Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Edao, Yuki*; Fukada, Satoshi*; Yamaguchi, Sho*; Wu, Y.*; Nakamura, Hiroo
Fusion Engineering and Design, 85(1), p.53 - 57, 2010/01
Times Cited Count：19 Percentile：75.48(Nuclear Science & Technology)Koizumi, Norikiyo; Hemmi, Tsutomu; Matsui, Kunihiro; Nakajima, Hideo; Okuno, Kiyoshi; Kuno, Kazuo*; Nomoto, Kazuhiro*
Fusion Engineering and Design, 84(2-6), p.210 - 214, 2009/09
Times Cited Count：16 Percentile：69.53(Nuclear Science & Technology)ITER-TF coil, whose height and width are 14 m and 9 m, respectively, is scaled up by about 3 times from TF model coil (TFMC), which was developed in ITER EDA and successfully tested in EU. Although major technique of TF coil fabrication has been demonstrated in the TFMC development, new technical issues are initiated because of the scale-up. The remaining major issues are feasibilities of high accuracy automatic winding and optimization of welding of cover plates to fix the conductor in a radial plate, which is structure to enhance mechanical and electrical reliability. The authors therefore develop one of major parts of automatic winding machine, bending roller head, and successfully performed trial winding for 1/3 scale D-shaped winding. In addition, cover plate welding test was carried out using 1-m RP section and distortion of the radial plate is estimated to be in the requirement.
Molla, J.*; Nakamura, Kazuyuki
Fusion Engineering and Design, 84(2-6), p.247 - 251, 2009/06
Times Cited Count：6 Percentile：39.83(Nuclear Science & Technology)IFMIF is considered as a key element in the achievement of fusion reactors. Experiments performed in IFMIF will produce valuable information on the behavior of materials under reactor conditions. The project entered in July 2007 into EVEDA phase. The engineering design will be completed by June 2013. The Test Facilities is one of the three main systems that compose IFMIF. Irradiation of samples, remote handling between different rooms, post-irradiation studies and all the required technologies to operate the Test Facilities will be part of this system. The intense neutron flux in the Test Cell, the 
heat load in the modules, the use of tritium for in-situ experiments and the presence of high radioactivity levels in the access Cell makes the design of the Test Cell system somewhat complicated. The main challenges for the design of the Test Cell and the required Validation of new technologies will be studied in this paper.
Obana, Tetsuhiro*; Takahata, Kazuya*; Hamaguchi, Shinji*; Yanagi, Nagato*; Mito, Toshiyuki*; Imagawa, Shinsaku*; Kizu, Kaname; Tsuchiya, Katsuhiko; Hoshi, Ryo; Yoshida, Kiyoshi
Fusion Engineering and Design, 84(7-11), p.1442 - 1445, 2009/06
Times Cited Count：18 Percentile：73.13(Nuclear Science & Technology)The superconductor test facility in National Institute for Fusion Science (NIFS) has been upgraded in order to test cable-in-conduit (CIC) conductors for the JT-60SA equilibrium field (EF) coil. In the test facility, supercritical helium (SHe) lines were newly assembled with transfer tubes and a heat exchanger. The CIC conductor was covered with a thermal insulation vessel filled with gas helium at atmospheric pressure. The temperature of the conductor was varied using a film heater attached to an inlet pipe. The Ic and Tcs measurements of the prototype CIC conductor have successfully been carried out with the upgraded test facility. In the measurements, the conductor temperature was precisely controlled.
Kizu, Kaname; Tsuchiya, Katsuhiko; Obana, Tetsuhiro*; Takahata, Kazuya*; Hoshi, Ryo; Hamaguchi, Shinji*; Nunoya, Yoshihiko; Yoshida, Kiyoshi; Matsukawa, Makoto; Yanagi, Nagato*; et al.
