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Nishitani, Takeo; Shikama, Tatsuo*; Reichle, R.*; Hodgson, E. R.*; Ishitsuka, Etsuo; Kasai, Satoshi; Yamamoto, Shin
Fusion Engineering and Design, 63-64, p.437 - 441, 2002/12
Times Cited Count：14 Percentile：65.6(Nuclear Science & Technology)no abstracts in English
Hori, Junichi; Maekawa, Fujio; Wada, Masayuki*; Ochiai, Kentaro; Yamauchi, Michinori*; Morimoto, Yuichi*; Terada, Yasuaki; Klix, A.; Nishitani, Takeo
Fusion Engineering and Design, 63-64, p.271 - 276, 2002/12
Times Cited Count：2 Percentile：17.03(Nuclear Science & Technology)In order to the waste management method and the safety design of future D-T fusion reactor, it is important to consider the radioactivity productions via not only primary neutron reactions but also sequential charged particle reactions (SCPR). Especially, on the surface of a coolant channel many recoiled protons are generated by the neutron irradiation with coolant water, so it is apprehensive that the undesirable radioactive nuclide production yields via SCPR are enhanced. In this work, the laminated sample pieces of fusion material foils (V, Fe, W, Ti, Pb, Cu) were made and attached on a polyethylene board to simulate water flowing inside a coolant channel. They were irradiated with D-T neutrons. The effective radioactivity cross section and the depth distribution of the radioactivity production yields due to SCPR were obtained for each material. On the other hand, the estimated values were compared with the experimental ones.
Ida, Mizuho*; Nakamura, Hideo; Nakamura, Hiroshi*; Nakamura, Hiroo; Ezato, Koichiro; Takeuchi, Hiroshi
Fusion Engineering and Design, 63-64, p.333 - 342, 2002/12
Times Cited Count：43 Percentile：91.34(Nuclear Science & Technology)no abstracts in English
Honda, Tsutomu*; Hattori, Yukiya*; Holloway, C.*; Martin, E.*; Matsumoto, Yasuhiro*; Matsunobu, Takashi*; Suzuki, Toshiyuki*; Tesini, A.*; Baulo, V.*; Haange, R.*; et al.
Fusion Engineering and Design, 63-64, p.507 - 518, 2002/12
Times Cited Count：16 Percentile：69.83(Nuclear Science & Technology)The requirement to reduce the construction cost for ITER as compared with the 1998 ITER design, has led to a reduction in the size of the ITER machine and a number of design changes which have an impact on the remote maintenance of ITER. Major components to be considered for remote handling (RH) include the divertor cassettes, shield blanket modules, neutral beamline components, as well as in-port components, which are integrated with the port shield plug such as auxiliary heating equipment, limiters and test blanket modules. The design of the following equipment has been adapted for the smaller machine with reduced access space for the RH equipment: the RH equipment used for the in-vessel RH operationsto be deployed from the casks, the RH equipment that is used to remove the in-port assemblies (port plugs), as well as the remotely operated casks, which can be attached to and removed from vacuum vessel ports by using double -door systems. Defective components are loaded in transfer casks and moved to the hot cell facility by means of a remotely-operated air floatation system attached underneath the cask, where they dock against identical port interfaces and unload the component for remote refurbishment and/or waste storage.
Matsukawa, Makoto; JT-60SC Design Team
Fusion Engineering and Design, 63-64, p.519 - 529, 2002/12
Times Cited Count：15 Percentile：67.86(Nuclear Science & Technology)no abstracts in English
Morioka, Atsuhiko; Sakasai, Akira; Masaki, Kei; Ishida, Shinichi; Miya, Naoyuki; Matsukawa, Makoto; Kaminaga, Atsushi; Oikawa, Akira
Fusion Engineering and Design, 63-64, p.115 - 120, 2002/12
Times Cited Count：12 Percentile：58.11(Nuclear Science & Technology)no abstracts in English
Tanaka, Satoru*; Ohira, Shigeru; Ichimasa, Yusuke*
Fusion Engineering and Design, 63-64, p.139 - 152, 2002/12
Times Cited Count：2 Percentile：17.03(Nuclear Science & Technology)no abstracts in English
Takase, Kazuyuki; Ose, Yasuo*; Kunugi, Tomoaki*
Fusion Engineering and Design, 63-64, p.421 - 428, 2002/12
Times Cited Count：20 Percentile：76.08(Nuclear Science & Technology)no abstracts in English
Konishi, Satoshi; Nishio, Satoshi; Tobita, Kenji; DEMO Design Team
Fusion Engineering and Design, 63-64, p.11 - 17, 2002/12
Times Cited Count：50 Percentile：93.5(Nuclear Science & Technology)The first fusion power plant DEMO must have some reality that ITER and other facilities in the same period are expected to prove its feasibility. The DEMO should also be so attractive and advanced that the future society would be interested in constructing based on its concept. The present DEMO plant concept intends to satisfy these antagonistic requirements assuming construction in 2030s immediately after successful completion of fundamental ITER mission. A steady tokamak is minimized to have 5.8m of major radius with 2.3GW with Q exceeds 30. Modestly ambitious plasma parameters are chosen. Technology improvement is assumed to make maximum 20 T magnet, metal first wall and super critical water cooled ITER-like blanket modules feasible. Tritium inventory is reduced to 1kg with improved safety system concept. This conceptual design identifies various technical issues that are expected to be solved by intensive R&D efforts during ITER period, and indicates a possible step immediately after ITER.
