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Journal Articles

Accuracy of prediction method of cryogenic tensile strength for austenitic stainless steels in ITER toroidal field coil structure

Sakurai, Takeru; Iguchi, Masahide; Nakahira, Masataka; Saito, Toru*; Morimoto, Masaaki*; Inagaki, Takashi*; Hong, Y.-S.*; Matsui, Kunihiro; Hemmi, Tsutomu; Kajitani, Hideki; et al.

Physics Procedia, 67, p.536 - 542, 2015/07

 Times Cited Count:3 Percentile:76.43

Japan Atomic Energy Agency (JAEA) has developed the tensile strength prediction method at liquid helium temperature (4K) using the quadratic curve as a function of the content of carbon and nitrogen in order to establish the rationalized quality control of the austenitic stainless steel used in the ITER superconducting coil operating at 4K. ITER is under construction aiming to verify technical demonstration of a nuclear fusion generation. Toroidal Field Coil (TFC), one of superconducting system in ITER, have been started procurement of materials in 2012. JAEA is producing materials for actual product which are the forged materials with shape of rectangle, round bar, asymmetry and etc. JAEA has responsibility to procure all ITER TFC Structures. In this process, JAEA obtained many tensile strength of both room temperature and 4K about these structural materials, for example, JJ1: High manganese stainless steel for structure (0.03C-12Cr-12Ni-10Mn-5Mo- 0.24N) and 316LN: High nitrogen containing stainless steel (0.2Nitrogen). Based on these data, accuracy of 4K strength prediction method for actual TFC Structure materials was evaluated and reported in this study.

Journal Articles

Progress of manufacturing trials for the ITER toroidal field coil structures

Iguchi, Masahide; Morimoto, Masaaki; Chida, Yutaka*; Hemmi, Tsutomu; Nakajima, Hideo; Nakahira, Masataka; Koizumi, Norikiyo; Yamamoto, Akio*; Miyake, Takashi*; Sawa, Naoki*

IEEE Transactions on Applied Superconductivity, 24(3), p.3801004_1 - 3801004_4, 2014/06

 Times Cited Count:6 Percentile:38.31(Engineering, Electrical & Electronic)

no abstracts in English

Journal Articles

ITER vacuum vessel, in-vessel components and plasma facing materials

Ioki, Kimihiro*; Barabash, V.*; Cordier, J.*; Enoeda, Mikio; Federici, G.*; Kim, B. C.*; Mazul, I.*; Merola, M.*; Morimoto, Masaaki*; Nakahira, Masataka*; et al.

Fusion Engineering and Design, 83(7-9), p.787 - 794, 2008/12

 Times Cited Count:18 Percentile:76.69(Nuclear Science & Technology)

This paper presents recent results of ITER activities on Vacuum Vessel (VV), blanket, limiter, and divertor. Major results can be summarized as follows. (1) The VV design is being developed in more details considering manufacturing and assembly methods, and cost. Incorporating manufacturing studies being performed in cooperation with parties, the regular VV sector design has been nearly finalized. (2) The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. (3) The divertor activities have progressed with the aim of launching the procurement according to the ITER project schedule.

Journal Articles

Design progress of the ITER in-wall shielding

Morimoto, Masaaki; Ioki, Kimihiro; Terasawa, Atsumi; Utin, Y.*

Fusion Science and Technology, 52(4), p.834 - 838, 2007/11

 Times Cited Count:1 Percentile:11.7(Nuclear Science & Technology)

The ITER in-wall shielding is mounted in between the double walls of the Vacuum Vessel. Boron-doped stainless steel and SS430 ferritic steel are used. The design improvement of the in-wall shielding has focused on reducing electromagnetic forces acting on shielding blocks. It has been found that the calculated electromagnetic forces have been significantly reduced. Magnetization forces have also been calculated for ferromagnetic inserts. Based on these load conditions, structural analyses have been performed and structural integrity has been validated. Shapes of boron-doped shielding plates which have low ductility are carefully designed to prevent excessive stress concentrations and not to take high mechanical loads. This makes shielding plate design simpler and more robust. Suitable dimensions and gaps between shielding blocks and between shielding block and the VV have been designed to fit to tolerances of the VV.

