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Journal Articles

Progress of ITER full tungsten divertor technology qualification in Japan

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:23 Percentile:4.82(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$$^{2}$$. 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$$^{2}$$ 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$$^{2}$$ and 1000 cycles 20 MW/m$$^{2}$$ 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$$^{2}$$ is three times longer than the requirement of ITER divertor.

Journal Articles

Assessment of operational space for long-pulse scenarios in ITER

Polevoi, A. R.*; Loarte, A.*; Hayashi, Nobuhiko; Kim, H. S.*; Kim, S. H.*; Koechl, F.*; Kukushkin, A. S.*; Leonov, V. M.*; Medvedev, S. Yu.*; Murakami, Masakatsu*; et al.

Nuclear Fusion, 55(6), p.063019_1 - 063019_8, 2015/05

 Times Cited Count:23 Percentile:13.62(Physics, Fluids & Plasmas)

Journal Articles

Kinetic modelling of divertor fluxes during ELMs in ITER

Hosokawa, Masanari*; Loarte, A.*; Huijsmans, G.*; Takizuka, Tomonori*; Hayashi, Nobuhiko

Europhysics Conference Abstracts (Internet), 38F, p.P5.003_1 - P5.003_4, 2014/06

Journal Articles

Results of the SINGAP neutral beam accelerator experiment at JAEA

DeEsch, H. P. L.*; Svensson, L.*; Inoue, Takashi; Taniguchi, Masaki; Umeda, Naotaka; Kashiwagi, Mieko; Fubiani, G.*

AIP Conference Proceedings 1097, p.353 - 363, 2009/03

CEA Cadarache and JAEA Naka have entered into a collaboration in order to test a SINGAP accelerator at the Megavolt Test Facility (MTF) at Naka, Japan. Whereas at the CEA testbed the acceleration current was limited to 0.1 A, at JAEA 0.5 A is available. This allows the acceleration of 15 H$$^{-}$$ beamlets in SINGAP to be tested and a direct comparison between SINGAP and MAMuG to be made. High-voltage conditioning in the SINGAP configuration has been quite slow, with 581 kV in vacuum achieved after 140 hours of conditioning. With 0.1 Pa of H$$_{2}$$ gas present in the accelerator 787 kV could be achieved. The conditioning curve for MAMuG is 200 kV higher. A beamlet divergence better than 5 mrad was obtained. SINGAP accelerates electrons to a higher energy than MAMuG. Based on the experiments described here, electron production by a SINGAP accelerator scaled up to ITER size was estimated to be too high for comfort.

JAEA Reports

Design study of 500 keV H$$^{-}$$ accelerator for ITER NB system

Kashiwagi, Mieko; Inoue, Takashi

JAEA-Research 2008-100, 82 Pages, 2009/02

JAEA-Research-2008-100.pdf:19.32MB

In a neutral beam (NB) system for heating and current drive of ITER, detailed designs of a multi aperture five stage accelerator to produce 1 MeV 40 A D$$^{-}$$ and 870 keV 46 A H$$^{-}$$ ion beams are ongoing. However, it was expected that the shinethrough power from 870 keV H$$^{0}$$ beam was above tolerable level for the maximum plasma density prior to any H mode. Therefore, it was required to reduce the beam energy to 500 keV with maintaining high beam current. The objective of this study is to identify necessary modifications from the original five stage accelerator to a three stage accelerator to produce 500 keV H$$^{-}$$ ion beam. The physics design of the three stage accelerator was carried out based on a beam optics study in a 2D beam analysis, a beamlet steering design in a 3D multi beamlets analysis and gas density / stripping loss of negative ions in a 3D gas analysis. Finally, the items for necessary modification were summarized.

Oral presentation

Experment of the ITER PF insert coil

Nunoya, Yoshihiko; Takahashi, Yoshikazu; Isono, Takaaki; Oshikiri, Masayuki; Kawano, Katsumi; Koizumi, Norikiyo; Hamada, Kazuya; Matsui, Kunihiro; Nabara, Yoshihiro; Hemmi, Tsutomu; et al.

no journal, , 

no abstracts in English

Oral presentation

Development of water choke model for high-voltage deck of ITER neutral beam

Tsuchida, Kazuki; Watanabe, Kazuhiro; Yamanaka, Haruhiko; Takemoto, Jumpei; Inoue, Takashi; Tanaka, Shigeru*; Yamashita, Yasuo*

no journal, , 

In ITER Neutral Beam (NB) injection system, the negative ion source and the accelerator should be cooled to generate high energy and high power beam for the prolonged operation. Those devices will be settled at high electric potential of -1 MV and an water choke is needed in the cooling water line to insulate the high voltage. The water choke for ITER should meet the following specifications; high water pressure (2 MPa) and high leak current (several ten mA) due to the reduction of resistivity in hot water (65 $$^{circ}$$C), etc. So, we developed a ceramic insulation tube for the water choke of ITER NB system and it's performance was confirmed by mechanical and withstand voltage tests. From the test results, it was confirmed the developed ceramic tube has sufficient performance ($$>$$110 kV/tube) as a part of ITER water choke. The anti-corrosion and anti-dissolution of the sealing materials in the ceramic tube was also confirmed under the high leak current condition.

Oral presentation

A Study of plasma facing tungsten components with electrical discharge machined surface exposed to cyclic thermal loads

Seki, Yohji; Ezato, Koichiro; Suzuki, Satoshi; Yokoyama, Kenji; Yamada, Hirokazu; Hirayama, Tomoyuki

no journal, , 

Through R&D for a plasma facing unit (PFU) of an ITER full-tungsten (W) divertor, Japan Atomic Energy Agency (JAEA) succeeded in demonstrating the durability of the cyclic heat loaded W divertor which was shaped by an electrical discharge machining (EDM). To prevent melting of W, an adequate technology to meet requirements of a geometrical shape and tolerance of the PFU is one of the most important key issues in a manufacturing process. JAEA has evaluated the EDM to control the final shape tolerance of $$pm$$0.25 mm. In order to examine an effect on durability of the micro-crack due to EDM, one polished W armor without the EDM and three W armors with the EDM were exposed to cyclic thermal loads. As the result, all of the W armors endured the repetitive heat load of 20 MW/m$$^{2} times$$ 1000 cycles without any macro-cracks, which strongly encourages the realization of the PFU of ITER full-W divertor with various geometrical shape and high accuracy tolerance.

Oral presentation

Manufacturing and tests of 1MV power supply for ITER NBTF

Kashiwagi, Mieko; Watanabe, Kazuhiro; Yamanaka, Haruhiko; Maejima, Tetsuya; Terunuma, Yuto*; Oda, Yuki; Tobari, Hiroyuki; Dairaku, Masayuki; Hanada, Masaya

no journal, , 

Toward the neutral beam (NB) system of ITER, the prototype of the ITER NB is under construction in the NB test facility (NBTF), Padova, Italy. For the NBTF, Japan Atomic Energy Agency manufactures, transports and constructs the 1MV high voltage power supply components to generate 1 MV, 60 A for 3600 s, which consist of fourteen components such as five DC generators and transmission lines with 100 m in a length. The manufacturing of the power supply components is in progress as scheduled. Three of five DCGs and 80% of the transmission lines have been completed. In the factory, the voltage holding test including the margin of 20% were successfully demonstrated. Then, these are under transportation to the NBTF. The construction work is started from Dec/2015 as scheduled. As the one of R&Ds results, the development of the water choke made of fiber reinforced plastic (FRP) is reported, which is the alternative of the conventional ceramic.

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