Effects of pressure profile and plasma shaping on the n=1 internal kink mode in JT-60/JT-60U pellet fuelled plasmas

International Nuclear Information System (INIS)

Ozeki, Takahisa; Azumi, Masafumi

1990-10-01

The stability of the n=1 internal kink mode in a tokamak is numerically analyzed for plasmas with a centrally peaked pressure profile. These studies are carried out with the strongly peaked pressure inside the q=1 surface, which is based on the experimentally observed plasmas by means of injections of hydrogen-ice pellets in JT-60 tokamak. The effects of peaked pressure and shaping, i.e., elongation and triangularity, are also studied for JT-60U tokamak. The plasma with the strongly peaked pressure profile has higher critical value of poloidal beta defined within the q=1 surface than that with a parabolic pressure profile. Though the beta limit reduces with the increase of the elongation, the plasma with the peaked pressure profile has larger improvement due to the triangularity than that with the parabolic pressure profile. To access the second stability of the n=1 internal kink mode, the plasma with a flat pressure profile and the large minor radius of the q=1 surface is effective. (author)

Very fast feedback control of coil-current in JT-60 tokamak

International Nuclear Information System (INIS)

Aoyagi, T.; Terakado, T.; Takahashi, M.; Nobusaka, H.; Yagyu, J.; Matsuzaki, Y.

1992-01-01

A direct digital control (DDC) system is adopted for controlling thyristor converters of power supplies in the JT-60 tokamak built in 1984. Microcomputers of the DDC were 5 MHz i8086 microprocessor and programs were written by assembler language and the processing time was under 1ms. They were, however, too old in hardware and too complicated in software. New DDC system has been made in the JT-60 Upgrade (JT-60U) to control the power supplies more quickly under 0.25 and 0.5 ms of the processing time and also to write the programs used by high-level language. The new system consists of a host computer and five microcomputers with microprocessor on VME bus system. The host computer AS3260 performs on-line processing such as setting the DDC under the discharge conditions and so on. Functions of the microcomputers with a 32-bit, 20 MHz microprocessor MC68030, whose OS are VxWorks and programs are written by C language, are real-time processing such as taking in instructions from a ZENKEI computer and in feedback control of currents and voltages of coils every 0.25 and 0.5 ms. The system is now operating very smoothly. (author)

Fabrication of the vacuum vessel for JT-60 machine upgrade

International Nuclear Information System (INIS)

Uchikawa, T.; Takanabe, K.; Tsujimura, S.; Ue, K.; Oka, K.; Kuri, S.; Ioki, K.; Namiki, K.; Suzuki, Y.; Horliike, H.; Ninomiya, H.; Yamamoto, M.; Neyatani, Y.; Ando, T.; Matsukawa, M.

1992-01-01

The JT-60 tokamak was upgraded to double the plasma current to 6 MA. In the JT-60 machine upgrade (JT-60U), the vacuum vessel and poloidal field (PF) coils were renewed. The new vacuum vessel features a three-dimensionally curved, thin double-skin torus with multi-arc D-shaped cross section. The double-skin structure is strengthened with square pipes placed in between the outer and inner skins. Fabrication and site installation of the vessel was smoothly completed in February, 1991. This paper describes an overview of the JT-60U vacuum vessel construction

JT-60 power tests from mechanical and thermal viewpoints of tokamak machine

International Nuclear Information System (INIS)

Takatsu, H.; Yamamoto, M.; Ohkubo, M.

1986-01-01

JT-60 power tests were carried out, to demonstrate, in advance of actual plasma operation, satisfactory performance of the tokamak machine, power suppliers and control system in combination. The tests began with low power ones of individual coil systems, progressed to full power ones and concluded successfully. The present paper describes the principal results of JT-60 power tests from mechanical and thermal viewpoints of tokamak machine. All of the coil systems were raised up to full power operation in combination and system performance was verified including thermal and mechanical integrity of tokamak machine. Measured strain and displacement showed good agreements with those predicted in the design, which was an evidence that electromagnetic loads were supported adequately as expected in the design. Vibration of the vacuum vessel was found to be large up to 48 m/s/sup 2/ and caused excessive vibration of the lateral port gate-valves. A few limitations to machine operation were also made clear quantatively

Physical design of JT-60 Super Upgrade

International Nuclear Information System (INIS)

Nagashima, K.; Kikuchi, M.; Kurita, G.; Ozeki, T.; Aoyagi, T.; Ushigusa, K.; Neyatani, Y.; Kubo, T.; Mori, K.; Nakagawa, S.; Kuriyama, M.; Nagami, M.

1997-01-01

The JT-60 Super Upgrade (JT-60SU) is an upgraded tokamak device of JT-60U for developing the steady-state reactor and advanced tokamak operation in the International Thermonuclear Experimental Reactor. The device is planned to utilize the JT-60 facilities fully and to minimize the needed modification. The major radius is 4.8 m and the maximum plasma current is 10 MA. Neutral beam injection with 750 keV beam energy is the primary heating method. The machine is capable of steady-state operation with high density up to 8.8 x 10 19 m -3 at 5 MA plasma current. The high operating density, over the Greenwald et al. limit, is critically important in order to achieve high bootstrap current fraction. Ballooning mode and low n ideal magnetohydrodynamic (MHD) mode including the bootstrap current were analyzed for steady-state operation. The current profile must be optimized to obtain a normalized beta up to 3. The plasma configuration with high triangularity was adopted in order to get good MHD stability and high energy confinement. A compact divertor was designed in order to get the large plasma space. (orig.)

Development of Gyrotron and JT-60U EC heating system for fusion reactor

International Nuclear Information System (INIS)

Sakamoto, K.; Kasugai, A.; Ikeda, Yo.

2003-01-01

The progress of ECH technology, for ITER and JT-60U tokamak, are presented. In the development of gyrotron, 0.9MW/9.2sec, 0.5MW/30sec, 0.3MW/60sec, etc. have been demonstrated at 170GHz. At 110GHz, 1.3MW/1.2sec, 1.2MW/4.1sec, 1MW/5sec were obtained. It is found that the reduction of the stray radiation and the enhancement of cooling capability are keys for CW operation. Four 110GHz gyrotrons are under operation in the ECH system of JT-60U. The power up to approximately 3MW/2.7sec was injected into the plasma through the poloidally movable mirrors, and contributed to the electron heating up to 26keV(n e ∼0.5x10 13 cm -3 ), and the suppression of the neo-classical tearing mode. (author)

Development of gyrotron and JT-60U EC heating system for fusion reactor

International Nuclear Information System (INIS)

Sakamoto, K.; Kasugai, A.; Ikeda, Yo.

2003-01-01

The progress of ECH technology, for ITER and JT-60U tokamak, are presented. In the development of gyrotron, 0.9MW/9.2sec, 0.5MW/30sec, 0.3MW/60sec, etc. have been demonstrated at 170GHz. At 110GHz, 1.3MW/1.2sec, 1.2MW/4. 1sec. 1MW/5sec were obtained. It is found that the reduction of the stray radiation and the enhancement of cooling capability are keys for CW operation. Four 110GHz gyrotrons are under operation in the ECH system of JT-60U. The power up to approximately 3MW/2.7sec was injected into the plasma through the poloidally movable mirrors, and contributed to the electron heating up to 26keV(n e ∼0.5x10 13 cm -3 ), and the suppression of the neo-classical tearing mode. (author)

Analyses of plasma parameter profiles in JT-60U

Energy Technology Data Exchange (ETDEWEB)

Shirai, Hiroshi; Shimizu, Katsuhiro; Hayashi, Nobuhiko [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Itakura, Hirofumi; Takase, Keizou [CSK Co. Ltd., Tokyo (Japan)

2001-01-01

The methods how diagnostics data are treated as the surface quantity of magnetic surface and processed to the profile data in the JT-60U plasmas are summarized. The MHD equilibrium obtained by solving Grad-Shafranov equation on the MHD equilibrium calculation and registration software FBEQU are saved shot by shot as a database. Various experimental plasma data measured at various geometrical positions on JT-60 are mapped onto the MHD equilibrium and treated as functions of the volume averaged minor radius {rho} on the experimental data time slice monitoring software SLICE. Experimental data are integrated and edited on SLICE. The experimental data measured as the line integral values are transformed by Able inversion. The mapped data are fitted to a functional form and saved to the profile database MAP-DB. SLICE can also read data from MAP-DB and redisplay and transform them. In addition, SLICE can generate the profile data TOKRD as run data for orbit following Monte-Carlo (OFMC) code, analyzer for current drive consistent with MHD equilibrium (ACCOME) code and tokamak predictive and interpretive code system (TOPICS). (author)

Analyses of plasma parameter profiles in JT-60U

International Nuclear Information System (INIS)

Shirai, Hiroshi; Shimizu, Katsuhiro; Hayashi, Nobuhiko

2001-01-01

The methods how diagnostics data are treated as the surface quantity of magnetic surface and processed to the profile data in the JT-60U plasmas are summarized. The MHD equilibrium obtained by solving Grad-Shafranov equation on the MHD equilibrium calculation and registration software FBEQU are saved shot by shot as a database. Various experimental plasma data measured at various geometrical positions on JT-60 are mapped onto the MHD equilibrium and treated as functions of the volume averaged minor radius ρ on the experimental data time slice monitoring software SLICE. Experimental data are integrated and edited on SLICE. The experimental data measured as the line integral values are transformed by Able inversion. The mapped data are fitted to a functional form and saved to the profile database MAP-DB. SLICE can also read data from MAP-DB and redisplay and transform them. In addition, SLICE can generate the profile data TOKRD as run data for orbit following Monte-Carlo (OFMC) code, analyzer for current drive consistent with MHD equilibrium (ACCOME) code and tokamak predictive and interpretive code system (TOPICS). (author)

Design and Structural Analysis for the Vacuum Vessel of Superconducting Tokamak JT-60SC

International Nuclear Information System (INIS)

Kudo, Y.; Sakurai, S.; Masaki, K.; Urata, K.; Sasajima, T.; Matsukawa, M.; Sakasai, A.; Ishida, S.

2003-01-01

A modification of the JT-60 is planned to be a superconducting tokamak (JT-60SC) in order to establish steady-state operation of high beta plasma for 100 s, and to ensure the applicability of ferritic steel as a reduced activation material for reactor relevant break-even class plasmas. This paper describes the detailed design of the vacuum vessel, which has a unique structure for cost effective manufacturing, as well as structural analysis results for a feasibility study

Improvement of JT-60U Negative Ion Source Performance

International Nuclear Information System (INIS)

Grisham, L.R.; Kuriyama, M.; Kawai, M.; Itoh, T.; Umeda, N.

2000-01-01

The negative ion neutral beam system now operating on JT-60U was the first application of negative ion technology to the production of beams of high current and power for conversion to neutral beams, and has successfully demonstrated the feasibility of negative ion beam heating systems for ITER and future tokamak reactors [1, 2]. It also demonstrated significant electron heating[3] and high current drive efficiency in JT-60U[4]. Because this was such a large advance in the state of the art with respect to all system parameters, many new physical processes appeared during the earlier phases of the beam injection experiments. We have explored the physical mechanisms responsible for these processes, and implemented solutions for some of them, in particular excessive beam stripping, the secular dependence of the arc and beam parameters, and nonuniformity of the plasma illuminating the beam extraction grid. This has reduced the percentage of beam heat loading on the downstream grids by roug hly a third, and permitted longer beam pulses at higher powers. Progress is being made in improving the negative ion current density, and in coping with the sensitivity of the cesium in the ion sources to oxidation by tiny air or water leaks, and the cathode operation is being altered

Tritium processing in JT-60U

International Nuclear Information System (INIS)

Miya, Naoyuki; Masaki, Kei

1997-01-01

Tritium retention analysis and tritium concentration measurement have been made during the large Tokamak JT-60U deuterium operations. This work has been carried out to evaluate the tritium retention for graphite tiles inside the vacuum vessel and tritium release characteristics in the tritium cleanup operations. JT-60U has carried out D-D experiments since July 1991. In the deuterium operations during the first two years, about 1.7 x 10 19 D-D fusion neutrons were produced by D (d, p) T reactions in plasma, which are expected to produce ∼31 GBq of tritium. The tritium produced is evacuated by a pumping system. A part of tritium is, however, trapped in the graphite tiles. Several sample tiles were removed from the vessel and the retained tritium Distribution in the tiles was measured using a liquid scintillator. The results of poloidal distribution showed that the tritium concentration in the divertor tiles was higher than that in the first wall tiles and it peaked in the tiles between two strike points of divertor magnetic lines. Tritium concentration in the exhaust gas from the vessel have also been measured with an ion chamber during the tritium cleanup operations with hydrogen divertor discharges and He-GDC. Total of recovered tritium during the cleanup operations was ∼ 7% of that generated. The results of these measurements showed that the tritium of 16-23 GBq still remained in the graphite tiles, which corresponded to about 50-70% of the tritium generated in plasma. The vessel is ventilated during the in-vessel maintenance works, then the atmosphere is always kept lower than the legal concentration guide level of 0.7 Bq/cm 3 for radiation work permit requirements. (author)

Numerical analysis of gas puff modulation experiment on JT-60U

International Nuclear Information System (INIS)

Nagashima, Keisuke; Sakasai, Akira

1992-03-01

In tokamak transport physics, source modulation experiments are one of the most effective methods. For an analysis of these modulation experiments, a simple numerical method was developed to solve the general transport equations. This method was applied to gas puff modulation experiments on JT-60U. From the comparison between the measured and calculated density perturbations, it was found that the particle diffusion coefficient is about 0.8 m 2 /sec in the edge region and 0.1-0.2 m 2 /sec in the central region. (author)

Requirements for tokamak remote operation: Application to JT-60SA

International Nuclear Information System (INIS)

Innocente, Paolo; Barbato, Paolo; Farthing, Jonathan; Giruzzi, Gerardo; Ide, Shunsuke; Imbeaux, Frédéric; Joffrin, Emmanuel; Kamada, Yutaka; Kühner, Georg; Naito, Osamu; Urano, Hajime; Yoshida, Maiko

2015-01-01

Highlights: • We analyzed the data management system (DMS) appropriate for international collaboration. • We define the principal requirements for all components of the DMS. • We evaluated application of DMS requirements to the JT-60SA experiment. • We evaluated the role network bandwidth and time delay between EU and Japan. - Abstract: Remote operation and data analysis are becoming key requirements of any fusion devices. In this framework a well-conceived data management system integrated with a suite of analysis and support tools are essential components for an efficient remote exploitation of any fusion device. The following components must be considered: data archiving data model architecture; remote data and computers access; pulse schedule, data analysis software and support tools; remote control room specifications and security issues. The definition of a device-generic data model plays also important role in improving the ability to share solution and reducing learning time. As for the remote control room, the implementation of an Operation Request Gateway has been identified as an answer to security issues meanwhile remotely proving all the required features to effectively operate a device. Previous requirements have been analyzed for the new JT-60SA tokamak device. Remote exploitation is paramount in the JT-60SA case which is expected to be jointly operated between Japan and Europe. Due to the geographical distance of the two parties an optimal remote operation and remote data-analysis is considered as a key requirement in the development of JT-60SA. It this case the effects of network speed and delay have been also evaluated and tests have confirmed that the performance can vary significantly depending on the technology used.

