuji*; Yokoyama, Kenji; Kanazawa, Jun*; Yamauchi, Yuji*; Hino, Tomoaki*; Suzuki, Satoshi; Ezato, Koichiro; Enoeda, Mikio; Akiba, Masato
Journal of Nuclear Materials, 417(1-3), p.607 - 611, 2011/10
Times Cited Count：2 Percentile：18.29(Materials Science, Multidisciplinary)Seki, Yohji; Onishi, Yoichi*; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ozu, Akira; Ezato, Koichiro; Tsuru, Daigo; Suzuki, Satoshi; Yokoyama, Kenji; et al.
Progress in Nuclear Science and Technology (Internet), 2, p.139 - 142, 2011/10
R&D of a test blanket module (TBM) with a water-cooled solid breeder has been performed for ITER. For our design, the temperature of a coolant pressurized up to 15 MPa is designed as 598 K in an outlet of the TBM, respectively. Establishment of estimation methods of the flow phenomena is important for designs of the channel network and predictions of the material corrosion and erosion. A purpose of our research is to establish and verify the method for the prediction of the flow phenomena. The Large-eddy simulation and Reynolds averaged Navier-Stokes simulation have been performed to predict the pressure drop and flow rates in the channels of the side wall. It results the inhomogeneous flow rates in each channel. At viewpoint of the heat removal capability, however, the smallest flow-rates near the first wall are evaluated with satisfying acceptance criteria. Moreover, the results of the numerical simulation correspond with those of experiment performed for the real size mock-up.
Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Enoeda, Mikio; Mori, Seiji
Fusion Engineering and Design, 85(7-9), p.1451 - 1454, 2010/12
Times Cited Count：17 Percentile：74.1(Nuclear Science & Technology)Japan Atomic Energy Agency (JAEA) is willing to procure the outer Vertical Targets (VT) in cooperation with ITER Organization. In advance of the start of the procurement, the JAEA has to first demonstrate its technical capability to carry out the procurement. This is achieved via the successful manufactures and quality tests of VT Qualification Prototypes. Non-Destructive Examination (NDE) with the infrared thermography is required as one of the quality tests to detect the defect in the CFC monoblock, and between the CFC/OFCu. In this research and development, the Facility of Infrared NDE for Divertor (FIND) has been built by the JAEA. The FIND successfully detects the position and the magnitude of the integrated defect in the CFC and in the bonding of CFC/OFCu. The infrared NDE system established in the JAEA contributes to keeping the quality of the ITER-divertor.
Ezato, Koichiro; Seki, Yohji; Tanigawa, Hisashi; Hirose, Takanori; Tsuru, Daigo; Nishi, Hiroshi; Dairaku, Masayuki; Yokoyama, Kenji; Suzuki, Satoshi; Enoeda, Mikio
Fusion Engineering and Design, 85(7-9), p.1255 - 1260, 2010/12
Times Cited Count：12 Percentile：62.3(Nuclear Science & Technology)no abstracts in English
Seki, Yohji; Onishi, Yoichi*; Yoshikawa, Akira; Tanigawa, Hisashi; Hirose, Takanori; Ozu, Akira; Ezato, Koichiro; Tsuru, Daigo; Suzuki, Satoshi; Yokoyama, Kenji; et al.
Proceedings of Joint International Conference of 7th Supercomputing in Nuclear Application and 3rd Monte Carlo (SNA + MC 2010) (USB Flash Drive), 4 Pages, 2010/10
R&D of a test blanket module (TBM) with a water-cooled solid breeder has been performed for ITER. For our design, the temperature of a coolant pressurized up to 15 MPa is designed as 598 K in an outlet of the TBM, respectively. Establishment of estimation methods of the flow phenomena is important for designs of the channel network and predictions of the material corrosion and erosion. A purpose of our research is to establish and verify the method for the prediction of the flow phenomena. The Large-eddy simulation and Reynolds averaged Navier-Stokes simulation have been performed to predict the pressure drop and flow rates in the channels of the side wall. It results the inhomogeneous flow rates in each channel. At viewpoint of the heat removal capability, however, the smallest flow-rates near the first wall are evaluated with satisfying acceptance criteria. Moreover, the results of the numerical simulation correspond with those of experiment performed for the real size mockup.