Iwai, Yasunori; Yamanishi, Toshihiko; Ohira, Shigeru; Suzuki, Takumi; Shu, Wataru; Nishi, Masataka
Fusion Engineering and Design, 61-62, p.553 - 560, 2002/11
Times Cited Count：15 Percentile：63.36(Nuclear Science & Technology)no abstracts in English
Shu, Wataru; Gentile, C. A.*; Skinner, C. H.*; Langish, S.*; Nishi, Masataka
Fusion Engineering and Design, 61-62, p.599 - 604, 2002/11
Times Cited Count：13 Percentile：63.36(Nuclear Science & Technology)no abstracts in English
Yoshida, Hiroshi; Glugla, M.*; Hayashi, Takumi; L
sser, R.*; Murdoch, D.*; Nishi, Masataka; Haange, R.*
Fusion Engineering and Design, 61-62, p.513 - 523, 2002/11
Times Cited Count：28 Percentile：84.16(Nuclear Science & Technology)ITER tritium plant is composed of tokamak fuel cycle systems, tritium confinement and detritation systems. The tokamak fuel cycle systems, composed of various tritium sumsystems such as vacuum vessel cleaning gas processing, tokamak exhaust processing, hydrogen isotope separation, fuel storage, mixing and delivery, and external tritium receiving and long-term storage, has been designed to meet not only ITER operation scenarios but safety requirements (minimization of equipment tritium inventory and reduction of environmental tritium release at different off-normal events and accident scenarios). Multiple confinement design was employed because tritium easily permeates through metals (at 
150 
C) and plastics (at ambient temperature) and mixed with moisture in room air. That is, tritium process equipment and piping are designed to be the primary confinement barrier, and the process equipments (tritium inventory 1 g) are surrounded by the secondary confinement barrier such as a glovebox. Tritium process rooms, which contains these facilities, form the tertiary confinement barrier, and equipped with emergency isolation valves in the heating ventillation and air conditioning ducts as well as atmosphere detritiation systems. This confinement approach has been applied to tokamak building, tritium building, and hotcell and radwaste building.
Masaki, Kei; Taniguchi, Masaki; Miyo, Yasuhiko; Sakurai, Shinji; Sato, Kazuyoshi; Ezato, Koichiro; Tamai, Hiroshi; Sakasai, Akira; Matsukawa, Makoto; Ishida, Shinichi; et al.
Fusion Engineering and Design, 61-62, p.171 - 176, 2002/11
Times Cited Count：19 Percentile：74.77(Nuclear Science & Technology)A flat carbon fiber composite tile mock-up with screw tubes, which have helical fins like a nut, was fabricated aiming at further improvement of the heat removal performance of the cost-effectively manufactured divertor target for JT-60SC (modified JT-60 as a superconducting coil tokamak). The heat removal performance of the mock-up was successfully demonstrated on the JAERI Electron Beam Irradiation Stand. The estimated heat transfer coefficient of the screw tube at the non-boiling region was roughly 3 times higher than that of the smooth tube. This corresponds to 1.5 times that of the swirl tube. A heat cycle test of 10 MW/m2 showed that the mock-up with the screw tubes could withstand for 1400 cycles. These results indicate that the divertor target plate with the flat carbon fiber composite tile and the screw tube can be a promising candidate for the JT-60SC divertor target.
Kurihara, Ryoichi
Fusion Engineering and Design, 61-62, p.209 - 216, 2002/11
Times Cited Count：4 Percentile：29.25(Nuclear Science & Technology)To attain high fusion power density, the divertor must suffer high heat flux from the fusion plasma. It is very difficult to remove a high heat flux more than 20 MW/m
using the only solid divertor plate from the viewpoint of severe mechanical state such as thermal stress and crack growth. Therefore, a concept of liquid divertor is proposed to remove high heat flux by liquid films flowing on a solid wall. This paper mainly descries a preliminary thermofluid analysis of the free surface liquid flow, made of the FliBe molten salt, using the finite element analysis code ADINA-F. The heat flux of 25
100 MW/m was given on the free surface liquid of the flow. I explored a possibility of applying the secondary flow to enhance the heat transfer of the liquid flow suffering high heat flux. This analysis shows that the heat flux of 100 MW/m can be removed by inducing the secondary flow in the free surface liquid FLiBe. And this paper shows that the liquid divertor using solid-liquid multi-phase flow makes possible large heat removal by utilizing the latent heat of fusion of solid phase.
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
Times Cited Count：11 Percentile：58.11(Nuclear Science & Technology)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: 0.9
0.250.07 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.
Kawamura, Hiroshi; Takahashi, Heishichiro*; Yoshida, Naoaki*; Shestakov, V.*; Ito, Yoshio*; Uchida, Munenori*; Yamada, Hirokazu*; Nakamichi, Masaru; Ishitsuka, Etsuo
Fusion Engineering and Design, 61-62, p.391 - 397, 2002/11
Times Cited Count：40 Percentile：89.18(Nuclear Science & Technology)no abstracts in English
/SiC compositesGiancarli, L.*; Golfire, H.*; Nishio, Satoshi; Raffray, R.*; Wong, C.*; Yamada, Reiji
Fusion Engineering and Design, 61-62, p.307 - 318, 2002/11
Times Cited Count：56 Percentile：94.58(Nuclear Science & Technology)no abstracts in English