Requirements for tokamak remote operation: Application to JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Innocente, Paolo, E-mail: paolo.innocente@igi.cnr.it [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Barbato, Paolo [Consorzio RFX, Corso Stati Uniti 4, 35127 Padova (Italy); Farthing, Jonathan [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Giruzzi, Gerardo [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Ide, Shunsuke [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Imbeaux, Frédéric; Joffrin, Emmanuel [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Kamada, Yutaka [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Kühner, Georg [Max-Planck-Institute for Plasma Physics, EURATOM Association, Wendelsteinstr. 1, 17491 Greifswald (Germany); Naito, Osamu; Urano, Hajime; Yoshida, Maiko [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

2015-10-15

Highlights: • We analyzed the data management system (DMS) appropriate for international collaboration. • We define the principal requirements for all components of the DMS. • We evaluated application of DMS requirements to the JT-60SA experiment. • We evaluated the role network bandwidth and time delay between EU and Japan. - Abstract: Remote operation and data analysis are becoming key requirements of any fusion devices. In this framework a well-conceived data management system integrated with a suite of analysis and support tools are essential components for an efficient remote exploitation of any fusion device. The following components must be considered: data archiving data model architecture; remote data and computers access; pulse schedule, data analysis software and support tools; remote control room specifications and security issues. The definition of a device-generic data model plays also important role in improving the ability to share solution and reducing learning time. As for the remote control room, the implementation of an Operation Request Gateway has been identified as an answer to security issues meanwhile remotely proving all the required features to effectively operate a device. Previous requirements have been analyzed for the new JT-60SA tokamak device. Remote exploitation is paramount in the JT-60SA case which is expected to be jointly operated between Japan and Europe. Due to the geographical distance of the two parties an optimal remote operation and remote data-analysis is considered as a key requirement in the development of JT-60SA. It this case the effects of network speed and delay have been also evaluated and tests have confirmed that the performance can vary significantly depending on the technology used.

Carbon transport and fuel retention in JT-60U with high temperature operation based on postmortem analysis

Energy Technology Data Exchange (ETDEWEB)

Yoshida, M., E-mail: yoshida.masafumi@jaea.go.jp [Japan Atomic Energy Agency, Mukoyama 801-1, Naka-shi, Ibaraki-ken 311-0193 (Japan); Tanabe, T.; Adachi, A. [Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Hayashi, T.; Nakano, T.; Fukumoto, M.; Yagyu, J.; Miyo, Y.; Masaki, K.; Itami, K. [Japan Atomic Energy Agency, Mukoyama 801-1, Naka-shi, Ibaraki-ken 311-0193 (Japan)

2013-07-15

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas, or tile gaps and remote areas, in JT-60U were measured. The total fuel retention rate in the plasma shadowed areas was more than two times higher than that in the carbon re-deposited layers on the plasma facing surfaces, or the divertor tiles. This is because of lower temperature in the plasma shadowed areas than in the plasma facing surfaces, which leads to high hydrogen saturation concentration, although the amount of the carbon re-deposited on the plasma shadowed areas was only 60% of that on the plasma facing surfaces. The total fuel retention rate in JT-60U, including previously determined for all the plasma facing areas, was evaluated to be 1.3 × 10{sup 20} H + D s{sup −1}, and this retention rate was lower than that in the other devices, due probably to high baking temperature operation in JT-60U. Distributions of the fuel retention and the carbon re-deposition in the whole in-vessel of a large tokamak were determined for the first time in the world.

Carbon transport and fuel retention in JT-60U with high temperature operation based on postmortem analysis

International Nuclear Information System (INIS)

Yoshida, M.; Tanabe, T.; Adachi, A.; Hayashi, T.; Nakano, T.; Fukumoto, M.; Yagyu, J.; Miyo, Y.; Masaki, K.; Itami, K.

2013-01-01

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas, or tile gaps and remote areas, in JT-60U were measured. The total fuel retention rate in the plasma shadowed areas was more than two times higher than that in the carbon re-deposited layers on the plasma facing surfaces, or the divertor tiles. This is because of lower temperature in the plasma shadowed areas than in the plasma facing surfaces, which leads to high hydrogen saturation concentration, although the amount of the carbon re-deposited on the plasma shadowed areas was only 60% of that on the plasma facing surfaces. The total fuel retention rate in JT-60U, including previously determined for all the plasma facing areas, was evaluated to be 1.3 × 10 20 H + D s −1 , and this retention rate was lower than that in the other devices, due probably to high baking temperature operation in JT-60U. Distributions of the fuel retention and the carbon re-deposition in the whole in-vessel of a large tokamak were determined for the first time in the world

Review of ICRF antenna development and heating experiments up to advanced experiment I, 1989 on the JT-60 tokamak

International Nuclear Information System (INIS)

Fujii, Tsuneyuki

1992-03-01

Two main subjects of ion cyclotron range of frequencies (ICRF) heating on JT-60 are described in this paper from development phase of the JT-60 ICRF heating system up to advanced experiment I, 1989. One is antenna design and development for the high power JT-60 ICRF heating system (6 MW for 10 s at a frequency range of 108 - 132 MHz). The other is the experimental investigation of characteristics of second harmonic ICRF heating in a large tokamak. (J.P.N.)

Progress of JT-60SA Project: EU-JA joint efforts for assembly and fabrication of superconducting tokamak facilities and its research planning

Energy Technology Data Exchange (ETDEWEB)

Shirai, Hiroshi, E-mail: shirai.hiroshi@jaea.go.jp [JT-60SA Project Team, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Barabaschi, Pietro [JT-60SA EU-Home Team, Fusion for Energy, Boltsmannstr 2, Garching 85748 (Germany); Kamada, Yutaka [JT-60SA JA-Home Team, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2016-11-01

Highlights: • JT-60SA Project is promoted under the BA Agreement and JA national programme. • JT-60SA is designed to operate in break-even equivalent condition for a long period. • JT-60SA Project supports and complements the ITER project, and promotes DEMO design. • Fabrication of JT-60SA components and assembly in Naka are steadily going on. • JT-60SA Research Plan has been developed jointly by EU and JA fusion communities. - Abstract: Aiming at supporting the early realization of fusion energy, the JT-60SA Project has shown steady progress for several years toward the first plasma in 2019 under the dual frameworks: the Satellite Tokamak Programme of the Broader Approach Agreement between EU and Japan, and the Japanese national programme. JT-60SA is a superconducting tokamak designed to operate in break-even equivalent conditions for a long pulse duration (typically 100 s) with a maximum plasma current of 5.5 MA. A variety of plasma control capabilities enable JT-60SA to contribute directly to the ITER project and also to DEMO by addressing key engineering and physics issues for advanced plasma operation. Design and fabrication of JT-60SA components, shared by the EU and Japan, started in 2007. Assembly in the torus hall started in January 2013, and welding work of the vacuum vessel sectors (seven 40° sectors and two 30° sectors) is currently ongoing on the cryostat base. Other components such as TF coils, PF coils, power supplies, cryogenic system, cryostat vessel, thermal shields and so on were or are being delivered to the Naka site for installation, assembly and commissioning. This paper gives technical progress on fabrication, installation and assembly of tokamak components and ancillary systems, as well as progress of the JT-60SA Research Plan being developed jointly by European and Japanese fusion communities.

Advanced tokamak research with integrated modeling in JT-60 Upgrade

International Nuclear Information System (INIS)

Hayashi, N.

2010-01-01

Researches on advanced tokamak (AT) have progressed with integrated modeling in JT-60 Upgrade [N. Oyama et al., Nucl. Fusion 49, 104007 (2009)]. Based on JT-60U experimental analyses and first principle simulations, new models were developed and integrated into core, rotation, edge/pedestal, and scrape-off-layer (SOL)/divertor codes. The integrated models clarified complex and autonomous features in AT. An integrated core model was implemented to take account of an anomalous radial transport of alpha particles caused by Alfven eigenmodes. It showed the reduction in the fusion gain by the anomalous radial transport and further escape of alpha particles. Integrated rotation model showed mechanisms of rotation driven by the magnetic-field-ripple loss of fast ions and the charge separation due to fast-ion drift. An inward pinch model of high-Z impurity due to the atomic process was developed and indicated that the pinch velocity increases with the toroidal rotation. Integrated edge/pedestal model clarified causes of collisionality dependence of energy loss due to the edge localized mode and the enhancement of energy loss by steepening a core pressure gradient just inside the pedestal top. An ideal magnetohydrodynamics stability code was developed to take account of toroidal rotation and clarified a destabilizing effect of rotation on the pedestal. Integrated SOL/divertor model clarified a mechanism of X-point multifaceted asymmetric radiation from edge. A model of the SOL flow driven by core particle orbits which partially enter the SOL was developed by introducing the ion-orbit-induced flow to fluid equations.

Development of computer-aided software engineering tool for sequential control of JT-60U

International Nuclear Information System (INIS)

Shimono, M.; Akasaka, H.; Kurihara, K.; Kimura, T.

1995-01-01

Discharge sequential control (DSC) is an essential control function for the intermittent and pulse discharge operation of a tokamak device, so that many subsystems may work with each other in correct order and/or synchronously. In the development of the DSC program, block diagrams of logical operation for sequential control are illustrated in its design at first. Then, the logical operators and I/O's which are involved in the block diagrams are compiled and converted to a certain particular form. Since the block diagrams of the sequential control amounts to about 50 sheets in the case of the JT-60 upgrade tokamak (JT-60U) high power discharge and the above steps of the development have been performed manually so far, a great effort has been required for the program development. In order to remove inefficiency in such development processes, a computer-aided software engineering (CASE) tool has been developed on a UNIX workstation. This paper reports how the authors design it for the development of the sequential control programs. The tool is composed of the following three tools: (1) Automatic drawing tool, (2) Editing tool, and (3) Trace tool. This CASE tool, an object-oriented programming tool having graphical formalism, can powerfully accelerate the cycle for the development of the sequential control function commonly associated with pulse discharge in a tokamak fusion device

Assembly work and transport of JT-60SA cryostat base

International Nuclear Information System (INIS)

Okano, Fuminori; Masaki, Kei; Yagyu, Jun-ichi; Shibama, Yusuke; Sakasai, Akira; Miyo, Yasuhiko; Kaminaga, Atsushi; Nishiyama, Tomokazu; Suzuki, Sadaaki; Nakamura, Shigetoshi; Shibanuma, Kiyoshi

2013-11-01

Japan Atomic Energy Agency started to construct a fully superconducting tokamak experiment device, JT-60SA, to support the ITER since January, 2013 at the Fusion Research and Development Directorate in Naka, Japan. The JT-60SA will be constructed with enhancing the previous JT-60 infrastructures, in the JT-60 torus hall, where the ex-JT-60 machine was disassembled. The JT-60SA Cryostat Base, for base of the entire tokamak structure, were assembly as the first step of this construction. The Cryostat Base (CB, 250tons) is consists of 7 main components made of stainless steel, in 12 m diameter and 3 m height. The CB was built in the Spain and transported to the Naka site, via Hitachi port. After pre-assembly work including preliminary measurements and sole plate adjustments of its height/flatness, the JT-60SA CB was carefully set on the sole plate. JT-60SA CB was assembled with high accuracy by using a laser tracker. The CB was adjusted in the height and flatness against the assembly reference position and determined by the absolute coordinates. This report introduces the concrete result of assembly work and transport of JT-60SA CB. (author)

Physics and operation oriented activities in preparation of the JT-60SA tokamak exploitation

Science.gov (United States)

Giruzzi, G.; Yoshida, M.; Artaud, J. F.; Asztalos, Ö.; Barbato, E.; Bettini, P.; Bierwage, A.; Boboc, A.; Bolzonella, T.; Clement-Lorenzo, S.; Coda, S.; Cruz, N.; Day, Chr.; De Tommasi, G.; Dibon, M.; Douai, D.; Dunai, D.; Enoeda, M.; Farina, D.; Figini, L.; Fukumoto, M.; Galazka, K.; Galdon, J.; Garcia, J.; Garcia-Muñoz, M.; Garzotti, L.; Gil, C.; Gleason-Gonzalez, C.; Goodman, T.; Granucci, G.; Hayashi, N.; Hoshino, K.; Ide, S.; Imazawa, R.; Innocente, P.; Isayama, A.; Itami, K.; Joffrin, E.; Kamada, Y.; Kamiya, K.; Kawano, Y.; Kawashima, H.; Kobayashi, T.; Kojima, A.; Kubo, H.; Lang, P.; Lauber, Ph.; de la Luna, E.; Maget, P.; Marchiori, G.; Mastrostefano, S.; Matsunaga, G.; Mattei, M.; McDonald, D. C.; Mele, A.; Miyata, Y.; Moriyama, S.; Moro, A.; Nakano, T.; Neu, R.; Nowak, S.; Orsitto, F. P.; Pautasso, G.; Pégourié, B.; Pigatto, L.; Pironti, A.; Platania, P.; Pokol, G. I.; Ricci, D.; Romanelli, M.; Saarelma, S.; Sakurai, S.; Sartori, F.; Sasao, H.; Scannapiego, M.; Shimizu, K.; Shinohara, K.; Shiraishi, J.; Soare, S.; Sozzi, C.; Stępniewski, W.; Suzuki, T.; Suzuki, Y.; Szepesi, T.; Takechi, M.; Tanaka, K.; Terranova, D.; Toma, M.; Urano, H.; Vega, J.; Villone, F.; Vitale, V.; Wakatsuki, T.; Wischmeier, M.; Zagórski, R.

2017-08-01

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised.

Effect of halo current and its toroidal asymmetry during disruptions in JT-60U

International Nuclear Information System (INIS)

Neyatani, Y.; Yoshino, R.; Ando, T.

