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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.
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.
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.
Asakawa, Shuji; Yoshida, Kiyoshi
Journal of Plasma and Fusion Research SERIES, Vol.9, p.226 - 231, 2010/08
no abstracts in English
Honda, Mitsuru; Takizuka, Tomonori; Fukuyama, Atsushi*; Shimizu, Katsuhiro
Journal of Plasma and Fusion Research SERIES, Vol.9, p.529 - 534, 2010/08
A model of the neutral transport in tokamaks has been developed suitable for a one-dimensional transport code, TASK/TX. The model consists of the one-dimensional diffusion equations for slow, thermal and halo neutrals, which are simultaneously solved together with two-fluid equations and Maxwell's equation. The behavior of the slow neutrals is easily modelled by a one-dimensional diffusion equation owing to short mean free path, while that of the thermal neutrals with long mean free path is not straightforwardly. The development of the way to evaluate an effective mean free path including the information of a flux surface structure realized the reasonable estimation of the thermal neutral diffusivity compatible with one-dimensional diffusion modeling. The validity of the model has been confirmed by using a Monte Carlo code.
Nakamura, Hirofumi; Kobayashi, Kazuhiro; Yokoyama, Sumi*; Saito, Shigeru; Yamanishi, Toshihiko; Kikuchi, Kenji*
Journal of Plasma and Fusion Research SERIES, Vol.9, p.326 - 331, 2010/08
Based on results of tritium measurement in the SS316 specimens irradiated up to 5.9 dpa in the SINQ target (580 MeV proton) using a thermal desorption (TDS) method, trap site density and trap energy in the materials induced by the high-energy proton irradiation have been evaluated by means of the numerical tritium transport analysis. The results indicate that almost residual tritium in the SS316 specimen exists in the trap site, whose trap density is maximum 238 appm (5.9 dpa) and trap energy is 
1.4 eV, and that tritium release by the TDS is mainly attributed to the disappearance of the trap sites by the specimen heating, whose activation energy is about 0.7 eV. The trap site density seems to be almost proportional to the irradiation dose (dpa). Additionally, irradiation conditions such as the dose or irradiation temperature do not affect on the trapping mechanism.
Tojo, Hiroshi; Hatae, Takaki; Sakuma, Takeshi; Hamano, Takashi; Itami, Kiyoshi
Journal of Plasma and Fusion Research SERIES, Vol.9, p.288 - 293, 2010/08
Significant loss in transmissivity of the used fiber after the 18 years operations in JT-60U are found. Therefore, an optimization of the filter's configuration of the polychromator in Thomson scattering system for the high performance plasmas in JT-60SA is an essential task. Especially, loss in optical components induced by radiation from neutrons and -ray should be examined. In this paper, an optimization of the filter configuration in the polychromator is executed and expected error degradation due to the radiation damages in the measurements are also estimated with a comparison between used fibers in JT-60U and radiation resistance fibers. In 
10 keV, the time evolution of the averaged relative errors for the radiation resistance fibers increase from 3
5% by 
. However, for the used fibers in JT-60U, the errors are 78% and they reaches more than 10% in some high conditions, suggesting some difficulties in the measurements.
Ohira, Shigeru; Utsumi, Shigeo*; Kubo, Takashi; Yonemoto, Kazuhiro; Kasuya, Kenichi; Ejiri, Shintaro; Kimura, Haruyuki; Okumura, Yoshikazu
Journal of Plasma and Fusion Research SERIES, Vol.9, p.665 - 669, 2010/08
Under the Agreement Between the Government of Japan and the EURATOM for the Joint Implementation of the Broader Approach Activities (BA Activities) in the Field of Fusion Energy Research, JAEA develop a new site at Rokkasho-mura in Aomori prefecture of Japan as the Japanese Implementing Agency. In this new site, two of the three projects of the BA Activities are to be implemented, namely, International Fusion Energy Research Center (IFERC) Project and International Fusion Material Irradiation Facility/Engineering Validation and Engineering Design Activity (IFMIF/EVEDA) Project. In March 2009, the Administration and Research Building was completed, and the other research facilities; CSC&REC Building, DEMO R&D Building and IFMIF/EVEDA Accelerator Building will be completed in March 2010. In this presentation, the specifications and construction schedule of the individual research buildings will be presented, especially special features of the IFMIF/EVEDA Accelerator Building.
Wakai, Eiichi; Kogawara, Takafumi; Kikuchi, Takayuki
Journal of Plasma and Fusion Research SERIES, Vol.9, p.242 - 247, 2010/08
Recent progress of preliminary engineering design of post irradiation examination (PIE) facilities of IFMIF-EVEDA (International Fusion Materials Irradiation Facility - Engineering Validation and Engineering Design Activities) was summarized. The PIE Facilities have mainly hot cells and preparation rooms for some examinations and tests of the materials irradiated in the high-, medium- and low-flux test modules, back plate of Li target, and the others of IFMIF. In this study a basic functional analysis was performed, and handling process of specimen reloading and radiation-shielding wall in hot cells was also evaluated. Based on the results, a layout of PIE facilities was designed.
Froese, A.*; Takizuka, Tomonori; Yagi, Masatoshi
Journal of Plasma and Fusion Research SERIES, Vol.9, p.557 - 562, 2010/08
Heat flux parallell to the magnetic field in scrape-off layer is studied using the particle code PARASOL. In order to evaluate the heat load on the divertor plate, the velocity distribution functions of ions and electrons, as well as heat fluxes, and their flux limiting factors are investigated in detail, mainly about the effects of collisionality and recycling. It is found that the flux limiting factors are basically about 0.1 but varied widely from 0.001 to about 1 depending on plasma parameters.
Takayama, Arimichi*; Shimizu, Katsuhiro; Tomita, Yukihiro*; Takizuka, Tomonori
Journal of Plasma and Fusion Research SERIES, Vol.9, p.604 - 609, 2010/08
We introduce a new effective framework to combine computer codes into integrated simulation model. The framework employs MPMD (Multiple Program Multiple Data) approach and each computer code is written with MPI (Message Passing Interface) library. Adopting MPMD approach makes each computer code independent, which leads to the capability for maintenance and improvement of the integrated simulation model. The validity and usability of this approach is shown through a simple example model, which simulates the integrated divertor code, SONIC.
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.
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.
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.
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.