Fusion Engineering and Design, 84(2-6), p.1058 - 1062, 2009/06
Times Cited Count：12 Percentile：60.62(Nuclear Science & Technology)The maximum magnetic field and maximum current of EF coils for JT-60SA is 6.2T, 20 kA, respectively. The EF coil conductors are NbTi cable-in-conduit (CIC) conductor with SS316L conduit. In order to confirm the performance of current sharing temperature (
) tests under coil operational condition was performed. As a results, the degradation of was 0.01-0.08 K indicating that the conductor design and its fabrication method is appropriate. Experimental results were compared with the 
and by standard plasma operation scenario. It was confirmed that the conductor has margin of 
1K.
Sato, Kazuyoshi; Omori, Junji; Kondoh, Takashi; Hatae, Takaki; Kajita, Shin*; Ishikawa, Masao; Neyatani, Yuzuru; Ebisawa, Katsuyuki*; Kusama, Yoshinori
Fusion Engineering and Design, 84(7-11), p.1713 - 1715, 2009/06
Times Cited Count：1 Percentile：9.92(Nuclear Science & Technology)Engineering analyses have been performed for the representative diagnostic upper port plug of ITER. Maintenance and integration design have been also carried out for the diagnostic components to be installed in the upper port plug. From the electromagnetic and structural analyses, it has come up an important problem to suppress the displacement of the upper port plug rather than to reduce the produced stress. Reducing the EM force will help to decrease the severity of potential displacement. Maximum displacement of the port plug decreases with increasing in the number of slits in a manner that the displacement would seem to be less than the design tolerance. A proposed low body roller and inner frame may enhance maintenance and integration. These studies and designs have established the design basis for the diagnostic upper port plug.
Miyashita, Makoto; Furuya, Kazuyuki*; Ida, Mizuho; Nakamura, Hiroo
Fusion Engineering and Design, 84(7-11), p.1333 - 1338, 2009/06
Times Cited Count：1 Percentile：9.92(Nuclear Science & Technology)The IFMIF is an accelerator-based intense neutron source for testing candidate fusion materials. Intense neutrons equivalent to neutron irradiation damage of about 50 dpa/y are emitted inside the Li flow through a back-plate. Around the back plate, a lip seal made of 316L is welded by laser welding system for replacement by remote handling. The back plate will be designed for replacement at least every year. According to material tests of the lip seal weld joint, significant deterioration was not observed. Remote handling procedure of the back plate are examined. At first, rip seal joints of connection piping will be cut by the laser welding device and then the target assembly with the back plate will be moved to a hot cell. The back plate lip seal will be cut by the laser arm in the hot cell. And a new back plate will be welded and moved to test cell/target room. In the presentation, conceptual design of the laser cutting/welding device and remote handling procedure will be shown.
Agostini, P.*; Ida, Mizuho; Miccich
, G.*; Nakamura, Hiroo; Turroni, P.*
Fusion Engineering and Design, 84(2-6), p.364 - 368, 2009/06
Times Cited Count：4 Percentile：29.39(Nuclear Science & Technology)In the International Fusion Materials Irradiation Facility (IFMIF), a back-plate of liquid lithium (Li) target is most severely irradiated by high flux of neutrons. The conceptual configuration of the IFMIF target, based on the bayonet back plate, has been developed. The life time analysis of the back-plate has to be made carefully, in order to credibly estimate its expected replacement frequency. The various interconnections between the main damaging causes were discussed in order to evidence the most plausible reasons of the back-plate malfunctioning. The results were as follows. Under IFMIF Li condition with nitrogen content of 10 wppm or less, the erosion/corrosion rate of the back-plate made of the reduced activation ferritic steel is 0.003 mm/year or less. Even in case that the Bragg peak of heat deposition by deuterons approaches to the back-plate by 1 mm, temperature of the back-plate increases by only 12 degrees or less, and induced fatigue load is well low. Thermal stress due to the neutron irradiation is 93 MPa at the back-plate center, which is well below the plastic limit. The positive shift in the ductile brittle transition temperature (DBTT), the most effective factor, reduces the back-plate lifetime to three months or shorter under irradiation damage rate of 60 dpa/year. However, the lifetime can be recovered by the post irradiation annealing heat treatment.