1995-01-01

A poloidal halo current due to a vertical displacement event (VDE) is observed in experimentally simulated VDE discharges and density limit disruptions in the JT-60U tokamak. In the case of a clockwise I p and B T discharge, the halo current flows into the vacuum vessel from the inside separatrix and goes back to the plasma from the outside separatrix. A maximum halo current is produced by a change in the poloidal flux generated by plasma current decay. A toroidal asymmetry factor of 2.5 is estimated from the requirements of the fracture of the carbon-fiber composite tiles. The toroidal asymmetry is caused by the poloidal field (PF) that is produced by the toroidal field (TF) ripple, the deformation of the vacuum vessel, the setting error between the vacuum vessel and the TF and PF coils, the low-n mode during current quench, etc. To consider this asymmetry, in JT-60U, one must estimate the total halo current as nearly 26% of the plasma current just before a current quench. 25 refs., 10 figs

Activation analysis for JT-60U experiments with deuterium gases

International Nuclear Information System (INIS)

Miya, Naoyuki

1993-11-01

Identification of radionuclides and evaluation of dose rate level have been made on the structural materials of the JT-60U tokamak device. A one-dimensional neutron and gamma-ray transport code ANISN and an induced activation calculation code CINAC are used in this work. Radionuclides of 56 Mn (High-Mn steel toroidal field coil case), 58 Co (Inconel-625 vessel) and 60 Co (SS-316 first wall supporting material) appeared on the structures, which contribute to the dose rate around a vacuum vessel. Cobalt-58 and 60 Co with long half-life time intensely make residual activation in the time of 3 days to 3 months corresponding to the maintenance time after shutdown. The calculated dose rate on the vessel agreed well with the measured data in the first 2 years D-D operations. The one-dimensional code provided a sufficient prediction for the dose rate, although an error due to the toroidal field coil modeling in the calculation is estimated within ∼30%. (author)

Progress in JT-60 innovative technologies

Energy Technology Data Exchange (ETDEWEB)

NONE

2005-09-01

This review report provides the synthetic archives of innovative technologies in 20-year facility developments for the large tokamak experimental device JT-60, first founded as the major magnetic fusion device in the Second Basic Program for Fusion Research and Development of Japan. Manufacture of JT-60 was started in 1978, and the first plasma was achieved on April 8, 1985. In 1989-1991, the vacuum vessel and poloidal field coils were entirely reconfigured to improve the plasma performance. The major original mission of the JT-60 project, a breakeven condition for a D-T equivalent plasma, was finally attained in 1996. After this, JT-60, as a leading device for magnetic fusion research in the world, continues to challenge many experimental issues, which has been achieved by collaboration of innovative facility developments and experimental improvements. In addition, at this time to start the ITER construction phase in 2005, this review is expected to contribute the construction and operation activities for the next generation tokamak by providing the basic ideas in facility developments. We classified a tremendous number of development items into the selected sections for this review. Since the authors have been in charge of each development activity of their own, the contents are full of essential stories, points, and keywords in spite of its compact handbook size. We believe this review could provide highly sophisticated, informative ideas matured in JT-60 technological developments. (author)

Avoidance of VDEs during plasma current quench in JT-60U

International Nuclear Information System (INIS)

Yoshino, R.; Nakamura, Y.; Neyatani, Y.

1996-01-01

Vertical displacement events (VDEs) during plasma current quench (I p quench) are one of the serious problems encountered in designing tokamak fusion reactors, owing to the generation of enormously high electromagnetic forces on the vacuum vessel and in-vessel components, but they have been passively and actively avoided in JT-60U. In JT-60U 'slow I p quench' is ended with very fast plasma current termination (final I p termination), and the halo current is frequently measured at this final I p termination. VDEs make the final I p termination severe by increasing the halo current and the electromagnetic force. A strong dependence of VDE growth rate on the initial vertical position of the plasma current centre (Z J ) has been clarified experimentally, and a neutral point of Z J for VDE has been found at ∼ 15 cm above the midplane of the vacuum vessel. According to these measurements, VDE has been avoided by the selection of Z J at the start of I p quench (passive control) and by the control of Z J during I p quench (active control) eventually obtained owing to the small deviation of Z J in real time calculations from its actual value. Furthermore, passive avoidance of VDEs by the injection of a neon ice pellet has been demonstrated. (author). 29 refs, 14 figs

Edge pedestal characteristics in JET and JT-60U tokamaks under variable toroidal field ripple

NARCIS (Netherlands)

Urano, H.; Saibene, G.; Oyama, N.; Parail, V.; P. de Vries,; Sartori, R.; Kamada, Y.; Kamiya, K.; Loarte, A.; Lonnroth, J.; Sakamoto, Y.; Salmi, A.; Shinohara, K.; Takenaga, H.; Yoshida, M.

2011-01-01

The effects of toroidal field (TF) ripple on the edge pedestal characteristics were examined in the TF ripple scan experiments at the plasma current I(p) of 1.1 MA in JET and JT-60U. The TF ripple amplitude delta(R) was defined as a value averaged over the existing ripple wells at the separatrix on

Operation and Development on the Positive-Ion Based Neutral Beam Injection System for JT-60 and JT-60U

International Nuclear Information System (INIS)

Kuriyama, M.; Akino, N.; Ebisawa, N.; Honda, A.; Itoh, T.; Kawai, M.; Mogaki, K.; Ohga, T.; Oohara, H.; Umeda, N.; Usui, K.; Yamamoto, M.; Yamamoto, T.; Matsuoka, M.

2002-01-01

The positive-ion based neutral beam injection (NBI) system for JT-60, which consists of 14 beamline units and has a beam energy of 70 to 100 keV, started operation in 1986 with hydrogen beams and injected a neutral beam power of 27 MW at 75 keV into the JT-60 plasma. In 1991, the NBI system was modified to be able to handle deuterium beams as part of the JT-60 upgrade modification. After executing some research and developments, deuterium beams of 40 MW at 95 keV were injected in 1996. As a result, NBI has contributed to the achievement of the highest performance plasmas, a DT-equivalent fusion power gain of 1.25 and a fusion triple product of 1.55 x 10 21 keVs/m 3 , in the world on JT-60U

Design study of a time-of-flight neutron spectrometer for JT-60U

International Nuclear Information System (INIS)

Elevant, T.; Hoek, M.; Nishitani, Takeo.

1993-06-01

A time-of-flight neutron spectrometer is proposed for measurements of neutron energy spectra from deuterium-deuterium reactions in JT-60U tokamak plasmas. The sensitivity of the instrument is 2 · 10 -2 cm 2 , energy resolution is 4.5 % (FWHM) and maximum useful count-rate is 6 kHz. Analysis of neutron energy spectra will provide information on central ion temperatures larger than ∼ 4 keV with an accuracy of ± 10 %, and neutron source fraction from reactions between thermal ions with an accuracy of ± 15 %. The minimum time required for data acquisition is 0.1 s. (author)

FIR-laser scattering for JT-60

International Nuclear Information System (INIS)

Itagaki, Tokiyoshi; Matoba, Tohru; Funahashi, Akimasa; Suzuki, Yasuo

1977-09-01

An ion Thomson scattering method with far infrared (FIR) laser has been studied for measuring the ion temperature in large tokamak JT-60 to be completed in 1981. Ion Thomson scattering has the advantage of measuring spatial variation of the ion temperature. The ion Thomson scattering in medium tokamak (PLT) and future large tokamak (JET) requires a FIR laser of several megawatts. Research and development of FIR high power pulse lasers with power up to 0.6 MW have proceeded in ion Thomson scattering for future high-temperature tokamaks. The FIR laser power will reach to the desired several megawatts in a few years, so JAERI plans to measure the ion temperature in JT-60 by ion Thomson scattering. A noise source of the ion Thomson scattering with 496 μm-CH 3 F laser is synchrotron radiation of which the power is similar to NEP of the Schottky-barrier diode. However, the synchrotron radiation power is one order smaller than that when a FIR laser is 385 μm-D 2 O laser. The FIR laser power corresponding to a signal to noise ratio of 1 is about 4 MW for CH 3 F laser, and 0.4 MW for D 2 O laser if NEP of the heterodyne mixer is one order less. A FIR laser scattering system for JT-60 should be realized with improvement of FIR laser power, NEP of heterodyne mixer and reduction of synchrotron radiation. (auth.)

Mechanical properties of JT-60 tokamak machine in power tests

International Nuclear Information System (INIS)

Takatsu, Hideyuki; Ohkubo, Minoru; Yamamoto, Masahiro; Ohta, Mitsuru

1986-01-01

JT-60 power tests were carried out from Dec. 10, 1984 to Feb. 20, 1985 to demonstrate, in advance of actual plasma operation, satisfactory performance of tokamak machine, power suppliers and control system in combination. The tests began with low power test of individual coil systems and progressed to full power tests. The coil current was raised step by step, monitoring the mechanical, thermal, electrical and vacuum data. Power tests were concluded with successful results. All of the coil systems were raised up to full power operation in combination and system performance was verified including the structural integrity of tokamak machine. Measured strain and deflection showed good agreements with those predicted in the design, which was an evidence that electromagnetic forces were supported as expected in the design. A few limitations to machine operation was made clear quantitatively. And it was found that existing detectors were insufficient to monitor machine integrity and two kinds of detector were proposed to be installed. (author)

Numerical simulation on current spike behaviour of JT-60U disruptive plasmas

International Nuclear Information System (INIS)

Takei, N; Nakamura, Y; Tsutsui, H; Yoshino, R; Kawano, Y; Ozeki, T; Tobita, K; Tsuji-Iio, S; Shimada, R; Jardin, S C

2004-01-01

Characteristics and underlying mechanisms for plasma current spikes, which have been frequently observed during the thermal quench of JT-60U disruptions, were investigated through tokamak simulation code simulations including the passive shell effects of the vacuum vessel. Positive shear and reversed shear (PS and RS) plasmas were shown to have various current spike features in the experiments, e.g. an impulsive increase in the plasma current (positive spike) in the majority of thermal quenches, and a sudden decrease (negative spike), that has been excluded from past consideration, as an exception. It was first clarified that the shell effects, which become significant especially at a strong pressure drop due to the thermal quench of high β p plasmas, play an important role in the current spike in accordance with the initial relation of the radial location between the plasma equilibria and the vacuum vessel. As a consequence, a negative current spike may appear at thermal quench when the plasma is positioned further out from the geometric centre of the vacuum vessel. It was also pointed out that a further lowering in the internal inductance, in contradiction to previous interpretation in the past, is a plausible candidate for the mechanism for positive current spikes observed even in RS plasmas. The new interpretation enables us to reason out the whole character of current spikes of JT-60U disruptions

2. Interferometry and polarimetry. 2.3. Polarimetry on JT-60U

International Nuclear Information System (INIS)

Kawano, Yasunori

2000-01-01

In order to establish an electron density measurement method with high reliability and stability for magnetic-confinement fusion devices, studies on infrared polarimetry have been carried out in JT-60U. Electron density measurement based on tangential Faraday rotation has been verified using a CO 2 laser polarimeter developed for JT-60U. In this article, basic ideas of studies, results from polarimetry experiments, and suggestions for future devices are presented. (author)

Recent results on steady state and confinement improvement research on JT-60U

International Nuclear Information System (INIS)

Ide, Shunsuke

2000-01-01

On the JT-60U tokamak, fusion plasma research for realization of a steady state tokamak reactor has been pursued. Towards that goal, confinement improved plasmas such as H-mode, high β p , reversed magnetic shear (RS) and latter two combined with H-mode edge pedestal have been developed and investigated intensively. A key issue to achieve non-inductive current drive relevant to a steady state fusion reactor is to increase the fraction of the bootstrap current and match the spatial profile to the optimum. In 1999, as the result of the optimization, the equivalent deuterium-tritium (D-T) fusion gain (Q DT eq ) of 0.5 was sustained for 0.8 s, which is roughly equal to the energy confinement time, in a RS plasma. In order to achieve a RS plasma in steady state two approach have been explored. One is to use external current driver such as lower hybrid current drive (LHCD), and by optimizing LHCD a quasi-steady RS discharge was obtained. The other approach is to utilize bootstrap current as much as possible, and with highly increased fraction of the bootstrap current, a confinement enhancement factor of 3.6 was maintained for 2.7 s in a RS plasma with H-mode edge. A heating and current drive system in the electron cyclotron range of frequency for localized heating and current drive has been installed on JT-60U, and in initial experiments a clear increase of the central electron temperature in a RS high density central region was confirmed only with injected power of 0.75 MW. (author)

Absolute calibration of the neutron yield measurement on JT-60 Upgrade

International Nuclear Information System (INIS)

Nishitani, Takeo; Takeuchi, Hiroshi; Barnes, C.W.

1991-10-01

Absolutely calibrated measurements of the neutron yield are important for the evaluation of the plasma performance such as the fusion gain Q in DD operating tokamaks. Total neutron yield is measured with 235 U and 238 U fission chambers and 3 He proportional counters in JT-60 Upgrade. The in situ calibration was performed by moving the 252 Cf neutron source toroidally through the JT-60 vacuum vessel. Detection efficiencies of three 235 U and two 3 He detectors were measured for 92 locations of the neutron point source in toroidal scans at two different major radii. The total detection efficiency for the torus neutron source was obtained by averaging the point efficiencies over the whole toroidal angle. The uncertainty of the resulting absolute plasma neutron source calibration is estimated to be ± 10%. (author)

User's manual of JT-60 experimental data analysis system

International Nuclear Information System (INIS)

Hirayama, Takashi; Morishima, Soichi; Yoshioka, Yuji

2010-02-01

In the Japan Atomic Energy Agency Naka Fusion Institute, a lot of experiments have been conducted by using the large tokamak device JT-60 aiming to realize fusion power plant. In order to optimize the JT-60 experiment and to investigate complex characteristics of plasma, JT-60 experimental data analysis system was developed and used for collecting, referring and analyzing the JT-60 experimental data. Main components of the system are a data analysis server and a database server for the analyses and accumulation of the experimental data respectively. Other peripheral devices of the system are magnetic disk units, NAS (Network Attached Storage) device, and a backup tape drive. This is a user's manual of the JT-60 experimental data analysis system. (author)

JT-60SA power supply system

International Nuclear Information System (INIS)

Coletti, A.; Baulaigue, O.; Cara, P.; Coletti, R.; Ferro, A.; Gaio, E.; Matsukawa, M.; Novello, L.; Santinelli, M.; Shimada, K.; Starace, F.; Terakado, T.; Yamauchi, K.

2011-01-01

The paper describes the main features of the Superconducting Magnets Power Supply to generate the toroidal and poloidal magnetic fields in JT-60SA tokamak, with special regard to coil current regulation mode and magnets protection.

Comparison of particle confinement in the high confinement mode plasmas with the edge localized mode of the Japan Atomic Energy Research Institute Tokamak-60 Upgrade and the DIII-D tokamak

International Nuclear Information System (INIS)

Takenaga, H.; Mahdavi, M.A.; Baker, D.R.

