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
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.
Morimoto, Masaaki; Ioki, Kimihiro; Terasawa, Atsumi; Utin, Y.*
Fusion Science and Technology, 52(4), p.834 - 838, 2007/11
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.
Ioki, Kimihiro; Elio, F.*; Barabash, V.*; Chuyanov, V.*; Rozov, V.*; Wang, X.*; Chen, J.*; Wang, L.*; Lorenzetto, P.*; Peacock, A.*; et al.
Fusion Engineering and Design, 82(15-24), p.1774 - 1780, 2007/10
In December 2005, the new procurement allocation plan of the ITER components among the seven Parties was prepared. The need to qualify for procurement of the specific components was especially introduced in the document. The main features and milestones of the qualification program are described in "Procurement Plan" for each specific component. Due to the complicated features of FW procurement, the procurement document has to be developed precisely. To guarantee high quality of 1700 FW panels produced by 6 different Parties, a qualification program is essential. The qualification mock-up is 80 mm wide, 240 mm long and 81 mm thick with 3 beryllium tiles 10 mm thick. Heat load tests will be performed on the qualification mock-ups in 2007 in EU and USA facilities. The maximum design heat load on the ITER FW is 0.5 MW/m (steady state) 30,000 shots. Mechanical tests of joints are also required using standardized methods. Only Parties which have satisfied the acceptance criteria of the qualification tests can proceed to the procurement stage of the ITER FW. Semi-prototypes (roughly 1000 mm 200 mm) are also requested before the ITER FW manufacturing.
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
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.
Fujieda, Hirobumi; Sugihara, Masayoshi*; Shimada, Michiya; Gribov, Y.*; Ioki, Kimihiro*; Kawano, Yasunori; Khayrutdinov, R.*; Lukash, V.*; Omori, Junji; Neyatani, Yuzuru
JAEA-Research 2007-052, 115 Pages, 2007/07
Impacts of plasma disruptions on ITER have been investigated to confirm the robustness of the design of the machine to the potential consequential loads. The loads include both electro-magnetic (EM) and heat on the in-vessel components and the vacuum vessel. Several representative disruption scenarios are specified. Disruption simulations with the DINA code and EM load analyses with a 3D finite element method code are performed for these scenarios. Some margins are confirmed in the EM load. Heat load on the first wall due to the vertical movement and the thermal quench (TQ) is calculated with a 2D heat conduction code. For vertical displacement event, beryllium () wall will not melt during the vertical movement, prior to the TQ. Significant melting is anticipated for the upper wall and tungsten baffle due to the TQ after the vertical movement. However, its impact could be mitigated by implementing a reliable detection system of the vertical movement and a mitigation system.
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.
Sato, Satoshi; Yamauchi, Michinori; Nishitani, Takeo; Ioki, Kimihiro; Iida, Hiromasa; Kataoka, Yoshiyuki
JAEA-Technology 2006-032, 91 Pages, 2006/03
no abstracts in English
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
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 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.
Shimada, Michiya; Costley, A. E.*; Federici, G.*; Ioki, Kimihiro*; Kukushkin, A. S.*; Mukhovatov, V.*; Polevoi, A. R.*; Sugihara, Masayoshi
Journal of Nuclear Materials, 337-339, p.808 - 815, 2005/03
ITER is an experimental fusion reactor for investigation and demonstration of burning plasmas, characterised of its heating dominated by alpha-particle heating. ITER is a major step from present devices and an indispensable step for fusion reactor development. ITER's success largely depends on the control of plasma-wall interactions(PWI), with power and particle fluxes and time scales one or two orders of magnitude larger than in present devices. The strategy for control of PWI includes the semi-closed divertor, strong fuelling and pumping, disruption and ELM control, replaceable plasma-facing materials and stepwise operation.
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
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.
Ioki, Kimihiro*; Barabaschi, P.*; Barabash, V.*; Chiocchio, S.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Gervash, A.*; Ibbott, C.*; Jones, L.*; et al.
Nuclear Fusion, 43(4), p.268 - 273, 2003/04
Although the basic concept of the vacuum vessel (VV) and in-vessel components of the ITER design has stayed the same, there have been several detailed design improvements resulting from efforts to raise reliability, to improve maintainability and to save money. One of the most important achievements in the VV R&D has been demonstration of the necessary fabrication and assembly tolerances. Recently the deformation due to cutting of the port extension was measured and it was shown that the deformation is small and acceptable. Further development of advanced methods of cutting, welding and NDT on a thick plate have been continued in order to refine manufacturing and improve cost and technical performance. With regard to the related FW/blanket and divertor designs, the R&D has resulted in the development of suitable technologies. Prototypes of the FW panel, the blanket shield block and the divertor components have been successfully fabricated.
Ioki, Kimihiro*; Akiba, Masato; Cardella, A.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Lorenzetto, P.*; Miki, Nobuharu*; Osaki, Toshio*; Rozov, V.*; et al.