Journal Articles

Design progress of the ITER vacuum vessel sectors and port structures

Utin, Y.*; Ioki, Kimihiro; Alekseev, A.*; Bachmann, C.*; Cho, S. Y.*; Chuyanov, V.*; Jones, L.*; Kuzmin, E.*; Morimoto, Masaaki; Nakahira, Masataka; et al.

Fusion Engineering and Design, 82(15-24), p.2040 - 2046, 2007/10

 Times Cited Count:2 Percentile:19.35(Nuclear Science & Technology)

Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV sector (#1) has been nearly finalized. Design improvement of other sectors is in progress - in particular, of the VV sector #2 and #3 which interface with the ports for the neutral beam injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB Ports was advanced with the focus on the heat-flux components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements.

Journal Articles

ITER limiters moveable during plasma discharge and optimization of ferromagnetic inserts to minimize toroidal field ripple

Ioki, Kimihiro; Chuyanov, V.*; Elio, F.*; Garkusha, D.*; Gribov, Y.*; Lamzin, E.*; Morimoto, Masaaki; Shimada, Michiya; Sugihara, Masayoshi; Terasawa, Atsumi; et al.

Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03

Two important design updates have been made in the ITER VV and in-vessel components recently. One is the introduction of limiters moveable during a plasma discharge, and the other is optimization of the ferromagnetic insert configuration to minimize the toroidal field ripple. In the new limiter concept, the limiters are retracted by 8 cm during the plasma flat top phase in the divertor configuration. This concept gives important advantages: (1) the particle and heat loads due to disruptions, ELMs and blobs on the limiters will be mitigated approximately by a factor 1.5 or more; (2) the gap between the plasma and the ICRH antenna can be reduced to improve the coupling of the ICRH power. The ferromagnetic inserts have previously not been planned to be installed in the outboard midplane region between equatorial ports due to irregularity of tangential ports for NB injection. The result is a relatively large ripple (1 %) in a limited region of the plasma, which nevertheless seems acceptable from the plasma performance viewpoint. However, toroidal field flux lines fluctuate 10 mm due to the large ripple in the FW region. To avoid problems due to the large TF flux line fluctuation, additional ferromagnetic inserts are now planned to be installed in the equatorial port region.

Journal Articles

Selection of design solutions and fabrication methods and supporting R&D for procurement of ITER vessel and FW/blanket

Ioki, Kimihiro*; Elio, F.*; Maruyama, So; Morimoto, Masaaki*; Rozov, V.*; Tivey, R.*; Utin, Y.*

Fusion Engineering and Design, 74(1-4), p.185 - 190, 2005/11

 Times Cited Count:5 Percentile:36.91(Nuclear Science & Technology)

The ITER project has started preparation of Procurement Specification Documents for the vacuum vessel (VV). The design of the VV and FW/Blanket has progressed in many aspects, such as an double curvature pressing instead of facet shape welding for inner and outer shells in the upper and lower inboard regions to improve the fabrication and NDT process. The plasma facing surface of the FW has been defined to avoid protruding the leading edges, especially in the inboard area. Separate FW panels are supported with a central beam, and selection of a race-track shape cross-section for the central beam provides a more robust structure against halo current EM loads and also leads to a new cooling configuration in the shield block, where the pressure drop is significantly reduced to $$sim$$0.05 MPa. A UT R&D program is also going on to achieve acceptable S/N ratio for small-angle launching waves (20-30 deg.) to a weld. Hydraulic testing has been performed to demonstrate natural convection cooling in the transient condition.

Journal Articles

ITER nuclear components, preparing for the construction and R&D results

Ioki, Kimihiro*; Akiba, Masato; Barabaschi, P.*; Barabash, V.*; Chiocchio, S.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Ezato, Koichiro; Federici, G.*; et al.