2001-01-01

Particle confinement was compared for the high confinement mode plasmas with the edge localized mode in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) [S. Ishida, JT-60 Team, Nucl. Fusion 39, 1211 (1999)] and the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] considering separate confinement times for particles supplied by neutral beam injection (NBI) (center fueling) and by recycling and gas-puffing (edge fueling). Similar dependence on the NBI power was obtained in JT-60U and DIII-D. The particle confinement time for center fueling in DIII-D was smaller by a factor of 4 in the low density discharges and by a factor of 1.8 in the high density discharges than JT-60U scaling, respectively, suggesting the stronger dependence on the density in DIII-D. The particle confinement time for edge fueling in DIII-D was comparable with JT-60U scaling in the low density discharges. However, it decreased to a much smaller value in the high density discharges

Review of JT-60U experimental results from March to October, 1991

International Nuclear Information System (INIS)

1992-06-01

Experimental results achieved in the initial operation of JT-60U are described in this paper. Experiments of JT-60U were initiated in March 1991, and deuterium experiments started in the middle of July. Multivariable non-interacting control, H-mode experiments, the high-q and high-β p regime with hot ion enhanced confinement, the divertor heat flux, etc. are reported. Achieved values of the first experiment of the JT-60U LHCD in 1991 were P LH = 1.5 MW, driven current I RF = 2MA, current drive efficiency η CD (=n-bar e R p I RF /P LH ) = 0.25 x 10 20 m -2 A/W and current driven product CDP(=n-bar e R p I RF ) = 3 x 10 20 m -2 MA. (J.P.N.)

Energetic particle physics in JT-60U and JFT-2M

Energy Technology Data Exchange (ETDEWEB)

Shinohara, K [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Takechi, M [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Ishikawa, M [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Kusama, Y [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Tsuzuki, K [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Urata, K [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Kawashima, H [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Tobita, K [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Fukuyama, A [Department of Nuclear Engineering, Kyoto University, 606-8501, (Japan); Cheng, C Z [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Darrow, D S [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Kramer, G J [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Gorelenkov, N N [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Nazikian, R [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543 (United States); Todo, Y [National Institute for Fusion Science, Oroshi-cho, Toki, Gifu, 509-5292, (Japan); Miura, Y [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan); Ozeki, T [Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki, 311-0193 (Japan)

2004-07-01

Recent energetic particle physics research in JT-60U and JFT-2M is reported. Alfven eigenmodes (AEs) are investigated in reversed-shear (RS) plasmas in JT-60U where frequency sweeping (FS) modes are observed to follow the q-profile evolution. The RS-induced AE model can explain the FS of the modes within the context of an evolving q-profile. Enhanced energetic ion transport is also investigated with the appearance of modes in the toroidicity-induced AE range of frequency in JT-60U using a multi-channel neutron profile monitor and in JFT-2M using a lost ion probe. Additionally, the ripple loss in the complex toroidal field ripple due to ferritic steel inserts in JFT-2M is shown and compared with model analysis. The simulation code developed to predict ripple loss in JFT-2M will be of use in estimating the heat flux in the complex ripple field of a future device such as ITER.

Conceptual design of JT-60SA cryostat

International Nuclear Information System (INIS)

Shibama, Y.K.; Sakurai, S.; Masaki, K.; Sukekawa, A.M.; Kaminaga, A.; Yoshida, K.; Matsukawa, M.

2007-01-01

JT-60U modification program to fully superconducting device has been proceeded, namely ''JT-60SA'', toward early realization of fusion energy based on tokamak concept. The design of JT-60SA cryostat is expected to achieve a vacuum thermal insulation for super conducting coils, a bio-shielding boundary and structural gravity support. The cryostat is required to cover JT-60SA tokamak device, which is 15 m of total height and 7 m of radius, but there is geometrical limit due to surrounding devices reutilized. Although the cryostat consists of vessel body and gravity support, and the structural material is low cobalt 304 stainless steel (Co: 2 , and the design of the leaf spring is considered to reduce thermal stress, and to withstand the mechanical loads of plasma disruption and seismic loads. The coolant is 80 K gas helium and both sides of panel are covered with multi-layers super insulation (SI) to reduce heat load (radiation) up to 1/100. Fraction of non-covered region is assumed to be 2% due to many port-joints and supports for the vacuum vessel. Total heat load for inner surface of cryostat (600 m 2 ) is 9kW and the heat load for the port-joints (-300 m 2 ) is assumed up to 9 kW. The operational pressure of the cryostat is required to keep less than 10 -2 Pa and about 100,000 m 2 of structural surfaces is considered for exhaust system specification. Another role of the cryostat is the radiation protection. Biological shielding up to 10 micro-Sv/h (for maintenance acceptance) is required of the cryostat surface after the 10 years operation. Thus the cryostat consists of boron (2 wt%) doped concrete of 220 mm thickness and structural SS304 of total 40 mm thickness. The concrete reduces the air activation (41Ar) in the torus hall by 90% rather than the normal one by the thermal neutron absorption of boron. (orig.)

A four-pellet pneumatic injection system in the JT-60

International Nuclear Information System (INIS)

Hiratsuka, Hajime; Kawasaki, Kouzo; Miyo, Yasuhiko; Yoshioka, Yuji; Ohta, Kazuya; Shimizu, Masatsugu; Kondo, Ikuo; Onozuka, Masanori; Shimomura, Tomoyoshi; Iwamoto, Syuichi; Hashiri, Noboru

1991-01-01

A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter x 3.0 mm long for two pellets and 4.0 mm in diameter x 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector. (orig.)

A four-pellet pneumatic injection system in the JT-60

Energy Technology Data Exchange (ETDEWEB)

Hiratsuka, Hajime; Kawasaki, Kouzo; Miyo, Yasuhiko; Yoshioka, Yuji; Ohta, Kazuya; Shimizu, Masatsugu; Kondo, Ikuo (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan)); Onozuka, Masanori; Shimomura, Tomoyoshi; Iwamoto, Syuichi; Hashiri, Noboru (Mitsubishi Heavy Industries Ltd., Kobe (Japan))

1991-05-01

A four-pellet pneumatic injection system has been developed for plasma fueling of the JT-60. The JT-60 pellet injector is capable of accelerating separately four cylindrical pellets 3.0 mm in diameter x 3.0 mm long for two pellets and 4.0 mm in diameter x 4.0 mm long for the remaining two. The JT-60 pellet injector was installed on the JT-60 tokamak machine at the end of 1988. Obtained pellet velocity was higher than 2300 m/s by propellant gases of up to 100 bar and the pellet fueling efficiency achieved was around 70% for both dimensions of pellets. This paper describes the design, injection operation and performance test results of the JT-60 pellet injector. (orig.).

Pneumatic pellet injector for JT-60

International Nuclear Information System (INIS)

Onozuka, Masanori; Hiratsuka, Hajime; Kawasaki, Kouzo.

1990-01-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas. (author)

Pneumatic pellet injector for JT-60

Energy Technology Data Exchange (ETDEWEB)

Onozuka, Masanori (Mitsubishi Heavy Industries Ltd., Tokyo (Japan)); Hiratsuka, Hajime; Kawasaki, Kouzo

1990-11-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proven that the device provides high speed pellets as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 2.3km/s at 100 bar propellant gas. (author).

Impurity and particle recycling reduction by boronization in JT-60U

International Nuclear Information System (INIS)

Higashijima, S.; Sugie, T.; Kubo, H.; Tsuji, S.; Shimada, M.; Asakura, N.; Hosogane, N.; Kawano, Y.; Nakamura, H.; Itami, K.; Sakasai, A.; Shimizu, K.; Ando, T.; Saidoh, M.

1995-01-01

In JT-60U boronization using decaborane was carried out. Boronization reduced the oxygen impurity in OH discharges and shortened the wall conditioning after the vacuum vessel vent and consequently enabled JT-60U to produce clean plasmas easily except for NB heated plasmas. After boronization, particle recycling was reduced drastically in OH and NB discharges. High confinement plasmas were obtained including high β p mode and H-mode discharges. In the latest boronization part of divertor plates were replaced with B 4 C coated tiles with a B 4 C thickness similar 300 μm. After introducing B 4 C divertor tiles, an explosive generation of boron particles from the tiles was observed. By the combined effects of boronization with decaborane and boron generation from B 4 C tiles, oxygen impurity was so low that oxygen line signals were reduced to noise levels after the latest boronization. On the other hand, boron burst from the B 4 C tiles restricted the operation of JT-60U. ((orig.))

Development of the pellet injector for JT-60

International Nuclear Information System (INIS)

Kawasaki, Kouzo; Hiratsuka, Hajimo; Takatsu, Hideyuki; Shimizu, Masatsugu; Onozuka, Masanori; Uchikawa, Takashi; Iwamoto, Syuichi; Hashiri, Nobuo

1989-01-01

The pneumatic 4-shot pellet injector has been installed and operated for JT-60 (JAERI Tokamak-60). The performance tests have proved that the device provides high speed hydrogen pellets just as planned. The maximum pellet velocity obtained in the hydrogen pellet tests is greater than 1.6 km/sec at 50 bar propellant gas. The device is now in use for JT-60 contributing to plasma study. In this paper the outline of features and performance of the device is presented. (author). 4 refs.; 8 figs

The archives of operational achievements in JT-60

International Nuclear Information System (INIS)

Seimiya, Munetaka

2007-08-01

Since the first plasma in JT-60 was achieved in April 1985, various experimental challenges have been successfully conducted, and currently producing many new findings. These achievements have been realized by large modifications for lower X-point divertor in 1987, for large plasma current upgrade in 1989-1991, for W-shaped divertor in 1997, and for long pulse discharge in 2002. Such developments contribute to have established JT-60 as the leading tokamak in the world. As a consequence of the 22-year operation, we have accumulated many operational and experimental data. This reports the operational records including troubles and availability, the outline of planning management, the safety control and the promotion procedure of operation in JT-60. (author)

Overview of JT-60U results toward high integrated performance in reactor-relevant regime

International Nuclear Information System (INIS)

Fujita, T.

2002-01-01

Toward steady sustainment of high integrated performance, we have developed weak magnetic shear (high β p mode) and reversed magnetic shear plasmas. As a large-sized tokamak equipped with a variety of devices for heating, current drive and profile/shape control, JT-60U has high ability to approach the conditions required in reactors: low values of normalized Larmor radius and collisionality, high temperatures with T e > or approx. T i , etc. This paper reviews recent JT-60U results with the emphasis on the projection to the reactor-relevant regime. Full non-inductive current drive has been achieved in a 1.8 MA high β p H-mode plasma with β N 2:4, HH y2 =1.2 and high fusion triple product (3 x 10 20 m -3 keVs) owing to increased N-NB power. In a reversed shear plasma, HH y2 =1.4 at n e /n GW 0.8 under the full non-inductive current drive has been achieved with injection of LHRF and N-NB. In box-type ITBs with reversed shear, barriers for ions and electrons were sustained in a regime with T e > or approx. T i . The pedestal pressure was doubled with increased total poloidal beta in pellet-injected high triangularity plasmas with type I and II ELMs. Stable existence of current hole was demonstrated. (author)

Stable existence of central current hole in the JT-60U tokamak

International Nuclear Information System (INIS)

Miura, Y.; Fujita, T.; Oikawa, T.

2003-01-01

In an extreme state of a reversed magnetic shear configuration, it was found in JT-60U that there is almost no plasma current in the central region (called Current Hole). The Current Hole region extends to 40% of the plasma minor radius and it exists stably for several seconds. The Current Hole is formed by the growth of the bootstrap current and it is impossible to drive current in either positive or negative direction by ECH or N-NB inside the Current Hole. In that region, there is almost no gradient of density, temperature and toroidal rotation velocity. It means that there is almost no confinement in the Current Hole and the large energy in that region is sustained only by an internal transport barrier (ITB). The effects of the Current Hole on particle orbits and the effects on an error field on the Current Hole are also discussed. (author)

Construction of negative-ion based NBI for JT-60U

International Nuclear Information System (INIS)

Kawai, Mikito; Akino, Noboru; Ebisawa, Noboru

2001-11-01

The world's first negative-ion based neutral beam injector (N-NBI) system has been developed for studies of non-inductive current drive and plasma core heating with high energy neutral beam injection in higher density plasma. Construction of the N-NBI system for JT-60U was completed in March 1996. The system is composed of a beamline with two ion sources, a set of ion source power supplies, control system and auxiliary sub-system such as cooling water, refrigeration and vacuum system. In July 2001, deuterium neutral beam injection of 400keV and 5.8MW into JT-60U plasma was achieved. In order to increase both beam power and energy we have to go on more improvement of the N-NBI. (author)

Operation and management manual of JT-60 experimental data analysis system

International Nuclear Information System (INIS)

Hirayama, Takashi; Morishima, Soichi

2014-03-01

In the Japan Atomic Energy Agency Naka Fusion Institute, a lot of experiments have been conducted by using the large tokamak device JT-60 aiming to realize fusion power plant. In order to optimize the JT-60 experiment and to investigate complex characteristics of plasma, JT-60 experimental data analysis system was developed and used for collecting, referring and analyzing the JT-60 experimental data. Main components of the system are a data analysis server and a database server for the analyses and accumulation of the experimental data respectively. Other peripheral devices of the system are magnetic disk units, NAS (Network Attached Storage) device, and a backup tape drive. This is an operation and management manual the JT-60 experimental data analysis system. (author)

Neural-net disruption predictor in JT-60U

International Nuclear Information System (INIS)

Yoshino, R.

