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Liu, C.; Tobita, Kenji
Journal of Plasma and Fusion Research SERIES, Vol.9, p.197 - 201, 2010/10
The critical heat flux margin of the tungsten armour and the accommodation of the thermal stress between the cooling tube and the W armour was considered. The wall thickness of the F82H tube variation or as a constant were considered to analysis the heat removal capability of the cooling tube. It was found a nonlinear distribution of the peak temperature for the W armour and the top cooling tube with the tube bore rising. And q = 5 MW/m
or under the value would be acceptable based on present engineering consideration. The structure coupled analysis indicated the primary stress of the cooling tube was safety, less than 50 MPa, but thermal stress would be closed to 3Sm due to the thermal expansion between W armour and F82H tube. Based on the coupled results, a thinner tube would be better than a thicker one by considering thermal conducting and thermal stress. Finally, for the issues on CHF and thermal stress, the possible optimizations were discussed.
Shinto, Katsuhiro; Vermare, C.*
Journal of Plasma and Fusion Research SERIES, Vol.9, p.174 - 179, 2010/08
The IFMIF/EVEDA project, one of the three projects under contract with the BA agreement between EU and Japan, was started in the middle of 2007. During these two years, the design of an accelerator prototype has been progressed as the engineering validation activity and the base point of the engineering design activity for the IFMIF. The accelerator components for the prototype are being shifted to the manufacturing phase through the design reviews. In this article, the summary of the design of the prototype and the beam test plan of the prototype at Rokkasho BA site are described.
Oshima, Takayuki; Fujita, Takaaki; Seki, Masami; Kawashima, Hisato; Hoshino, Katsumichi; Shibanuma, Kiyoshi; Verrecchia, M.*; Teuchner, B.*
Journal of Plasma and Fusion Research SERIES, Vol.9, p.620 - 624, 2010/08
For interface control and assembly, the CAD data will be exchanged and integrated in a new Data Base server installed at Naka for JT-60SA, where a common computer network efficiently connected between the Naka site for JAEA and the Garching site for F4E is needed to be established. To ensure the design environments, a VPN (Virtual Private Network) was introduced with CAD LAN on computer network physically-separated from JAEA intranet area and firewall. In July 2009, a new VPN connection between the Naka and Garching sites has been successfully demonstrated using IPSec-VPN technology with a commercial and cost-effective firewall/router for security. The VPN technology would provide a common platform for the development of remote experimentation techniques on JT-60SA between Rokkasho and Naka in collaboration with activities of the ITER Remote Experimentation Centre for the IFERC Project at Rokkasho.
Sakamoto, Yoshiteru; Tobita, Kenji; Araki, Masanori
Journal of Plasma and Fusion Research SERIES, Vol.9, p.375 - 380, 2010/08
Recent tokamak experiments have achieved high fusion performances. Moreover, the ITER as the next step device will demonstrate the fusion burning with Q = 10, which provides the physics basis of burning plasma, including the behavior of energetic particle and the effect of self-heating towards DEMO reactors. The DEMO reactor requires not only the fusion performance but also the integrated performance. This paper summarizes the present status of the integrated performance achieved in the experiments to clarify the critical issues towards the DEMO reactor.
Asakura, Nobuyuki; Shimizu, Katsuhiro; Kawashima, Hisato; Tobita, Kenji; Takizuka, Tomonori
Journal of Plasma and Fusion Research SERIES, Vol.9, p.136 - 141, 2010/08
Design of the power handling for the demo tokamak reactor, SlimCS (R = 5.5 m, R/a = 2.6) with the fusion power of 3 GW and the exhausted power of 500 MW was investigated. First results of the SONIC simulation (two dimensional plasma fluid code, i.e. SOLDOR, neutral Monte Carlo code, i.e. NEUT2D, and two dimensional impurity Monte Carlo code, i.e. IMPMC) of the edge and divertor plasmas with the intensive Ar seeding were presented. Distributions of the Ar ions and radiation loss are compared with those in the previous work with a constant impurity concentration of Ar (nAr/ni) and the non-coronal radiation model for the radiation power function. At the same time, heat loading to the divertor and the first wall is investigated in the radiative plasma operation of the demo reactor.