Journal of Nuclear Materials, 329-333(1), p.31 - 38, 2004/08

 Times Cited Count:13 Percentile:65.42(Materials Science, Multidisciplinary)

The preparation of the procurement specifications is being progressed for key components. Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20 or 30 degree, on flow distribution tests of a two-channel model, on fabrication and testing of FW mockups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

Journal Articles

Glow discharge cleaning system

; ; ; Arai, Takashi; Masaki, Kei; Morimoto, Masaaki*; *

KEK Proceedings 99-17 (CD-ROM), 4 Pages, 1999/00

no abstracts in English

Journal Articles

JT-60U W-shaped divertor with in-out divertor pumping slots

Masaki, Kei; Morimoto, Masaaki*; Sakasai, Akira; Takenaga, Hidenobu; Sasajima, Tadayuki; Kodama, Kozo; Miyachi, Kengo; Hosogane, Nobuyuki

Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering (SOFE '99), p.123 - 126, 1999/00

no abstracts in English

JAEA Reports

Design and installation of W-shaped divertor in JT-60U

; Masaki, Kei; ; Morimoto, Masaaki*; *; Sakurai, Shinji; *; Saido, Masahiro; *; *; et al.

JAERI-Tech 98-049, 151 Pages, 1998/11

JAERI-Tech-98-049.pdf:6.45MB

no abstracts in English

Journal Articles

Inspection of JT-60 W-shaped divertor after the initial operation

Masaki, Kei; ; ; Morimoto, Masaaki*; *; Hosogane, Nobuyuki; Sakurai, Shinji; Saido, Masahiro

Purazuma, Kaku Yugo Gakkai-Shi, 74(9), p.1048 - 1053, 1998/09

no abstracts in English

Journal Articles

Development of a compact W-shaped pumped divertor in JT-60U

Sakurai, Shinji; Hosogane, Nobuyuki; Masaki, Kei; ; ; *; *; Shimizu, Katsuhiro; Akino, Noboru; Miyo, Yasuhiko; et al.

Fusion Engineering and Design, 39-40, p.371 - 376, 1998/00

 Times Cited Count:5 Percentile:44.99(Nuclear Science & Technology)

no abstracts in English

Journal Articles

The First inspection of JT-60U W-shaped divertor after high power operation

Masaki, Kei; ; Morimoto, Masaaki*; ; *; Hosogane, Nobuyuki; Saido, Masahiro

Fusion Technology 1998, p.67 - 70, 1998/00

no abstracts in English

Journal Articles

Installation of the W-shaped divertor in JT-60U

; ; Masaki, Kei; Hosogane, Nobuyuki; Sakurai, Shinji; Morimoto, Masaaki*; Miyo, Yasuhiko; Hiratsuka, Hajime; Akino, Noboru; *; et al.

Proc. of 17th IEEE/NPSS Symposium Fusion Engineering (SOFE'97), 2, p.365 - 368, 1998/00

no abstracts in English

Oral presentation

Status of ITER TF coil procurement

Matsui, Kunihiro; Hemmi, Tsutomu; Iguchi, Masahide; Kajitani, Hideki; Nishi, Hiroshi; Chida, Yutaka; Morimoto, Masaaki; Koizumi, Norikiyo

no journal, , 

no abstracts in English

Oral presentation

Progress of procurement of TF coil structures

Iguchi, Masahide; Hemmi, Tsutomu; Chida, Yutaka*; Morimoto, Masaaki; Hong, Y.-S.*; Nishi, Hiroshi; Koizumi, Norikiyo; Tokai, Daisuke*; Niimi, Kenichiro*; Yamada, Hirokazu*

no journal, , 

no abstracts in English

Oral presentation

Manufacturing strategy for ITER toroidal field coil structure

Hong, Y.-S.*; Iguchi, Masahide; Morimoto, Masaaki; Nakahira, Masataka; Hemmi, Tsutomu; Nishi, Hiroshi; Koizumi, Norikiyo

no journal, , 

Oral presentation

ITER TF coil structure actual manufacturing progress

Sakurai, Takeru; Iguchi, Masahide; Nakahira, Masataka; Morimoto, Masaaki; Inagaki, Takashi; Tanaka, Nobuhiko; Hong, Y.-S.*; Koizumi, Norikiyo

no journal, , 

no abstracts in English

Oral presentation

Progress of procurement of ITER TF coil structure in Japan

Iguchi, Masahide; Sakurai, Takeru; Morimoto, Masaaki*; Hong, Y.-S.*; Inagaki, Takashi; Tanaka, Nobuhiko; Nakahira, Masataka; Hemmi, Tsutomu; Matsui, Kunihiro; Koizumi, Norikiyo

no journal, , 

no abstracts in English

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