Enoeda, Mikio; Hirose, Takanori; Tanigawa, Hisashi; Tsuru, Daigo; Yoshikawa, Akira; Seki, Yohji; Nishi, Hiroshi; Yokoyama, Kenji; Ezato, Koichiro; Suzuki, Satoshi
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), p.645 - 649, 2010/05
This paper overviews the research and development activity of Water Cooled Ceramic Breeder (WCCB) Blanket in Japan. Japan is performing development of WCCB Blanket as the primary candidate of the breeding blanket for the fusion DEMO reactor. Regarding the development of blanket module fabrication technology, a real scale First Wall (FW) was fabricated using Reduced Activation Ferritic Martensitic Steel (RAFMS) F82H. Using fabricated FW mockup, thermo-hydraulic performance and high heat flux tests were successfully performed with the heat flux equivalent to ITER TBM condition, 0.5 MW/m, 80 cycles with the coolant condition as DEMO, 15 MPa 300 C. Also, real scale Side Wall (SW) and real scale breeder pebble bed structure have been successfully fabricated. Furthermore, assembling of the real scale FW plate mockup and SW plate mockup was successfully performed. Development of major key technologies for the WCCB TBM structure fabrication has been almost completed.
Yoshikawa, Akira; Tanigawa, Hisashi; Seki, Yohji; Hirose, Takanori; Tsuru, Daigo; Ezato, Koichiro; Yokoyama, Kenji; Nishi, Hiroshi; Suzuki, Satoshi; Tanzawa, Sadamitsu; et al.
JAEA-Technology 2009-077, 23 Pages, 2010/03
JAEA-Technology-2009-077.pdf:2.62MB
In the side wall of TBM, parallel flow channels are considered. In the cooling channels structure, the flow distribution probably arises from the pressure drop in the channels. The purpose of this study is to clarify the water flow distribution in the side wall and design the cooling channels structure so that structural material of the side wall can be kept under the allowable temperature. The structural material for assumed flow rates and the flow distribution were estimated, and then the cooling channels structure was designed. The design was verified using the mockup made of the vinyl chloride pipe. For the verified design, the mockup made of F82H is manufactured, and the water flow distribution and the pressure drop were measured. It was found that the heat removal capability was sufficient in this design. From these results, the design for the cooling channels structure in the side wall is established so that enough water flow to cool the structural material is kept.
Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji; Yokoyama, Kenji; Hirose, Takanori; Mori, Seiji; Enoeda, Mikio
Physica Scripta, T138, p.014003_1 - 014003_5, 2009/12
Times Cited Count：3 Percentile：26.3(Physics, Multidisciplinary)JAEA is going to procure the whole ITER divertor outer vertical target. The qualification of the ITER divertor components has been started to validate the technical capability of the domestic agencies. JAEA has developed vertical target qualification prototypes which cover most of the critical technical issues toward the series of production. The prototypes have been high heat flux tested in Efremov institute, and showed sufficient durability at 20 MW/m. JAEA has successfully obtained the certification from ITER organization. Development of a breeding blanket is one of the most important issues to realize DEMO plant. The engineering testing of the Test Blanket Module (TBM) in ITER is a key milestone toward the DEMO blanket. R&D's on the water-cooled solid breeder blanket has been performed in JAEA. A full-scale length TBM first wall mock-up made of F82H, has been developed by using HIP bonding technique. This mock-up showed sound thermal performance in the high heat flux test.
Ezato, Koichiro; Seki, Yohji; Tanigawa, Hisashi; Hirose, Takanori; Tsuru, Daigo; Nishi, Hiroshi; Dairaku, Masayuki; Yokoyama, Kenji; Suzuki, Satoshi; Enoeda, Mikio
Proceedings of 13th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-13) (CD-ROM), 12 Pages, 2009/09
In-vessel components such as Blanket and Divertor in a fusion reactor have a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Experimental Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, there are two kinds of blanket systems will be install. One is a shield blanket, which consists of a first wall (FW) and a block module shielding against neutron flux to a vacuum chamber and a superconducting magnet system. The other blanket system is called as a Test Blanket Module (TBM). TBM is a kind of prototype blanket for a fusion power plant and has functions of breeding of tritium (T) and extraction of energy from fusion plasma. TBM consists of FW and T-breeding/neutron (n)-multiplier zone. A concept of TBM developed by JAEA is water-cooled pebble-bed type, which means that FW and other structures are cooled by pressurized high temperature water and T-breeding/n-multiplier zone consists of multiple layers of pebble bed made of T-breeding and n-multiplier material. This paper describes the status of R&Ds on FW and pebble beds from the view of thermo-hydraulics and mechanics.