Uto, Hiroyasu; Isono, Takaaki; Hasegawa, Mitsuru*; Tobita, Kenji; Asakura, Nobuyuki
Journal of Plasma and Fusion Research SERIES, Vol.9, p.304 - 309, 2010/08
To survey reactor concept widely, a maneuverable design tool of superconductor TF coils is required. For the purpose, a design code for systematic calculating magnetic field generated by the toroidal field (TF) coils has been developed. In this design code, the maximum magnetic field (Bmax) is obtained in that case of the superconducting material, the coil size and the operating condition. In this design code, magnetic fields generated by TF coils are estimated, taking account of the critical current density of the superconductor wires, safety at quench using the hot spot method and mechanical properties. For the quench protection, the area of stabilized copper is determined by the balance of Joule heating and heat capacity of materials in the conductor. From the area of structural materials, the von Mises stress of TF coils at a cylinder formed by inboard legs are calculated. Compared to ITER TF coils, it was confirmed that the Bmax derived from this design code is consistent.
Kobayashi, Takayuki; Isayama, Akihiko; Fasel, D.*; Yokokura, Kenji; Shimono, Mitsugu; Hasegawa, Koichi; Sawahata, Masayuki; Suzuki, Sadaaki; Terakado, Masayuki; Hiranai, Shinichi; et al.
Journal of Plasma and Fusion Research SERIES, Vol.9, p.363 - 368, 2010/08
Improvements are required for expanding the pulse length of the JT-60 ECRF system (5s) for JT-60SA (100s). Newly developed power supplies will be fabricated and installed by EU. The conditioning operation of an improved gyrotron equipping a newly designed mode convertor has been started. The mode convertor will reduce heat flux on the internal components and therefore expected to enable long pulse operation at 1 MW. Pre-programmed and/or feedback control of the heater current and anode voltage, which was successfully demonstrated in JT-60U, will be key techniques because the beam current decreases during a shot. The evacuated transmission lines have a capability of 1 MW per line. Since maintenance of the components in the vacuum vessel is difficult, a linear motion antenna concept was proposed to reduce risks of water leakage and fault of the driving mechanism in the vacuum vessel. The detailed design and the low power test of a mock-up antenna have been started.
Shibama, Yusuke; Masaki, Kei; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira
Journal of Plasma and Fusion Research SERIES, Vol.9, p.180 - 185, 2010/08
JT-60SA is a combined JA-EU satellite tokamak program, aiming at the ITER program supports as well as the supplements toward the DEMO, under both broader approach agreement and the JA domestic program. The VV is a vessel to ensure sufficient ultrahigh vacuum space and one turn toroidal resistance for plasma breakdown. A double wall structure is selected to secure the higher rigidity against operational mechanical loads. The space between walls is utilized for the neutron shielding by 323 K boron water circulation, as well as for baking at 473 K by nitrogen gas flow to achieve the vacuum less than 10
Pa. Present design status of the structural integrity is discussed with numerical analyses, which are issues of a seismic event and plasma disruptions. The feasibility of the VV manufacture is studied and latest status is presented.
Ishikawa, Masao; Kondoh, Takashi; Nishitani, Takeo; Kawano, Yasunori; Kusama, Yoshinori
Journal of Plasma and Fusion Research SERIES, Vol.9, p.43 - 47, 2010/08
Neutron transport analysis is very important for design and optimization of diagnostics in ITER. Especially, in-vessel diagnostics are exposed to strong neutron and 
radiation and then it could lead to damage and temperature increase due to nuclear heating of the components of those diagnostics. High dose rate due to strong radiation also makes those maintenances difficult. Therefore, evaluation of neutron/ flux, spectrum and nuclear heating at the location of the diagnostics with neutron transport analysis are essential to design a neutron radiation shield system and/or a cooling system. In this paper, results of neutron transport analysis applied to in-vessel components of the microfission chamber (MFC) and the poloidal polarimeter, which are developed by Japan Atomic Energy Agency, are presented.