Nakamura, Hiroo; Agostini, P.*; Ara, Kuniaki; Fukada, Satoshi*; Furuya, Kazuyuki*; Garin, P.*; Gessii, A.*; Giusti, D.*; Groeschel, F.*; Horiike, Hiroshi*; et al.
Fusion Engineering and Design, 84(2-6), p.252 - 258, 2009/06
Times Cited Count：26 Percentile：82.93(Nuclear Science & Technology)DeEsch, H. P. L.*; Svensson, L.*; Inoue, Takashi; Taniguchi, Masaki; Umeda, Naotaka; Kashiwagi, Mieko; Fubiani, G.*
Fusion Engineering and Design, 84(2-6), p.669 - 675, 2009/06
Times Cited Count：13 Percentile：63.16(Nuclear Science & Technology)Jacquinot, J.*; Albajar, F.*; Beaumont, B.*; Becoulet, A.*; Bonicelli, T.*; Bora, D.*; Campbell, D.*; Chakraborty, A.*; Darbos, C.*; Decamps, H.*; et al.
Fusion Engineering and Design, 84(2-6), p.125 - 130, 2009/06
Times Cited Count：24 Percentile：81.06(Nuclear Science & Technology)The electron cyclotron (EC), ion cyclotron (IC), neutral beam (NB) and, lower hybrid (LH) systems for ITER have been reviewed in 2007/2008 in light of progress of physics and technology. Although the overall specifications are unchanged, notable changes have been approved. Firstly, the full 73MW should be commissioned and available on a routine basis before the D/T phase. Secondly, the possibility to operate the NB at full power during the hydrogen phase requiring new shine through protection; IC with 2 antennas with increased robustness; 2 MW transmission systems to provide an easier upgrading of the EC power; the addition of a building dedicated to the RF power sources and to a testing facility for acceptance of diagnostics and heating port plugs. Thirdly, the need of a plan for developing, in time for the active phase, a CD system such as LH suitable for very long pulse operation of ITER was recognized.
Kondo, Hiroo*; Kanemura, Takuji*; Sugiura, Hirokazu*; Yamaoka, Nobuo*; Miyamoto, Seiji*; Ida, Mizuho; Nakamura, Hiroo; Matsushita, Izuru*; Muroga, Takeo*; Horiike, Hiroshi*
Fusion Engineering and Design, 84(7-11), p.1086 - 1090, 2009/06
Times Cited Count：9 Percentile：51.65(Nuclear Science & Technology)A liquid lithium(Li) target of International Fusion Materials Irradiation Facility (IFMIF) is formed as flat plane free-surface flow by a nozzle and flows at high speed around 15 m/s. This paper focuses on flatness of the liquid Li target. A Li flow experiment was conducted in Osaka University Li Loop with a test section which was 1/2.5 scaled model of IFMIF. A thickness of the Li flow was measured and obtained by a contact method which was developed for the measurement. Analytical study on Kelvin wake and numerical calculation on wakes near side walls of the flow channel were also conducted and compared with the experimental results. As the results, positions of wake crest obtained from both of the experiment and numerical calculation assuming contact angle 140
agreed well with an iso-phase line of the analytical model. Generation of the wake are likely depends on wettability between Li and a structural material which is 304SS in the present study.
Gaio, E.*; Novello, L.*; Piovan, R.*; Shimada, Katsuhiro; Terakado, Tsunehisa; Kurihara, Kenichi; Matsukawa, Makoto
Fusion Engineering and Design, 84(2-6), p.804 - 809, 2009/06
Times Cited Count：21 Percentile：77.56(Nuclear Science & Technology)This paper deals with the conceptual design of the Quench Protection Circuits (QPC) of JT-60SA which have to provide a fast removal of the energy stored in the superconducting coils in case of quench. The core of the QPC units is constituted by a dc current breaker, which diverts the coil current into a resistor for a fast machine de-energization. In this paper, a hybrid solution, composed of a mechanical bypass switch paralleled to a static switch based on Integrated Gate Commutated Thyristor (IGCT) technology, has been chosen and worked out; a pyrobreaker is added in series to the hybrid switch as a backup protection. The resulting design allows benefiting from the fast breaking and very low maintenance requirement of the static switches, besides maintaining the advantage of the much lower power losses of the mechanical bypass in normal operation.
