Edao, Yuki*; Fukada, Satoshi*; Yamaguchi, Sho*; Wu, Y.*; Nakamura, Hiroo
Fusion Engineering and Design, 85(1), p.53 - 57, 2010/01
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
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.
Nakamichi, Masaru; Ishitsuka, Etsuo; Shimakawa, Satoshi; Kan, Satoshi*
Fusion Engineering and Design, 84(7-11), p.1399 - 1403, 2009/06
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
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
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
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.
Hoshino, Tsuyoshi; Kato, Kenichi*; Natori, Yuri*; Nakamura, Mutsumi*; Sasaki, Kazuya*; Hayashi, Kimio; Terai, Takayuki*; Tatenuma, Katsuyoshi*
Fusion Engineering and Design, 84(2-6), p.956 - 959, 2009/06
LiTiO is one of the most promising candidates among the proposed solid breeder materials for fusion reactors. Addition of H to inert sweep gas has been proposed for enhancing the release of bred tritium from breeder material. However, the mass of LiTiO was found to decrease with time in the hydrogen atmosphere. This mass change indicates that the oxygen content of the sample decreased, suggesting the change from Ti to Ti. In order to control the mass change at the time of high temperature use, the development of lithium titanate which has LiTiO additive is expected to be effective.
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
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.
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
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.
Molla, J.*; Nakamura, Kazuyuki
Fusion Engineering and Design, 84(2-6), p.247 - 251, 2009/06
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.
Hemmi, Tsutomu; Koizumi, Norikiyo; Matsui, Kunihiro; Okuno, Kiyoshi; Nishimura, Arata*; Sakai, Masahiro*; Asano, Shiro*
Fusion Engineering and Design, 84(2-6), p.923 - 927, 2009/06
The insulation for ITER-TF coils is required to withstand a total radiation dose of 10 fast n/m. For this purpose, cyanate ester resins with high radiation-resistant have been considered instead of epoxy resins. In order to evaluate an applicability of cyanate ester resins for ITER-TF coils, the developments of the vacuum pressure impregnation technology and evaluation of the high radiation-resistant properties have been carried out. This paper presents results of these developments with the cyanater ester resin.
Kondo, Keitaro; Murata, Isao*; Klix, A.*; Seidel, K.*; Freiesleben, H.*
Fusion Engineering and Design, 84(7-11), p.1076 - 1086, 2009/06
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 Universitt 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.
Murata, 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
Lithium zirconate, LiZrO, 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.
Ota, Masayuki*; Kondo, Keitaro; Matsunaka, Masayuki*; Miyamaru, Hiroyuki*; Murata, Isao*; Iida, Toshiyuki*; Ochiai, Kentaro; Konno, Chikara
Fusion Engineering and Design, 84(7-11), p.1446 - 1449, 2009/06
In order to validate evaluated nuclear data libraries for fusion reactor designs, various integral benchmark experiments with DT neutrons have been carried out so far on structural and advanced blanket materials at the FNS facility of JAEA. In this study, a neutron spectrum shifter, which will be placed between a sample and the DT neutron source to moderate DT neutrons incident to the sample, was adopted in order to carry out the nuclear data benchmarking induced with several MeV neutrons effectively. In order to estimate effects of the spectrum shifter, the ratio of contribution of initial 14 MeV neutrons in leakage neutron and -ray spectra was calculated for the experimental configuration at FNS with a modified MCNP-4C code. The calculations were carried out for a LiTiO sample with a Be, LiD, or DO spectrum shifter. It was found that the Be shifter was superior to others and the contribution of initial 14 MeV neutrons varied depending on material and size of the sample and shifter. The present analysis also suggested that the Be shifter was effective for secondary -ray experiments.
Sakata, Shinya; Kiyono, Kimihiro; Sato, Minoru; Kominato, Toshiharu; Sueoka, Michiharu; Hosoyama, Hiroki; Kawamata, Yoichi
Fusion Engineering and Design, 84(7-11), p.1680 - 1683, 2009/06
no abstracts in English
Umeda, Naotaka; Taniguchi, Masaki; Kashiwagi, Mieko; Dairaku, Masayuki; Hanada, Masaya; Tobari, Hiroyuki; Watanabe, Kazuhiro; Sakamoto, Keishi; Inoue, Takashi
Fusion Engineering and Design, 84(7-11), p.1875 - 1880, 2009/06
The R&D of MeV class accelerator and high voltage bushing toward ITER neutral beam system are in progress at JAEA. To demonstrate the 1 MeV class negative ion beam acceleration, a MeV accelerator has been developed at MeV Test Facility. A H ion beam of 320 mA (current density: 140 A/m) was successfully accelerated up to 796 keV by fixing air leak in the ion source chamber caused by heat loading due to backstream positive ions. For development of the HV bushing, a half-size ( 800 mm) mockup single stage bushing was manufactured and tested, that demonstrated the dc voltage holding capability of up to 220 kV. Electrostatic analyses and design of the HV bushing are in progress, in which some design improvements are suggested to reduce electric field strength on the surface of ceramic insulator.
Higashijima, Satoru; Sakurai, Shinji; Suzuki, Satoshi; Yokoyama, Kenji; Kashiwa, Yoshitoshi; Masaki, Kei; Shibama, Yusuke; Takechi, Manabu; Shibanuma, Kiyoshi; Sakasai, Akira; et al.
Fusion Engineering and Design, 84(2-6), p.949 - 952, 2009/06
An upgrading device of JT-60 tokamak with fully superconducting coils (JT-60SA) is constructed under both the Japanese domestic program and the international program "Broader Approach". The maximum heat flux to JT-60SA divertor is estimated to 15 MW/m for 100 s, and a monoblock-type CFC divertor armor is promising. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin OFHC-Cu buffer layer, and the brazed joints are essential for the armor. Metalization inside CFC monoblock is applied for further improvement, and we confirmed again that the mock-up has heat removal capability in excess of ITER requirement. For optimization of the fabrication method and understanding of the production yield, the mock-ups corresponding to quantity produced in one furnace is also produced, and the half of the mock-ups could remove 15 MW/m as required. This summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.
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
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.
Bigi, M.*; De Lorenzi, A.*; Grando, L.*; Watanabe, Kazuhiro; Yamamoto, Masanori
Fusion Engineering and Design, 84(2-6), p.446 - 450, 2009/06
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.