Fusion Engineering and Design, 61-62, p.399 - 405, 2002/11
We report progress on the ITER-FEAT Blanket design and R&D during 2001-2002. Four major sub-components (FW, shield body, flexible support and electrical connection) have been highlighted. Regarding the FW, design on a separate FW panel concept has progressed, and heat load tests on a small-scale mock-up have been successfully performed with 0.7 MW/m, 13000 cycles. Full-scale separate FW panels (dimensions: 0.90.250.07 m) have been fabricated by HIPing and brazing. Regarding the shield body, a radial flow cooling design has been developed, and full-scale partial mock-ups have been fabricated by using water-jet cutting. A separate FW panel was assembled with one the shield body mock-ups. Regarding the flexible support, mill-annealed Ti (easier fabricability) alloy has been selected, and the remote assembly has been considered in the design. In mechanical tests, the requires buckling strength and mechanical fatigue characteristics have been confirmed. Regarding the electrical connection, one-body structure design without welding joints has been developed. Mechanical fatigue tests in the 3 directions, have been carried out, and thermal fatigue tests and electrical tests in a solenoidal magnetic field have been performed. Feasibility of the design has been confirmed. Through progress on design and R&D of the blanket, cost reduction has been achieved, and feasibility of design and fabricability of the components have been confirmed.
Shimomura, Yasuo; Tsunematsu, Toshihide; Yamamoto, Shin; Maruyama, So; Mizoguchi, Tadanori*; Takahashi, Yoshikazu; Yoshida, Kiyoshi; Kitamura, Kazunori*; Ioki, Kimihiro*; Inoue, Takashi; et al.
Purazuma, Kaku Yugo Gakkai-Shi, 78(Suppl.), 224 Pages, 2002/01
no abstracts in English
Onozuka, Masanori*; Ioki, Kimihiro*; Sannazzaro, G.*; Utin, Y.*; Yoshimura, Hideto*
Fusion Engineering and Design, 58-59, p.857 - 861, 2001/11
no abstracts in English
Onozuka, Masanori*; Alfile, J. P.*; Aubert, P.*; Dagenais, J.-F.*; Grebennikov, D.*; Ioki, Kimihiro*; Jones, L.*; Koizumi, Koichi; Krylov, V.*; Maslakowski, J.*; et al.
Fusion Engineering and Design, 55(4), p.397 - 410, 2001/09
Development of welding, cutting and non-destructive testing (NDT) techniques, and development of remotized systems, have been conducted for on-site manufacturing and maintenance of the thick wall structure of the ITER vacuum vessel (VV). Conventional techniques, including TIG (tungsten inert gas) welding, plasma cutting and ultrasonic inspection, have been improved and optimized for the application to thick austenitic stainless steel plates. In addition, advanced methods have been investigated including reduced-pressure electron-beam and multi-pass NdYAG (neodymium-doped yttrium aluminum garnet) laser welding, NdYAG laser cutting, and EMAT (electro-magnetic acoustic transducer) inspection to improve cost and technical performance. Two types of remotized systems with different payloads have been investigated and one of them has been fabricated and demonstrated in field joint welding, cutting, and NDT tests on test mockups and full-scale ITER VV sector models. The progress and results of this development to date provide a high level of confidence that the manufacturing and maintenance of the ITER VV is feasible.
Nakahira, Masataka; Takahashi, Hiroyuki*; Koizumi, Koichi; Onozuka, Masanori*; Ioki, Kimihiro*
Nuclear Fusion, 41(4), p.375 - 380, 2001/04
no abstracts in English
Bosia, G.*; Ioki, Kimihiro*; Kobayashi, Noriyuki*; Bibet, P.*; Koch, R.*; Chavan, R.*; Tran, M. Q.*; Takahashi, Koji; Kuzikov, S.*; Vdovin, V.*
Proceedings of IAEA 18th Fusion Energy Conference (CD-ROM), 6 Pages, 2001/00
no abstracts in English
Onozuka, Masanori*; Ioki, Kimihiro*; Johnson, G.*; Kodama, T.*; Sonnazzaro, G.*; Utin, Y.*
Fusion Engineering and Design, 51-52(Part.B), p.249 - 255, 2000/11
no abstracts in English
Ozeki, Takahisa; Iio, Shunji*; Yoshino, Ryuji; Shimada, Michiya; Fujisawa, Noboru; Ebisawa, Katsuyuki*; Ioki, Kimihiro*; Nakamura, Yukiharu; Tokuda, Shinji; Hayashi, Nobuhiko
Purazuma, Kaku Yugo Gakkai-Shi, 76(7), p.694 - 696, 2000/07
no abstracts in English
Nakahira, Masataka; Takahashi, Hiroyuki*; Inoue, O.*; Koizumi, Koichi; Shibanuma, Kiyoshi; Tada, Eisuke; Onozuka, Masanori*; Ioki, Kimihiro*
Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering (SOFE '99), p.245 - 248, 1999/00
no abstracts in English