Kondo, Keitaro; Murata, Isao*; Klix, A.*; Seidel, K.*; Freiesleben, H.*
Fusion Engineering and Design, 84(7-11), p.1076 - 1086, 2009/06
Times Cited Count：3 Percentile：23.62(Nuclear Science & Technology)Several participants of the International Thermonuclear Experimental Reactor (ITER), such as Japan and EU, intend to introduce a Test Blanket Module (TBM) using a liquid lithium lead eutectic, which is used for the neutron multiplier and the tritium breeder. Recently a preliminary experiment in which a LiAlPb assembly was irradiated with 14 MeV neutrons was conducted at Technische Universit
t Dresden. We found out that the neutron flux inside the assembly calculated with JENDL-3.3 underestimates an experimental value in the 10-16 MeV region by around 30% and that in the 0.5-5 MeV region by around 15%, while the calculated flux with JEFF-3.1 overestimates the measurement in the 5-10 MeV region by around 20%. In order to reveal a reason of the discrepancy, problems of the nuclear data libraries for lead were investigated. As a result, the following problems of the evaluated libraries were pointed out: The cross sections of the elastic scattering in JENDL-3.3 for lead isotopes are too small and cause a significant underestimation of the neutron flux above 10 MeV, which appeared in the analysis of the above experiment. Inelastic scattering data for 
Pb in JENDL-3.3 reproduce previous experimental double-differential cross section data most well. However, those for the other lead isotopes have some problems and cause a large underestimation of the neutron flux from 0.5 to 5 MeV. The reason of the overestimation in the energy region of 5-10 MeV with JEFF-3.1 is still unclear.
Zr(n,2n) reaction cross-section at 14 MeVMurata, Isao*; Shiken, Kimiaki*; Kondo, Keitaro; Matsunaka, Masayuki*; Ota, Masayuki*; Miyamaru, Hiroyuki*; Ochiai, Kentaro; Konno, Chikara; Nishitani, Takeo
Fusion Engineering and Design, 84(7-11), p.1376 - 1379, 2009/06
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Lithium zirconate, Li
ZrO
, is known as a candidate blanket material in a fusion reactor. According to the independent benchmark studies for zirconium by JAERI, Kyoto University and Osaka University, the neutron spectrum calculations show fairly large overestimation for most evaluated nuclear data libraries. The author's group expects that the overestimation be due to a problem of evaluation for the (n,2n) reaction, because the (n,2n) reaction cross section is not well determined experimentally. In the present study, two neutrons emitted from Zr(n,2n) reaction have been measured directly to reveal the problem. As a result of measurements, the cross section obtained for energies above 1 MeV, which is the lower measurable limit energy, shows a little larger than JENDL-3.3. This is an opposite result to the benchmark analysis. However, an extrapolation for the low energy region by the evaporation spectrum with the nuclear temperature of 1 MeV brought the smaller total (n,2n) reaction cross section than JENDL-3.3, which is comparable to ENDF/B-VI. This result suggests that the discrepancies reported previously might be due to inappropriate evaluation of nuclear temperature.
Bigi, M.*; De Lorenzi, A.*; Grando, L.*; Watanabe, Kazuhiro; Yamamoto, Masanori
Fusion Engineering and Design, 84(2-6), p.446 - 450, 2009/06
Times Cited Count：16 Percentile：69.53(Nuclear Science & Technology)The ITER NBI requires a five stage -1 MV power supply system under strict load protection for electric breakdowns. A circuit model of the power supply with MAMuG accelerator has been developed for the simulation of fast transients related to the accelerator breakdowns. The complex inductive and capacitive couplings were assessed by calculations with finite element techniques. The circuit model, developed to address a number of design issues requiring simulations at system level, is now ready for use - the optimization of passive protections being the most significant application.