Yatsuka, Eiichi; Hatae, Takaki; Kusama, Yoshinori
Journal of Plasma and Fusion Research SERIES, Vol.9, p.12 - 17, 2010/08
Iwai, Yasunori; Sato, Katsumi; Yamanishi, Toshihiko
Journal of Plasma and Fusion Research SERIES, Vol.9, p.332 - 337, 2010/08
We have developed some hydrophobic Pt catalysts applicable for hydrogen oxidation in the presence of saturated water vapor at room temperature. A new type of hydrophobic catalyst, Pt/ASDBC, has been prepared by depositing platinum on alkyl-styrene diviyl-benzene copolymer (ASDBC). The deposited platinum used to prepare Pt/ASDBC catalyst was 1.0 g/L. The value was approximately half of a commercial Pt/SDBC catalyst. Oxidation tests of the catalysts using 10000 ppm of hydrogen were performed in the presence of saturated water vapor at room temperature. Hydrogen oxidation more than 99% was achieved using Pt/ASDBC catalyst in the range of superficial velocity from 320 to 3300 h
. Moreover, radiation technology was applied to extend the pore size of catalyst, since the rate-controlling step of hydrogen oxidation reaction is pore diffusion. Hydrogen oxidation performance has been much improved with Pt/ASDBC catalyst irradiated with electron beams.
Miyato, Naoaki; Scott, B. D.*; Tokuda, Shinji*
Journal of Plasma and Fusion Research SERIES, Vol.9, p.546 - 551, 2010/08
Generally guiding-centre (GC) or gyro-centre fluid moments are different from corresponding particle fluid moments due to finite Larmor radius effects. Recently we derived a modified GC fundamental 1-form with strong E
B flow from which a GC Vlasov-Poisson system was also constructed through the field theory. In contrast to conventional formulations with strong EB flow, the symplectic part of our GC 1-form is the same as that in the standard gyrokinetic model formally. The GC Hamiltonian also agrees with the standard gyrokinetic Hamiltonian in the long wavelength limit. Therefore it is expected that the relation between the fluid moments in the modified GC coordinates and the particle-fluid moments is similar to the one obtained from the standard gyrokinetic model in the long wavelength limit. We represent the particle fluid moment in terms of the modified GC fluid moments. The representation is compared with the standard gyrokinetic result.
Shimada, Katsuhiro; Terakado, Tsunehisa; Matsukawa, Makoto; Cara, P.*; Baulaigue, O.*; Gaio, E.*; Coletti, R.*; Candela, G.*; Coletti, A.*
Journal of Plasma and Fusion Research SERIES, Vol.9, p.163 - 168, 2010/08
Umeda, Naotaka; Mizuno, Takatoshi; Taniguchi, Masaki; Kashiwagi, Mieko; Ezato, Koichiro; Tobari, Hiroyuki; Dairaku, Masayuki; Watanabe, Kazuhiro; Sakamoto, Keishi; Inoue, Takashi
Journal of Plasma and Fusion Research SERIES, Vol.9, p.259 - 263, 2010/08
Long pulse acceleration of ITER class H
ion beam has carried out at MeV accelerator. Melts of the acceleration grids were found around grid apertures. To accelerate higher power beam, compensation of the beam deflection and design of a new grid which has high cooling performance is required. In this study, 3D thermal transport analysis was carried out and a new acceleration grid was designed. From the analysis, it was found that the grid temperature exceeded the melting point in a few seconds. To overcome this problem, a new acceleration grid was designed whose cooling channel was drilled near upper surface. This countermeasure is effective not only to reduce the temperature rise but to enlarge the aperture size from 14 mm to 16 mm. From the result of heat analysis, temperature rise of the new grid is greatly reduced than that of the previous grid. It is expected that higher power and longer pulse beam would be accelerated at next test campaign.
Yamauchi, Kunihito; Shimada, Katsuhiro; Terakado, Tsunehisa; Matsukawa, Makoto; Cara, P.*; Gaio, E.*; Santinelli, M.*; Coletti, R.*; Coletti, A.*
Journal of Plasma and Fusion Research SERIES, Vol.9, p.220 - 225, 2010/08
Yoshida, Kiyoshi; Tsuchiya, Katsuhiko; Kizu, Kaname; Murakami, Haruyuki; Kamiya, Koji; Peyrot, M.*; Barabaschi, P.*
Journal of Plasma and Fusion Research SERIES, Vol.9, p.214 - 219, 2010/08
The upgrade of JT-60U magnet system to superconducting coils (JT-60SA) is carried out by both parties of Japan and European commission (EU) in the framework of the Broader Approach agreement. The magnet system for JT-60SA consists of 18 Toroidal Field (TF) coils, a Central Solenoid (CS) with four modules, six Equilibrium Field (EF) coils. The TF coil case encloses the winding pack and is the main structural component of the magnet system. The CS consists of four independent winding pack modules, which is support from the bottom of the TF coils. The six EF coils are attached to the TF coil cases through supports with flexible plates. The feeder system is connected from each coil to the helium refrigerator and power supply. High temperature superconducting current leads are installed in the coil terminal box. The construction of CS and EF coils was started in 2008 in Japan.