2003-01-01

The prediction of major disruptions caused by the density limit, the plasma current ramp-down with high internal inductance l i , the low density locked mode and the β-limit has been investigated in JT-60U. The concept of 'stability level', newly proposed in this paper to predict the occurrence of a major disruption, is calculated from nine input parameters every 2 ms by the neural network and the start of a major disruption is predicted when the stability level decreases to a certain level, the 'alarm level'. The neural network is trained in two steps. It is first trained with 12 disruptive and six non-disruptive shots (total of 8011 data points). Second, the target output data for 12 disruptive shots are modified and the network is trained again with additional data points generated by the operator. The 'neural-net disruption predictor' obtained has been tested for 300 disruptive shots (128 945 data points) and 1008 non-disruptive shots (982 800 data points) selected from nine years of operation (1991-1999) of JT-60U. Major disruptions except for those caused by the -limit have been predicted with a prediction success rate of 97-98% at 10 ms prior to the disruption and higher than 90% at 30 ms prior to the disruption while the false alarm rate is 2.1% for non-disruptive shots. This prediction performance has been confirmed for 120 disruptive shots (56 163 data points), caused by the density limit, as well as 1032 non-disruptive shots (1004 611 data points) in the last four years of operation (1999-2002) of JT-60U. A careful selection of the input parameters supplied to the network and the newly developed two-step training of the network have reduced the false alarm rate resulting in a considerable improvement of the prediction success rate. (author)

Design study of a new P-NBI control system for 100-s injection in JT-60SA

International Nuclear Information System (INIS)

Honda, Atsushi; Okano, Fuminori; Shinozaki, Shinichi; Ooshima, Katsumi; Ikeda, Yoshitaka; Numazawa, Susumu

2007-03-01

The modification of the JT-60U to a fully superconducting coil tokamak, JT-60SA (Super Advanced), has been programmed as the satellite devise for the ITER (International Thermonuclear Experimental Reactor) and as the national centralized tokamak. The present positive-ion-based NBI system (P-NBI), which has been operated for 20 years and will be the main heating system on JT-60SA, is required to manage the long pulse injection extended from 30 s to 100 s at the power of 24 MW with 12 units. To realize such a requirement, the original control system handling more than 4000 digital data is to be fully remodeled. Design study of the new control system has been conducted from viewpoint of market availability, system extensibility, cost-effectiveness and independent development in programming. It has been concluded that a distributed control system using PLC (Programmable Logic Controller) could be applied to the large-scale control system for 100-s operations with satisfaction of the evaluation viewpoints. (author)

Study of neutral particle behavior and particle confinement in JT-60U

International Nuclear Information System (INIS)

Takenaga, Hidenobu; Shimizu, Katsuhiro; Asakura, Nobuyuki; Shimada, Michiya; Kikuchi, Mitsuru; Tsuji-Iio, Shunji; Uchino, Kiichiro; Muraoka, Katsunori.

1995-07-01

In order to understand the particle confinement properties in JT-60U, the particle confinement time was estimated through analyses of the neutral particle behavior. First, the neutral particle transport simulation code DEGAS using a Monte-Carlo technique was combined with the simple divertor code for calculating the edge plasma parameters, and was developed to calculate under the experimental conditions in JT-60U. Then, the charged particle source in the main plasma due to the ionization of the neutral particles was evaluated from the analyses of the neutral particle penetration to the main plasma based on results of the simulation code and measurements of D α emission intensities. Finally, the particle confinement time was estimated from the analysis of particle balance. The analyses were performed systematically for the L-mode plasma and H-mode plasma of JT-60U, and a data base of the particle confinement time was obtained. The dependence of the particle confinement time on the plasma parameters and the relationship between the properties of the particle confinement and the energy confinement were examined. (author)

Development of fast charge exchange recombination spectroscopy by using interference filter method in JT-60U

International Nuclear Information System (INIS)

Kobayashi, Shinji; Sakasai, Akira; Koide, Yoshihiko; Sakamoto, Yoshiteru; Kamada, Yutaka; Hatae, Takaki; Oyama, Naoyuki; Miura, Yukitoshi

2003-01-01

Recent developments and results of fast charge exchange recombination spectroscopy (CXRS) using interference filter method are reported. In order to measure the rapid change of the ion temperature and rotation velocity under collapse or transition phenomena with high-time resolution, two types of interference filter systems were applied to the CXRS diagnostics on the JT-60U Tokamak. One can determine the Doppler broadening and Doppler shift of the CXR emission using three interference filters having slightly different center wavelengths. A rapid estimation method of the temperature ad rotation velocity without non-linear least square fitting is presented. The modification of the three-filters system enables us to improve the minimum time resolution up to 0.8 ms, which is better than that of 16.7 ms for the conventional CXRS system using the CCD detector in JT-60U. The other system having seven wavelength channels is newly fabricated to crosscheck the results obtained by the three-filters assembly, that is, to verify that the CXR emission forms a Gaussian profile under collapse phenomena. In a H-mode discharge having giant edge localized modes, the results obtained by the two systems are compared. The applicability of the three-filters system to the measurement of rapid changes in temperature and rotation velocity is demonstrated. (author)

Study of carbon impurity generation by chemical sputtering in JT-60U

International Nuclear Information System (INIS)

Higashijima, S.; Kubo, H.; Sugie, T.; Shimizu, K.; Asakura, N.; Itami, K.; Hosogane, N.; Sakasai, A.; Konoshima, S.; Sakurai, S.; Takenaga, H.

1997-01-01

CD/CH-band intensities emitted from hydrocarbon molecules have been measured in the divertor region of JT-60U and the chemical sputtering yield of methane was estimated as a function of the surface temperature and the deuterium ion flux. The chemical sputtering yield increases with the surface temperature and decreases with increasing ion flux density in the L-mode plasmas. The B 4 C converted CFC tiles are installed in JT-60U and it is found that the chemical sputtering of B 4 C converted CFC tiles is suppressed in comparison to normal CFC tiles. (orig.)

Conceptual design of JT-60SA cryostat

International Nuclear Information System (INIS)

Shibama, Y.K.; Sakurai, S.; Masaki, K.; Sukekawa, A.M.; Kaminaga, A.; Sakasai, A.; Matsukawa, M.

2008-01-01

This paper describes the conceptual design of cryostat for the JT-60SA, which is a research device for the commercial production of electricity from the controlled fusion reaction in the future. JT-60SA is designed to be a fully superconducting device and cryostat is one of the main components to allow the normal operation. Cryostat covers up the tokamak device, which is 15 m of total height and 7 m of radius, and supports the total weight of 25 MN. Cryostat components consist of vessel body, gravity support and auxiliary systems, such as 80 K thermal shield and vacuum exhaust. The functions required of cryostat are these three, thermal insulation for superconducting magnets, gravity support for the tokamak device, and bio-shielding. The design conditions for each cryostat component are outlined and the features of auxiliary systems such as capacity of vacuum exhaust related to 80 K thermal shield design are summarized

Development of magnetic sensors for JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Takechi, M., E-mail: takechi.manabu@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Matsunaga, G.; Sakurai, S.; Sasajima, T.; Yagyu, J.; Hoshi, R.; Kawamata, Y.; Kurihara, K. [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Nishikawa, T.; Ryo, T.; Kagamihara, S. [Okazaki Manufacturing Company, Kobe, Hyogo 651-0087 (Japan); Nakamura, K. [RIAM, Kyushu Univ., Kasuga, Fukuoka 816-8580,Japan (Japan)

2015-10-15

JT-60SA has been designed and is being constructed to demonstrate and develop steady-state high-beta operation. Resistive wall mode (RWM) control, error field correction, and edge-localized mode (ELM) control will be performed using in-vessel coils. For these controls, we have developed a biaxial magnetic sensor to determine 3D magnetic configuration of the plasma. Moreover, for obtaining basic information about JT-60SA plasma, magnetic sensors, in particular, one-turn loops, Rogowski coils, diamagnetic loops, and saddle coils have been developed. Because the length of the vacuum vessel in the poloidal direction of JT-60SA is 16 m and almost twice as long as that of JT-60U, the length of the Rogowski coil and the diamagnetic loop of JT-60SA are also twice as long as those on JT-60U. We have devised new types of sensors and a connector for the mineral-insulated cable because construction and installation of these sensors are much more difficult in JT-60SA. We will report the design and specification of the magnetic sensors for JT-60SA from the physics and engineering aspects.

Overview of engineering design, manufacturing and assembly of JT-60SA machine

Energy Technology Data Exchange (ETDEWEB)

Di Pietro, Enrico, E-mail: enrico.dipietro@jt60sa.org [JT-60SA EU Home Team, Fusion for Energy, Boltzmannstrasse 2, Garching 85748 (Germany); Barabaschi, Pietro [JT-60SA EU Home Team, Fusion for Energy, Boltzmannstrasse 2, Garching 85748 (Germany); Kamada, Yutaka [JT-60SA JA Home Team, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Ishida, Shinichi [JT-60SA JA Project Team, Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2014-10-15

The JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The JT-60SA device is a fully superconducting tokamak capable of confining break-even equivalent deuterium plasmas with equilibria covering high plasma shaping with a low aspect ratio at a maximum plasma current of I{sub p} = 5.5 MA. This makes JT-60SA capable to support and complement ITER in all the major areas of fusion plasma development necessary to decide DEMO reactor construction. After a complex start-up phase due to the necessity to carry out a re-baselining effort with the purpose to fit in the original budget while aiming to retain the machine mission, performance, and experimental flexibility, in 2009 detailed design could start. With the majority of time-critical industrial contracts in place, in 2012, it was possible to establish a credible time plan, and now, the project is progressing on schedule towards the first plasma in March 2019. After careful and focused R and D and qualification tests, the procurement of the major components and plant is now well advanced in manufacturing design and/or fabrication. In the meantime the disassembly of the JT-60U machine has been completed and the engineering of the JT-60SA assembly process has been developed. The actual assembly of JT-60SA started in January 2013 with the installation of the cryostat base. The paper gives an overview of the present status of the engineering design, manufacturing and assembly of the JT-60SA machine.

Review of JT-60U experimental results from February to October, 1995

International Nuclear Information System (INIS)

1996-03-01

Renewed theme group organization started from October 1994 for the upcoming experiments in JT-60U. This regime has three theme groups each of which is composed of two sub-theme groups as; (1) Plasma Operation Theme (Leader Y. Neyatani) with Operation Sub-Theme and Disruption Sub-Theme, (2) High Performance (Leader S. Ishida) with Confinement and MHD Sub-Theme and High Energy Particle Sub-Theme and (3) Steady State Theme (Leader A. Sakasai) with Current Drive Sub-Theme and Divertor Sub-Theme. The main results from the JT-60U experiments in 1995 are summarized in the overviews of the three theme group activities. (J.P.N.)

Review of JT-60U experimental results from February to October, 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-03-01

Renewed theme group organization started from October 1994 for the upcoming experiments in JT-60U. This regime has three theme groups each of which is composed of two sub-theme groups as; (1) Plasma Operation Theme (Leader Y. Neyatani) with Operation Sub-Theme and Disruption Sub-Theme, (2) High Performance (Leader S. Ishida) with Confinement and MHD Sub-Theme and High Energy Particle Sub-Theme and (3) Steady State Theme (Leader A. Sakasai) with Current Drive Sub-Theme and Divertor Sub-Theme. The main results from the JT-60U experiments in 1995 are summarized in the overviews of the three theme group activities. (J.P.N.).

High performance experiments in JT-60U reversed shear discharges

International Nuclear Information System (INIS)

Fujita, T.; Kamada, Y.; Ishida, S.

2001-01-01

The operation of JT-60U reversed shear discharges has been extended to a high plasma current, low-q regime keeping a large radius of the internal transport barrier (ITB) and the record value of equivalent fusion multiplication factor in JT-60U, Q DT eq =1.25, has been achieved at 2.6 MA. Operational schemes to reach the low-q regime with good reproducibility have been developed. The reduction of Z eff was obtained in the newly installed W-shaped pumped divertor. The beta limit in the low-q min regime, which limited the performance of L-mode edge discharges, has been improved in H-mode edge discharges with a broader pressure profile, which was obtained by power flow control with ITB degradation. Sustainment of ITB and improved confinement for 5.5 seconds has been demonstrated in an ELMy H reversed shear discharge. (author)

Injection control development of the JT-60U electron cyclotron heating system

Energy Technology Data Exchange (ETDEWEB)

Hiranai, Shinichi; Shinozaki, Shin-ichi; Yokokura, Kenji; Moriyama, Shinichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sato, Fumiaki [Nippon Advanced Technology Co., Ltd., Tokai, Ibaraki (Japan); Suzuki, Yasuo [Atomic Energy General Service Co., Ltd., Tokai, Ibaraki (Japan); Ikeda, Yoshitaka [Japan Atomic Energy Research Inst., Kashiwa, Chiba (Japan)

2003-03-01

The JT-60U electron cyclotron heating (ECH) System injects a millimeteric wave at 110 GHz into the JT-60 Plasma, and heats the plasma or drives a current locally to enhance the confinement performance of the JT-60 plasma. The system consists of four sets of high power gyrotrons, high voltage power supplies and transmission lines, and two antennas that launch electron cyclotron (EC) beams toward the plasma. The key features of the injection control system are streering of the direction of the EC beam by driving the movable mirror in the antenna, and capability to set any combination of polarization angle and ellipticity by rotating the two grooved mirrors in the polarizers. This report represents the design, fabrication and improvements of the injection control system. (author)

Study of grounding system of large tokamak device JT-60

International Nuclear Information System (INIS)

Arakawa, Kiyotsugu; Shimada, Ryuichi; Kishimoto, Hiroshi; Yabuno, Kohei; Ishigaki, Yukio.

1982-01-01

In the critical plasma testing facility JT-60 constructed by the Japan Atomic Energy Research Institute, high voltage, large current is required in an instant. Accordingly, for the protection of human bodies and the equipment, and for realizing the stable operation of the complex, precise control and measurement system, a large scale facility of grounding system is required. In case of the JT-60 experimental facility, the equipments with different functions in separate buildings are connected, therefore, it is an important point to avoid high potential difference between buildings. In the grounding system for the JT-60, a reticulate grounding electrode is laid for each building, and these electrodes are connected with a low impedance metallic duct called grounding trunk line. The power supply cables for various magnetic field coils, control lines and measurement lines are laid in the duct. It is a large problem to grasp quantitatively the effect of a grounding trunk line by analysis. The authors analyzed the phenomenon that large current flows into a grounding system by lightning strike or grounding. The fundamental construction of the grounding system for the JT-60, the condition for the analysis and the result of simulation are reported. (Kako, I.)

Progress in JT-60 joint research

International Nuclear Information System (INIS)

Kimura, Haruyuki; Kikuchi, Mitsuru; Inutake, Masaaki

2007-01-01

It consists of five chapters; 1) introduction, 2) management system of joint plan and researches, 3) progress of joint researches, 4) results of researches and 5) summary. The second chapter stated the structure of management system of JT-60 joint researches, progress of management of the JT-60 experimental theme system, invitation the public to joint researches and selection of the subjects. The progress of joint researches contained the number of subjects, research members and organizations, change of joint research fields, remote control system of experiments, analysis code group, and number of reports. The main results of researches such as development of operation without center solenoid, Magneto-Hydro-Dynamics (MHD) control by electron cyclotron wave, plasma-wall interaction, application of laser technologies to plasma measurement, and comparison between tokamak and helical are reported. (S.Y.)

Development of upgraded pellet injector for JT-60

International Nuclear Information System (INIS)

Onozuka, M.; Shimomura, T.; Tanaka, N.; Iwamoto, S.; Hashiri, N.; Oda, Y.; Minami, M.; Hiratsuka, H.; Kawasaki, K.; Takatsu, H.; Shimizu, M.