Nakamichi, Masaru; Ishitsuka, Etsuo; Shimakawa, Satoshi; Kan, Satoshi*
Fusion Engineering and Design, 84(7-11), p.1399 - 1403, 2009/06
Times Cited Count：1 Percentile：9.92(Nuclear Science & Technology)no abstracts in English
Nakahira, Masataka; Matsumoto, Yasuhiro; Kakudate, Satoshi; Takeda, Nobukazu; Shibanuma, Kiyoshi; Tesini, A.*
Fusion Engineering and Design, 84(7-11), p.1394 - 1398, 2009/06
Times Cited Count：23 Percentile：80.11(Nuclear Science & Technology)Invessel components of ITER have to be maintained by remote handling (RH) equipment due to high radiation level in the vacuum vessel (VV) after D-D operation. Blanket module (BM) is maintained by a manipulator mounted on a vehicle traveled through an articulated rail deployed inside the VV. Towards the construction, the BLRH equipment design has been improved and developed in more detail. The overview of design results are introduced in this paper. The design of rail deployment system of the BLRH has been updated to enable the rail connection in the transfer cask in order to minimize occupation space. For this purpose, design works have been performed for concept, sequence and typical simulation of BL replacement in the VV and rail deployment of the RH equipment in the cask, including cask docking. The technical issues of the rail connection in the cask are (1) tight tolerance of a pin at a hinge, (2) limited space of the connection inside a cask and (3) tight positioning accuracy. This paper summarizes the idea to solve these issues and a result of the design work. The paper also introduces a new cable handling equipment, rail support equipment and BL receiver/transporter.
Takeda, Nobukazu; Kakudate, Satoshi; Nakahira, Masataka; Matsumoto, Yasuhiro; Taguchi, Ko; Kozaka, Hiroshi; Shibanuma, Kiyoshi; Tesini, A.*
Fusion Engineering and Design, 84(7-11), p.1813 - 1817, 2009/06
Times Cited Count：11 Percentile：58.00(Nuclear Science & Technology)The maintenance operation of the ITER in-vessel component, such as a blanket and divertor, must be executed by the remote equipment because of the high -ray environment. During the Engineering Design Activity (EDA), the Japan Atomic Energy Agency had been fabricated the prototype of the vehicle manipulator system for the blanket remote handling and confirmed feasibility of this system including automatic positioning of the blanket and rail deployment procedure of the articulated rail. The JAEA is continuing several R&Ds so that the system can be procured smoothly to ITER. The residual key issues after the EDA are rail connection, cable handling and in-situ replacement of first wall. The last issue is newly raised and currently under the discussion. This presentation concentrates on the former two issues.
Kobayashi, Takayuki; Moriyama, Shinichi; Fujii, Tsuneyuki; Takahashi, Koji; Kajiwara, Ken; Sakamoto, Keishi
Fusion Engineering and Design, 84(2-6), p.1063 - 1067, 2009/06
Times Cited Count：10 Percentile：55.13(Nuclear Science & Technology)Design study of an ECRF antenna for JT-60SA is being carried out. An antenna concept which equips a linearly-driven flat mirror and a fixed-curved mirror was evaluated by using a numerical code. This antenna enables to alter millimeter-wave injection angle widely, only by a linear motion of the flat mirror. Therefore, the antenna eliminates a flexible tube for coolant supply and a link mechanism in vacuum vessel. In this work, it was clarified that the beam radius was expanded wider in poloidal direction compared with the expected one which was evaluated by the Gaussian optics. Then, the surface structures of the mirrors were modified in order to obtain the narrower beam radius on the resonance surface. It was found that the beam radius on the resonance surface became effectively narrower than that of the previous design by an increase in the curvature radius of the fixed-curved mirror or by a modification of the linearly-driven mirror as a diverging mirror.