Tobari, Hiroyuki; Inoue, Takashi; Dairaku, Masayuki; Umeda, Naotaka; Kashiwagi, Mieko; Taniguchi, Masaki; Mizuno, Takatoshi; Watanabe, Kazuhiro; Sakamoto, Keishi
Journal of Plasma and Fusion Research SERIES, Vol.9, p.152 - 156, 2010/08
The high voltage (HV) bushing in ITER NBI acts as a feed through for electric power and cooling water from the -1 MV power supply in pressurized SF
gas atmosphere to negative ion source / accelerator inside vacuum. The HV bushing has five-stage structure each of which consists of a large bore ceramic ring with 1.56 m in diameter as the insulator. The ceramic is metalized and brazed with Kovar plate and then metal flanges to form the vacuum boundary as a whole. However, there is no practical example of brazing with such a large ceramic. JAEA has successfully accomplished brazing of the world's largest ceramic with Kovar plate for the first time through sample tests and mechanical analyses. Following the result, manufacturing of a mock-up simulating one-stage of the HV bushing has been completed and its vacuum insulation test is now ongoing. Electric field design inside the HV bushing for -1 MV insulation is also ongoing. In this conference, recent progresses above are reported.
Kawamura, Yoshinori; Isobe, Kanetsugu; Yamanishi, Toshihiko
Journal of Plasma and Fusion Research SERIES, Vol.9, p.358 - 362, 2010/08
Electrochemical hydrogen pump using ceramic proton conductor has been investigated to discuss about application of it for the blanket tritium recovery system of the nuclear fusion reactor. As the series of that work, the transportation experiments of H
-D mixture via ceramic proton conductor membrane have carried out. Then, the phenomenon that was caused by the exchange reaction between the deuterium in the ceramic and the protium in the gas phase have observed. So, the ceramic proton conductor which was doped deuterium was exposed to protium under the control of zero current, and the gas sampled at downstream was analyzed. It is considered that the protium in the gas phase is taken into the ceramic by isotope exchange reaction, and penetrates to the ceramic by diffusion with substitution of deuterium. And then, the reaction rate constant of isotope exchange reaction and the diffusion coefficient were also estimated.
Tanaka, Shigeru; Abe, Yuichi; Kawabe, Masaru; Kutsukake, Chuzo; Oginuma, Yoshikazu; Yamada, Masayuki; Suzuki, Takumi; Yamanishi, Toshihiko; Konno, Chikara
Journal of Plasma and Fusion Research SERIES, Vol.9, p.338 - 341, 2010/08
We have conducted a small tritium target production R&D for FNS inside JAEA. The tritium target is produced by adsorbing tritium in a thin titanium layer. Since titanium is very active to oxygen, glow discharge cleaning was carried out to remove an oxidation film of the titanium surface. Through many tests with deuterium, we found out that it was not an oxidation film but humidity to disturb tritium absorption. The following procedures were necessary; (1) to outgas the inside of an absorption chamber, (2) to keep environmental humidity under 3% in handling the titanium-deposited substrate, (3) to keep the titanium-deposited target substrate in the vacuum. The DT neutron generation performance of the tritium target produced with the above procedures was the same as that with discharge cleaning. The manufacture condition of the small target was established.
Onishi, Seiki; Maebara, Sunao; Sakaki, Hironao; Sato, Satoshi; Ochiai, Kentaro; Konno, Chikara
Journal of Plasma and Fusion Research SERIES, Vol.9, p.190 - 192, 2010/08
A new deuteron accelerator is planned to be build at Rokkasho-site in IFMIF/EVEDA and its shielding design is required urgently. Therefore, a shielding analysis has done with the prototype model of IFMIF/EVEDA accelerator vault by the Monte-Carlo transport calculation code, MCNP5, and cross section library, FENDL/MC-2.1. The neutron dose rates become 0.5 
Sv/h at the side of the beam dump and 0.05 Sv/h at the center of the beam axis. Those are smaller than the limitation dose rate of the regularly accessible controlled area, 25 Sv/h.