1989-01-01

The pneumatic 4-shot pellet injector had been in use for JT-60 (JAERI Tokamak-60) contributing to plasma studies in 1988. It could propel the pellets up to 1.6 km/sec at 50 bar propellant gas. In 1989, the new gun assembly has been reinstalled in the upgraded system to provide higher performance and reliability. The supply pressure of the propellant gas is to be raised to 100 bar to obtain higher pellet velocity up to 2.3 km/sec. The device is now in use for JT-60, and is expected to contribute to further plasma studies. In this paper the outline of features and performance of the device is presented. 5 refs., 9 figs

High Beta Steady State Research and Future Directions on JT-60U and JFT-2M

Science.gov (United States)

Ishida, Shinichi

2003-10-01

JT-60U and JFT-2M research is focused on high beta steady state operation towards economically and environmentally attractive reactors. In JT-60U, a high-βp H-mode plasma was sustained with βN 2.7 for 7.4 s in which neoclassical tearing modes (NTMs) limited the attainable β_N. Real-time tracking NTM stabilization system using ECCD demonstrated complete suppression of NTM leading to recovery of βN before onset of NTM. Performance in a fully non-inductive H-mode plasma was improved up to n_i(0) τE T_i(0) = 3.1 x 10^20 keV s m-3 using N-NBCD with βN 2.4, HH_y,2=1.2 and bootstrap fraction f_BS 0.5. ECH experiments extended the confinement enhancement for dominantly electron heated reversed shear plasmas up to HH_y,2 2 at T_e/Ti 1.25. A world record ECCD efficiency, 4.2 x 10^18 A/W/m^2, was achieved at Te 23 keV with a highly localized central current density. Innovative initiation and current build-up without center solenoid currents were established by LHCD/ECH and bootstrap current up to f_BS 0.9. In JFT-2M, the inside of the vacuum vessel wall was fully covered with low-activation ferritic steel plates to investigate their use in plasmas near fusion conditions. High βN plasmas were produced up to βN = 3.3 with an internal transport barrier (ITB) and a steady H-mode edge. A new H-mode regime with steady high recycling (HRS) and an ITB was exploited leading to βN H_89P 6.2 at n_e/nG 0.7. In 2003, JT-60U will be able to operate for the duration up to 65 s at 1 MA/2.7 T and the heating/current-drive duration up to 30 s at 17 MW to prolong high-βN and/or high-f_BS discharges with feedback controls. JFT-2M is planning to implement wall stabilization experiments in 2004 to pursue plasmas above the ideal no-wall limit using a ferritic wall. The modification of JT-60 to a fully superconducting tokamak is under discussion to explore high-β steady state operation in collision-less plasmas well above no-wall limit with ferritic wall in a steady state.

Long pulse operation of high performance plasmas in JT-60U

International Nuclear Information System (INIS)

Ide, Shunsuke

2005-01-01

Recent experimental progress in JT-60U advanced tokamak research is presented; sustainment of the normalized beta (β N ) - 3 in a normal magnetic shear plasma, the bootstrap current fraction (f BS ) - 45% in a weak shear plasma and ∼75% in a reversed magnetic shear plasma in a nearly full non-inductive current drive condition for longer than the current relaxation time. Achievement of high-density high-radiation fraction together with high-confinement in advanced plasmas was demonstrated. Achievement and foundings in long pulse operations after system modification are presented as well. A 65 s discharge of I p =0.7 MA was successfully obtained. As a result, high-β N of 2.3 was successfully sustained for a very long period of 22.3 s. In addition, a 30 s standard ELMy H-mode plasma of I p up to 1.4 MA has also been obtained. Effectiveness of divertor pumping to control particle recycling and the electron density under the wall retention was saturated was demonstrated. These achievement and issues in the development will be discussed. (author)

Fast collimated neutron flux measurement using stilbene scintillator and flashy analog-to-digital converter in JT-60U

International Nuclear Information System (INIS)

Ishikawa, M.; Itoga, T.; Okuji, T.; Nakhostin, M.; Shinohara, K.; Hayashi, T.; Sukegawa, A.; Baba, M.; Nishitani, T.

2006-01-01

A line-integrated neutron emission profile is routinely measured using the radial neutron collimator system in JT-60U tokamak. Stilbene neuron detectors (SNDs), which combine a stilbene organic crystal scintillation detector (SD) with an analog neutron-gamma pulse shape discrimination (PSD) circuit, have been used to measure collimated neutron flux. Although the SND has many advantages as a neutron detector, the maximum count rate is limited up to ∼1x10 5 counts/s due to the analog PSD circuit. To overcome this issue, a digital signal processing system (DSPS) using a flash analog-to-digital converter (Acqiris DC252, 8 GHz, 10 bits) has been developed at Cyclotron and Radioisotope Center in Tohoku University. In this system anode signals from photomultiplier of the SD are directory stored and digitized. Then, the PSD between neutrons and gamma rays is performed using software. The DSPS has been installed in the vertical neutron collimator system in JT-60U and applied to deuterium experiments. It is confirmed that the PSD is sufficiently performed and collimated neutron flux is successfully measured with count rate up to ∼5x10 5 counts/s without the effect of pileup of detected pulses. The performance of the DSPS as a neutron detector, which supersedes the SND, is demonstrated

3. Laser Thomson scattering by plasmas. 3.2. Applications of incoherent Thomson scattering. 3.2.2. Incoherent Thomson scattering systems for JT-60U and JFT-2M

International Nuclear Information System (INIS)

Hatae, Takaki; Yoshida, Hidetoshi; Naito, Osamu; Yamauchi, Toshihiko

2000-01-01

Development of Thomson scattering diagnostics for the JT-60U and JFT-2M Tokamaks are described. Two Thomson scattering systems have been installed on JT-60U. The first system uses two ruby lasers (10 J, 0.25 Hz) and measures electron temperature (T e ) and density (n e ) profiles of 60 spatial points with high spatial resolution (8 mm). The second system uses a YAG laser (2 J, 30 Hz) and measures time evolution of T e and n e profiles with 15 spatial points. On JFT-2M, a TV Thomson Scattering system (TVTS) has been installed and measures at 81 spatial points with high spatial resolution (8.6 mm). These systems have provided not only profiles of all over the plasma, but successfully measured local structures to study various physics issues; e.g. H-mode edge pedestal, internal transport barrier, local MHD event. (author)

The H-mode pedestal, ELMs and TF ripple effects in JT-60U/JET dimensionless identity experiments

International Nuclear Information System (INIS)

Saibene, G.; Oyama, N.; Loennroth, J.; Andrew, Y.; Luna, E. de la; Giroud, C.; Huysmans, G.T.A.; Kamada, Y.; Kempenaars, M.A.H.; Loarte, A.; Donald, D. Mc; Nave, M.M.F.; Meiggs, A.; Parail, V.; Sartori, R.; Sharapov, S.; Stober, J.; Suzuki, T.; Takechi, M.; Toi, K.; Urano, H.

2007-01-01

This paper summarizes results of dimensionless identity experiments in JT-60U and JET, aimed at the comparison of the H-mode pedestal and ELM behaviour in the two devices. Given their similar size, dimensionless matched plasmas are also similar in their dimensional parameters (in particular, the plasma minor radius a is the same in JET and JT-60U). Power and density scans were carried out at two values of I p , providing a q scan (q 95 = 3.1 and 5.1) with fixed (and matched) toroidal field. Contrary to initial expectations, a dimensionless match between the two devices was quite difficult to achieve. In general, p ped in JT-60U is lower than in JET and, at low q, the pedestal pressure of JT-60U with a Type I ELMy edge is matched in JET only in the Type III ELM regime. At q 95 = 5.1, a dimensionless match in ρ*, ν* and β p,ped is obtained with Type I ELMs, but only with low power JET H-modes. These results motivated a closer investigation of experimental conditions in the two devices, to identify possible 'hidden' physics that prevents obtaining a good match of pedestal values over a large range of plasmas parameters. Ripple-induced ion losses of the medium bore plasma used in JT-60U for the similarity experiments are identified as the main difference with JET. The magnitude of the JT-60U ripple losses is sufficient to induce counter-toroidal rotation in co-injected plasma. The influence of ripple losses was demonstrated at q 95 = 5.1: reducing ripple losses by ∼2 (from 4.3 to 1.9 MW) by replacing positive with negative neutral beam injection at approximately constant P in resulted in an increased p ped in JT-60U, providing a good match to full power JET H-modes. At the same time, the counter-toroidal rotation decreased. Physics mechanisms relating ripple losses to pedestal performance are not yet identified, and the possible role of velocity shear in the pedestal stability, as well as the possible influence of ripple on thermal ion transport are briefly

Dynamic transport study of the plasmas with transport improvement in LHD and JT-60U

International Nuclear Information System (INIS)

Ida, K.; Inagaki, S.; Sakamoto, R.; Tanaka, K.; Fujita, T.; Funaba, H.; Kubo, S.; Yoshinuma, M.; Shimozuma, T.; Takeiri, Y.; Ikeda, K.; Michael, C.; Tokuzawa, T.; Sakamoto, Y.; Takenaga, H.; Isayama, A.; Matsunaga, G.; Ide, S.

2009-01-01

Transport analysis during the transient phase of heating (a dynamic transport study) applied to the plasma with internal transport barriers (ITBs) in the Large Helical Device (LHD) heliotron and the JT-60U tokamak is described. In the dynamic transport study the time of transition from the L-mode plasma to the ITB plasma is clearly determined by the onset of flattening of the temperature profile in the core region and a spontaneous phase transition from a zero curvature ITB (hyperbolic tangent shaped ITB) or a positive curvature ITB (concaved shaped ITB) to a negative curvature ITB (convex shaped ITB) and its back-transition are observed. The flattening of the core region of the ITB transition and the back-transition between a zero curvature ITB and a convex ITB suggest the strong interaction of turbulent transport in space.

Mechanical and thermal characteristics of JT-60 tokamak machine demonstrated in its power tests

International Nuclear Information System (INIS)

Takatsu, Hideyuki; Yamamoto, Masahiro; Ohkubo, Minoru

1985-09-01

JT-60 power tests were carried out from Dec. 10, 1984 to Feb. 20, 1985 to demonstrate, in advance of actual plasma operation, satisfactory performance of tokamak machine, power suppliers and control system in combination. The tests began with low power test of individual coil systems and progressed to full power tests. Power tests were successfully concluded with the following conclusions. (1) All of the coil systems were raised up to full power operation in combination and system performance was verified including thermal and structural integrity of tokamak machine. (2) Measured strain and deflection showed good agreements with those predicted in the design, which was an evidence that electromagnetic loads were supported adequately as expected in the design. (3) Vibration of lateral port was found to be large up to 50 m/s 2 and caused excessive vibration of gate-valves. (4) A few limitations to machine operation were made clear quantatively. (5) It was found that the existing detectors were insufficient to monitor the machine integrity and a few kinds of detectors were necessary to be installed. (author)

Present status of the JT-60 control system

International Nuclear Information System (INIS)

Kimura, T.

1992-01-01

The present status of the control system for a large fusion device of the JT-60 upgrade tokamak is reported including its original design concept, the progress of the system in the past five-year operation and modification for the upgrade. The control system has the features of hierarchical structure, computer control, adoption of CAMAC interfaces and protective interlock by both software and hard-wired systems. Plant monitoring and control are performed by an efficient data communication via CAMAC highways. Sequential discharge control of is executed by a combination of computers and a timing system. A plasma feedback control system with fast 32-bit microprocessors and a man/machine interface with modern workstations have been newly developed for the operation of the JT-60 upgrade. (author)

Review of JT-60U experimental results in 2007 and 2008

International Nuclear Information System (INIS)

Isayama, Akihiko; Oyama, Naoyuki; Suzuki, Takahiro; Shinohara, Kouji; Sakamoto, Yoshiteru; Matsunaga, Go; Yoshida, Maiko; Asakura, Nobuyuki; Nakano, Tomohide; Kamiya, Kensaku; Itami, Kiyoshi

2010-02-01

Results in JT-60U experiments in 2007 and 2008 are reviewed. In this campaign, which is the final experimental period in JT-60U, development of advanced tokamak plasma was extensively performed toward establishment of physics basis of ITER and DEMO. High integrated performance plasma with high normalized beta (β N -2.6) and high confinement enhancement factor (H H98(y,2) -1.0-1.1), which are comparable to those in the ITER Hybrid Scenario, and at the same time with high bootstrap current fraction (f BS -40%) was sustained for 25 s. High density and high radiation loss fraction plasma was sustained for 12 s by adding argon and neon to a deuterium plasma. The duration of the high-performance plasmas is more than 10 times longer than the current diffusion time, τ R . In a high beta regime exceeding the ideal MHD limit without conducting wall (no-wall limit), a new instability was observed. By suppressing the instability a high beta plasma was sustained for 5 s, which corresponds to several times longer than τ R . Performance of reversed shear plasmas was significantly improved by utilizing the stabilizing effect of the conducting wall, and β N -2.7 and f BS -90% were obtained. These results significantly exceed those in the previous experimental campaign. In addition, real-time control system was improved, and ion temperature and current profile were independently or simultaneously controlled in real time. Development of new diagnostics was also continuously performed. For example, profiles of electron density and current were measured using the lithium beam probe diagnostic with high resolution. A number of important results from physics experiments were obtained in the area of transport, confinement, instability, plasma-wall interaction etc. Performance of heating and current drive systems was also extended significantly. In the electron cyclotron wave system, 2.9 MW for 5 s injection and 0.4 MW for 30 s injection to plasma were successfully demonstrated. Power

Neutronic analysis of fusion tokamak devices by PHITS

International Nuclear Information System (INIS)

Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

2011-01-01

A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

ASDEX Upgrade-JT-60U comparison and ECRH power requirements for NTM stabilization in ITER

International Nuclear Information System (INIS)

Urso, L.; Zohm, H.; Maraschek, M.; Poli, E.; Isayama, A.

2010-01-01

Neoclassical tearing modes (NTMs) are experimentally controlled with local electron cyclotron current drive (ECCD) and the island width decay during NTM stabilization is modelled using the so-called modified Rutherford equation (MRE). In this paper, a modelling of the MRE is carried out and simulations of the island width decay are compared with the experimentally observed ones in order to fit the two free machine-independent parameters present in the equation. A systematic study on a database of NTM stabilization discharges from ASDEX Upgrade and JT-60U is done for extrapolating the ECCD power requirements for ITER. The extrapolation to ITER of the NTM stabilization results from ASDEX Upgrade and JT-60U shows that 10 MW of ECCD power are enough to stabilize large NTMs. The 10 MW power estimate for ITER is based on the assumption that the free parameters in the MRE are machine independent. Indeed, this assumption is verified in this paper for ASDEX Upgrade and JT-60U. An interesting consequence of the relatively modest power requirement for ITER is that the installed 20 MW will suffice for simultaneous 2/1 and 3/2 NTM stabilization.

Compatibility of advanced tokamak plasma with high density and high radiation loss operation in JT-60U

International Nuclear Information System (INIS)

Takenaga, H.; Asakura, N.; Kubo, H.; Higashijima, S.; Konoshima, S.; Nakano, T.; Oyama, N.; Ide, S.; Fujita, T.; Takizuka, T.; Kamada, Y.; Miura, Y.; Porter, G.D.; Rognlien, T.D.; Rensink, M.E.

2005-01-01

Compatibility of advanced tokamak plasmas with high density and high radiation loss has been investigated in both reversed shear (RS) plasmas and high β p H-mode plasmas with a weak positive shear on JT-60U. In the RS plasmas, the operation regime is extended to high density above the Greenwald density (n GW ) with high confinement (HH y2 >1) and high radiation loss fraction (f rad >0.9) by tailoring the internal transport barriers (ITBs). High confinement of HH y2 =1.2 is sustained even with 80% radiation from the main plasma enhanced by accumulated metal impurity. The divertor radiation is enhanced by Ne seeding and the ratio of the divertor radiation to the total radiation is increased from 20% without seeding to 40% with Ne seeding. In the high β p H-mode plasmas, high confinement (HH y2 =0.96) is maintained at high density (n-bar e /n GW =0.92) with high radiation loss fraction (f rad ∼1) by utilizing high-field-side pellets and Ar injections. The high n-bar e /n GW is obtained due to a formation of clear density ITB. Strong core-edge parameter linkage is observed, as well as without Ar injection. In this linkage, the pedestal β p , defined as β p ped =p ped /(B p 2 /2μ 0 ) where p ped is the plasma pressure at the pedestal top, is enhanced with the total β p . The radiation profile in the main plasma is peaked due to Ar accumulation inside the ITB and the measured central radiation is ascribed to Ar. The impurity transport analyses indicate that Ar accumulation by a factor of 2 more than the electron, as observed in the high β p H-mode plasma, is acceptable even with peaked density profile in a fusion reactor for impurity seeding. (author)

The design study of the JT-60SU device. No.8. Nuclear shielding and safety design

Energy Technology Data Exchange (ETDEWEB)

Miya, Naoyuki; Kikuchi, Mitsuru; Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1998-03-01

Results of nuclear shielding design study and safety analysis for the steady-state tokamak device JT-60SU are described. D-T operation (option) for two years is adopted in addition to ten years operation using deuterium. Design work has been done in accordance with general laws for radioisotopes handling in Japan as a guideline of safety evaluation, which is applied to the operation of present JT-60U device. Optimization of the shielding design for the device structure including vacuum vessel has been presented to meet with allowable limits of biological shielding determined in advance. It is shown that JT-60SU can be operated safely in the present JT-60 experimental building. It is planed to use 100g/year of tritium in D-T operation phase. A concept of multiple -barrier system is applied to the facility design to prevent propagation of tritium, in which the torus hall and the tritium removal room provide the tertiary confinement. From the design of atmosphere detritiation system for accidental tritium release, it is shown that tritium concentration level can be reduced to the allowable level after two weeks with reasonable compact size components. Safety assessment related to activation of coolant/air, and atmospheric tritium effluents are discussed. (author)

Experimental study of radiation losses on the JT-60 tokamak

International Nuclear Information System (INIS)

Nishitani, Takeo

1990-06-01

Bolometric measurement system and associated diagnostics, soft x-ray pulse-height-analyzer, soft x-ray intensity and Balmer α line measurement systems, were developed to investigate the radiation losses of the JT-60 plasmas. The bolometric measurement is the most important diagnostics in the radiation loss study. The soft x-ray pulse-height-analyzer is useful to estimate the metallic impurity concentration, and the soft x-ray intensity and Balmer α line measurements are monitors of radiation in x-ray region and particle recycling in the plasma edge, respectively. In JT-60, the marfe has been observed frequently in high-Ip and high density limited discharges with NB heating after the replacement of the first wall from TiC coated molybdenum tiles to graphite ones. The threshold electron density of the marfe onset increased with the NB power. The empirical scaling of the marfe onset taking account of the NB power was obtained. This scaling was useful to predict the marfe onset condition in NB heated discharges on JT-60. The marfe was modelled based on the radiative thermal instability. The simple model can explain the marfe onset condition. The radiated power from the plasma with marfe was about 90 % of the absorbed power. Both stored energy and central electron temperatures did not change by the marfe onset in spite of the such intense radiation loss. Finally, this study revealed that the most clean plasma was obtained in the metallic first wall with the divertor on JT-60. This fact is suggesting the capability of the metallic material for the first wall of next devices. Enhance radiation localized in the peripheral plasma such as marfe and IDC does not degrade the core plasma confinement or somewhat improves it, so that marfe and IDC are suitable operational regime in the high density region for future devices because they have strong remote-radiative-cooling-effect. (J.P.N.)

Machine performance and its effects on experiments in JT-60U

International Nuclear Information System (INIS)

Kondo, I.

1995-01-01

The operational results of JT-60U were reviewed in light of the strategy made at the design stage. The operational plan for better confinement shifted from that of low q to high poloidal beta plasma configuration with higher q value according to the revealed machine properties. Some technical and operational skills helped bring about the recent results out of the machine. (orig.)

Design, fabrication and test of double-wall vacuum vessel for JT-60U

International Nuclear Information System (INIS)

Uchikawa, Takashi; Ioki, Kimihiro; Ninomiya, Hiromasa.

1994-01-01

A double-wall vacuum vessel was designed and fabricated for JT-60U (an upgraded machine of JT-60), which has a plasma current up to 6 MA and a large plasma volume (100 m 3 ). A new concept of Inconel 625 all-welded structure was adopted to the vessel, that comprises an inner plate, square tubes and an outer plate. The vacuum vessel with a multi-arc D-shaped cross section was fabricated by using hot-sizing press. The electromagnetic and structural analysis has been performed for plasma disruption loads. Dynamic responses of the vessel were measured during plasma disruptions, and the observed displacement had a good agreement with the result of FEM analysis. (author)

Study of particle pumping characteristics for different pumping geometries in JT-60U and DIII-D divertors

International Nuclear Information System (INIS)

Takenaga, H.; Sakasai, A.; Kubo, H.

2001-01-01

Particle pumping characteristics were compared between pumping from the inner side private flux region (IPP) and pumping from both sides of the private flux region (BPP) in the JT-60U W shaped divertor, and between JT-60U IPP and pumping in the DIII-D lower baffled divertor. The pumping flux for BPP is smaller than that for IPP by about a factor of 2 with weak in-out asymmetry of recycling neutral flux and by a factor of 3.5-6.5 with strong in-out asymmetry. The reduction of the pumping flux for BPP is consistent with Monte Carlo simulations, where backflow at the outer pumping slot is observed due to in-out recycling asymmetry. The pumping flux in DIII-D at I p =0.8 MA and B T =1.6 T is comparable to or smaller than that for JT-60U IPP at I p =1.0 MA, B T =3.8 T and I p =1.5 MA, B T =3.5 T in the same density regime. In the DIII-D divertor with pumping from the private flux region, the pumping flux decreases with increasing in-out asymmetry. The pumping flux normalized by the integrated D α emission over the whole plasma exhibits a similar dependence on the distance between the pumping slot and the strike point in JT-60U IPP and the DIII-D lower divertor with pumping through the outer divertor plasma region. (author)

High Mach flow associated with plasma detachment in JT-60U

International Nuclear Information System (INIS)

Hatayama, A.; Hoshino, K.; Miyamoto, K.

2003-01-01

Recent new results of the high Mach flows associated with plasma detachment are presented on the basis of numerical simulations by a 2-D edge simulation code (the B2-Eirene code) and their comparisons with experiments in JT-60U W-shaped divertor plasma. High Mach flows appear near the ionization front away from the target plate. The plasma static pressure rapidly drops, while the total pressure is kept almost constant near the ionization front, because the ionization front near the X-point is clearly separated from the momentum loss region near the target plate. Redistribution from static to dynamic pressure without a large momentum loss is confirmed to be a possible mechanism of the high Mach flows. It has been also shown that the radial structure of the high Mach flow near the X point away from the target plate has a strong correlation with the DOD (Degree of Detachment) at the target plate. Also, we have made systematic analyses on the high Mach flows for both the 'Open' geometry and the 'W-shaped' geometry of JT-60U in order to clarify the geometric effects on the flows. (author)

Dynamic transport study of the plasmas with transport improvement in LHD and JT-60U

International Nuclear Information System (INIS)

Ida, K.; Inagaki, S.; Sakamoto, R.; Tanaka, K.; Funaba, H.; Kubo, S.; Yoshinuma, M.; Shimozuma, T.; Takeiri, Y.; Ikeda, K.; Michael, C.; Tokuzawa, T.; Sakamoto, Yoshiteru; Takenaga, Hidenobu; Isayama, Akihiko; Ide, Shunsuke; Fujita, Takaaki

2006-10-01

A transport analysis during the transient phase of heating (a dynamic transport study) applied to the plasma with internal transport barriers (ITBs) in the Large Helical Device (LHD) heliotron and JT-60U tokamak is described. In the dynamic transport study 1) a slow transition between two transport branches is observed, 2) the time of the transition from the L-mode plasma to the ITB plasma is clearly determined by the onset of the flattening of the temperature profile in the core region and 3) a spontaneous phase transition from a weak, wide ITB to a strong, narrow ITB and its back-transition are observed. The flattening of the core region of the ITB transition and the back-transition between a wide ITB and a narrow ITB suggest the strong interaction of turbulent transport in space, where turbulence suppression at certain locations in the plasma causes the enhancement of turbulence and thermal diffusivity nearby. (author)

Ion transport analysis of a high beta-poloidal JT-60U discharge

International Nuclear Information System (INIS)

Horton, W.; Tajima, T.; Dong, J.-Q.; Kim, J.-Y.; Kishimoto, Y.

1997-01-01

The high beta-poloidal discharge number 17110 in JT-60U (JT-60 Team, IAEA, Vienna, 1993) that developes an internal transport barrier is analysed for the transport of ion energy and momentum. First, the classical ion temperature gradient stability properties are calculated in the absence of sheared plasma flows to establish the L-mode transport level prior to the emergence of the transport barrier. Then the evolving toroidal and poloidal velocity profiles reported by Koide et al (1994 Phys. Rev. Lett. 72 3662) are used to show how the sheared mass flows control the stability and transport. Coupled energy-momentum transport equations predict the creation of a transport barrier. The balance of the steep ion temperature gradient against the magnetic shear and sheared mass flow is calculated for the profiles in the 17110 discharge. (Author)

Deuterium depth profiling in JT-60U W-shaped divertor tiles by nuclear reaction analysis

International Nuclear Information System (INIS)

Hayashi, T.; Ochiai, K.; Masaki, K.; Gotoh, Y.; Kutsukake, C.; Arai, T.; Nishitani, T.; Miya, N.

2006-01-01

Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JAERI Tokamak-60 Upgrade (JT-60U) were investigated by nuclear reaction analysis (NRA). The highest deuterium concentration of D/ 12 C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium-carbon co-deposition. In the outer and inner divertor target tiles, the D/ 12 C data were lower than 0.006. Additionally, the maximum (H + D)/ 12 C in the dome top tile was estimated to be 0.023 from the results of NRA and secondary ion mass spectroscopy (SIMS). Orbit following Monte-Carlo (OFMC) simulation showed energetic deuterons caused by neutral beam injections (NBI) were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, scanning electron microscope (SEM) observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium-carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk

Long Pulse Operation on NBI Systems for JT-60U

International Nuclear Information System (INIS)

Akino, N.; Ebisawa, N.; Honda, A.; Ikeda, Y.; Kawai, M.; Kazawa, M.; Mogaki, K.; Ohga, T.; Umeda, N.; Usui, K.; Yamamoto, T.; Grisham, L.

2005-01-01

In the neutral beam injection (NBI) system, an extension of the pulse duration up to 30 sec has been intended to study quasi-steady state plasma on JT-60U. The four positive-ion based (P-NBI) units, which tangentially inject neutral beam to plasma, were mainly modified on the electric power supplies and the beam limiters to extend the pulse duration up to 30 sec with 2 MW at 80 keV per each. The seven P-NBI units, each of which perpendicularly injects for 10 sec, were conducted to operate in series for the total pulse duration of 30 sec. The ion source of the negative-ion based (N-NBI) unit, whose target beam energy is 500 keV for 10 sec, was also modified to reduce the heat load of the grid for long pulse operation. The reduction of the re-ionization of the neutral beam in the beam drift duct was a key to achieve a long pulse injection. It was found that the pressure rise in the beam drift duct, which gives the re-ionization rate, depended on the temperature of the re-ionization plates during NBI injection. Up to now, it was attained successfully that the pulse duration of the tangential P-NBI unit was extended up to 30 sec. 310 MJ of the total integrated injection energy into JT-60U plasma was achieved, including the negative-ion based NBI operation for 17 sec at 366 keV

Design of JT-60SA magnets and associated experimental validations

International Nuclear Information System (INIS)

Zani, L.; Barabaschi, P.; Peyrot, M.; Meunier, L.; Tomarchio, V.; Duglue, D.; Decool, P.; Torre, A.; Marechal, J.L.; Della Corte, A.; Di Zenobio, A.; Muzzi, L.; Cucchiaro, A.; Turtu, S.; Ishida, S.; Yoshida, K.; Tsuchiya, K.; Kizu, K.; Murakami, H.

2011-01-01

In the framework of the JT-60SA project, aiming at upgrading the present JT-60U tokamak toward a fully superconducting configuration, the detailed design phase led to adopt for the three main magnet systems a brand new design. Europe (EU) is expected to provide to Japan (JA) the totality of the toroidal field (TF) magnet system, while JA will provide both Equilibrium field (EF) and Central Solenoid (CS) systems. All magnet designs were optimized trough the past years and entered in parallel into extensive experimentally-based phases of concept validation, which came to maturation in the years 2009 and 2010. For this, all magnet systems were investigated by mean of dedicated samples, e.g. conductor and joint samples designed, manufactured and tested at full scale in ad hoc facilities either in EU or in JA. The present paper, after an overall description of magnet systems layouts, presents in a general approach the different experimental campaigns dedicated to qualification design and manufacture processes of either coils, conductors and electrical joints. The main results with the associated analyses are shown and the main conclusions presented, especially regarding their contribution to consolidate the triggering of magnet mass production. The status of respective manufacturing stages in EU and in JA are also evoked. (authors)

Characteristics of divertor plasma and scrape-off layer in JT-60U

International Nuclear Information System (INIS)

Itami, K.; Shimada, M.; Hosogane, N.

1992-01-01

Heat flux to the divertor is measured by thermography and the heat transport in the scrape-off layer is studied in beam heated discharges of JT-60U. The heat flux onto the divertor is ∝50% of total beam power at maximum. The in-out asymmetry of the heat flux P HEAT in /P HEAT out is as large as 20-40% when the ion grad-B drift is toward the divertor. Differences in P HEAT in /P HEA T out due to the direction of ion grad-B drift are as large as large as ∝40%. A scaling of the peaking factor Y of heat flux, defined by Y=2πRfq max /P HEAT , is obtained for beam heated discharges in JT-60U with a wide range of plasma parameters. The Y corresponds to the inverse of the thickness of the scrape-off layer. From a statistical analysis, it is found that the peaking factor Y of heat flux scales as P HEAT 0.49±0.18 anti n e -0.45±0.22 q eff -0.67±0.18 . (orig.)

Development of JT-60 diagnostics system

International Nuclear Information System (INIS)

Suzuki, Yasuo

1988-01-01

The various kinds of plasma diagnostics have been developed and utilized in the JT-60 experiments. The features of JT-60 diagnostics system and the historical proceeding of the development are described in this paper. Taking account of the design consideration, JT-60 diagnostics system is sorted out into eight groups, which include six diagnostics systems, the data processing system and diagnostics supporting system. The all devices in the JT-60 diagnostics system were instrumented on schedule in the end of the fiscal year of 1985 and have contributed to JT-60 experiments. (author)

Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

Energy Technology Data Exchange (ETDEWEB)

Matsukawa, M. E-mail: matsukaw@naka.jaeri.go.jp; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T

2003-09-01

The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control.

Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

International Nuclear Information System (INIS)

Matsukawa, M.; Ishida, S.; Sakasai, A.; Urata, K.; Senda, I.; Kurita, G.; Tamai, H.; Sakurai, S.; Miura, Y.M.; Masaki, K.; Shimada, K.; Terakado, T.

2003-01-01

The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad-Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control

Characteristics of large scale ionic source for JT-60

Energy Technology Data Exchange (ETDEWEB)

Fujiwara, Yukio; Honda, Atsushi; Inoue, Takashi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1997-02-01

The Neutral Beam Injection (NBI) apparatus is expected for important role sharing apparatus to realize the plasma electric current drive and the plasma control in not only temperature upgrading of the plasma but also Tokamak nuclear fusion reactor for the next generation such as JT-60, ITER and so forth. Japan Atomic Energy Research Institute has developed the ionic source with high energy and large electric current for about 10 years. Some arrangement tests of the large negative ion source for JT-60 No. 1 were executed from June to October, 1995. As a series of arrangement tests, 400 KeV and 13.5 A of deuterium negative ion beam was successfully accelerated for 0.12 sec. under 0.22 Pa of low gas pressure. And, it was elucidated that electron electric current could be controlled efficiently even in deuterium negative ion beam. Here is described on the testing results in details. (G.K.)

Analysis of JT-60SA operational scenarios

Science.gov (United States)

Garzotti, L.; Barbato, E.; Garcia, J.; Hayashi, N.; Voitsekhovitch, I.; Giruzzi, G.; Maget, P.; Romanelli, M.; Saarelma, S.; Stankiewitz, R.; Yoshida, M.; Zagórski, R.

2018-02-01

Reference scenarios for the JT-60SA tokamak have been simulated with one-dimensional transport codes to assess the stationary state of the flat-top phase and provide a profile database for further physics studies (e.g. MHD stability, gyrokinetic analysis) and diagnostics design. The types of scenario considered vary from pulsed standard H-mode to advanced non-inductive steady-state plasmas. In this paper we present the results obtained with the ASTRA, CRONOS, JINTRAC and TOPICS codes equipped with the Bohm/gyro-Bohm, CDBM and GLF23 transport models. The scenarios analysed here are: a standard ELMy H-mode, a hybrid scenario and a non-inductive steady state plasma, with operational parameters from the JT-60SA research plan. Several simulations of the scenarios under consideration have been performed with the above mentioned codes and transport models. The results from the different codes are in broad agreement and the main plasma parameters generally agree well with the zero dimensional estimates reported previously. The sensitivity of the results to different transport models and, in some cases, to the ELM/pedestal model has been investigated.

Steady-state exhaust of helium ash in the W-shaped divertor of JT-60U

International Nuclear Information System (INIS)

Sakasai, A.; Takenaga, H.; Hosogane, N.

2001-01-01

By injecting a neutral beam of 60 keV helium (He) atoms as central fueling of helium into the ELMy H-mode plasmas, helium exhaust has been studied in the W-shaped pumped divertor on JT-60U. Efficient He exhaust was realized by He pumping using argon frosted cryopumps in the JT-60U new divertor. In steady state, good He exhaust capability (τ He */τ E =4 and high enrichment factor, where τ He * is a global particle confinement time of helium and τ E is the energy confinement time) was successfully demonstrated in attached ELMy H-mode plasmas. Good He exhaust capability was also obtained in detached ELMy H-mode plasmas, which was comparable to one in attached plasmas. This result of the helium exhaust is sufficient to support a detached divertor operation on ITER. After the divertor modification, helium exhaust in reversed shear plasmas has been investigated using He gas puff. Helium removal inside the internal transport barrier (ITB) is about two times as difficult as that outside the ITB in reversed shear discharges. (author)

Role of low order rational q values on the ITB-events in JT-60U plasmas

International Nuclear Information System (INIS)

Neudatchin, S.V.; Takizuka, T.; Hayashi, N.; Shirai, H.; Fujita, T.; Isayama, A.; Kamada, Y.; Koide, Y.; Suzuki, T.

2003-01-01

The formation of internal transport barriers (ITBs) near q=2,3 surfaces in normal (NrS) or optimized shear discharges of JT-60U and JET is well known. In reverse shear (RS) JT-60U plasmas, the role of q minimum (q min ) equal to 3, 5, 3, 2, 5, 2 is not obvious for ITB formation. ITB-events (non-local confinement bifurcations inside and around ITB in a ms timescale) are found in various JT-60U NrS and RS plasmas. Under sufficient power, ITB-events are seen at rational and not rational values of q min . The space-time evolution of T e and T i is similar even being strongly varied in space and time, suggesting same mechanism(s) of T e and T i transport. The temporal formation of strong ITB in H-mode under passing of q min =3 (after periodical improvements and degradations via ITB-events with 8ms period) in RS mode with P nbi =8MW is presented. Under smaller power, ITB-events are observed only at rational values of q min . In a weak RS shot with P nbi =4MW, abrupt rise of T e is seen at q min =3.5, while more cases of T i rise are observed. The difference of the T e and T i evolution seen regularly under the low power, suggests decoupling of T e and T i transport. (author)

Role of low order rational q values in the ITB-events in JT-60U plasmas

International Nuclear Information System (INIS)

Neudatchin, S.V.

2002-01-01

The formation of internal transport barriers (ITBs) near q=2,3 surfaces in normal (NrS) or optimized shear discharges of JT-60U and JET is well known. In reverse shear (RS) JT-60U plasmas, the role of q minimum (q min ) equal to 3.5, 3, 2.5, 2 is not obvious for ITB evolution. ITB-events (non-local confinement bifurcations inside and around ITB in a ms timescale) are found in various JT-60U NrS and RS plasmas. (a) Under sufficient power, ITB-events are seen at rational and not rational values of q min . The space-time evolution of Te and Ti is similar. The temporal creation of stronger ITB in H-mode (after periodical improvements and degradations via ITB-events with 8 ms period) under passing of q min = 3 is presented (P nbi = 8 MW, 1.5 MA / 3.7 T). (b) Under smaller power, the influence of some rational q min is seen clearly for ITB-events on T e . In 1.3MA/3.7T shot with very weak RS (Pnbi = 4 MW), abrupt rise of T e is seen at q min 3:5, while more cases of T i rise are observed. (c) The possible role of MHD-activity as ITB-events trigger (ms time scale correlation in some NrS and RS cases) is under investigation. (author)

Status of the cold test facility for the JT-60SA tokamak toroidal field coils

Energy Technology Data Exchange (ETDEWEB)

Abdel Maksoud, Walid, E-mail: walid.abdelmaksoud@cea.fr; Bargueden, Patrick; Bouty, André; Dispau, Gilles; Donati, André; Eppelle, Dominique; Genini, Laurent; Guiho, Patrice; Guihard, Quentin; Joubert, Jean-Michel; Kuster, Olivier; Médioni, Damien; Molinié, Frédéric; Sinanna, Armand; Solenne, Nicolas; Somson, Sébastien; Vieillard, Laurence

2015-10-15

Highlights: • The 5 K cryogenic loop includes a 500 W refrigerator and a She cold pump. • The coils are energized thanks to a 25.7 kA power supply and HTS current leads. • Temperature margin tests between 5 K and 7.5 K will be made on each coil. • A magnet safety system protects each double pancake of the coil in case of quench. • Instrumentation is monitored on a 1 Hz to 10 kHz fast acquisition system. - Abstract: JT-60SA is a fusion experiment which is jointly constructed by Japan and Europe and which shall contribute to the early realization of fusion energy, by providing support to the operation of ITER, and by addressing key physics issues for ITER and DEMO. In order to achieve these goals, the existing JT-60U experiment will be upgraded to JT-60SA by using superconducting coils. The 18 TF coils of the JT-60SA device will be provided by European industry and tested in a Cold Test Facility (CTF) at CEA Saclay. The coils will be tested at the nominal current of 25.7 kA and will be cooled with supercritical helium between 5 K and 7.5 K to check the temperature margin against a quench. The main objective of these tests is to check the TF coils performance and hence mitigate the fabrication risks. The most important components of the facility are: a 11.5 m × 6.5 m large cryostat in which the TF coils will be thermally insulated by vacuum; a 500 W helium refrigerator and a valve box to cool the coils down to 5 K and circulate 24 g/s of supercritical helium through the winding pack and through the casing; a power supply and HTS current leads to energize the coil; the control and instrumentation equipment (sensors, PLC's, supervision system, fast data acquisition system, etc.) and the Magnet Safety System (MSS) that protects the coils in case of quench. The paper will give an overview of the design of this large facility and the status of its realization.

Recent results and near-term expectations in Tokamak fusion research in the U.S., Europe, and Japan

International Nuclear Information System (INIS)

Meade, D.

1993-01-01

The development of fusion is often thought about in terms of three different activities: scientific feasibility, engineering feasibility, and economic feasibility. This paper discusses the scientific feasibility of fusion. Reactor temperatures, reactor densities and confinement, particle control, plasma power handling, and self-heating are some of the issues examined. Collaboration and results from research at the Tokamak Fusion Test Reactor (TFTR) at Princeton, the JT-60U in Japan, and JET, the Joint European Torus Tokamak in Oxford are presented

Design study of an AC power supply system in JT-60SA

International Nuclear Information System (INIS)

Shimada, Katsuhiro; Baulaigue, Olivier; Cara, Philippe; Coletti, Alberto; Coletti, Roberto; Matsukawa, Makoto; Terakado, Tsunehisa; Yamauchi, Kunihito

2011-01-01

In the initial research phase of JT-60SA, which is the International Thermonuclear Experimental Reactor (ITER) satellite Tokamak with superconducting toroidal and poloidal magnetic field coils, the plasma heating operation of 30 MW-60 s or 20 MW-100 s is planned for 5.5 MA single null divertor plasmas. To achieve this operation, AC power source of the medium voltage of 18 kV and ∼7 GJ has to be provided in total to the poloidal field coil power supplies and additional heating devices such as neutral beam injection (NBI) and electron cyclotron radio frequency (ECRF). In this paper, the proposed AC power supply system in JT-60SA was estimated from the view point of available power, and harmonic currents based on the standard plasma operation scenario during the initial research phase. This AC power supply system consists of the reused JT-60 power supply facilities including motor generators with flywheel, AC breakers, harmonic filters, etc., to make it cost effective. In addition, the conceptual design of the upgraded AC power supply system for the ultimate heating power of 41 MW-100 s in the extended research phase is also described.

Mock-up test results of monoblock-type CFC divertor armor for JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Higashijima, S. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)], E-mail: higashijima.satoru@jaea.go.jp; Sakurai, S.; Suzuki, S.; Yokoyama, K.; Kashiwa, Y.; Masaki, K.; Shibama, Y.K.; Takechi, M.; Shibanuma, K.; Sakasai, A.; Matsukawa, M.; Kikuchi, M. [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2009-06-15

The JT-60 Super Advanced (JT-60SA) tokamak project starts under both the Japanese domestic program and the international program 'Broader Approach'. The maximum heat flux to JT-60SA divertor is estimated to {approx}15 MW/m{sup 2} for 100 s. Japan Atomic Energy Agency (JAEA) has developed a divertor armor facing high heat flux in the engineering R and D for ITER, and it is concluded that monoblock-type CFC divertor armor is promising for JT-60SA. The JT-60SA armor consists of CFC monoblocks, a cooling CuCrZr screw-tube, and a thin oxygen-free high conductivity copper (OFHC-Cu) buffer layer between the CFC monoblock and the screw-tube. CFC/OFHC-Cu and OFHC-Cu/CuCrZr joints are essential for the armor, and these interfaces are brazed. Needed improvements from ITER engineering R and D are good CFC/OFHC-Cu and OFHC-Cu/CuCrZr interfaces and suppression of CFC cracking. For these purposes, metalization inside CFC monoblock is applied, 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 at the same time is also produced, and the half of the mock-ups could remove 15 MW/m{sup 2} as required. This paper summarizes the recent progress of design and mock-up test results for JT-60SA divertor armor.

Dynamic simulations for preparing the acceptance test of JT-60SA cryogenic system

Science.gov (United States)

Cirillo, R.; Hoa, C.; Michel, F.; Poncet, J. M.; Rousset, B.

2016-12-01

Power generation in the future could be provided by thermo-nuclear fusion reactors like tokamaks. There inside, the fusion reaction takes place thanks to the generation of plasmas at hundreds of millions of degrees that must be confined magnetically with superconductive coils, cooled down to around 4.5 K. Within this frame, an experimental tokamak device, JT-60SA is currently under construction in Naka (Japan). The plasma works cyclically and the coil system is subject to pulsed heat loads. In order to size the refrigerator close to the average power and hence optimizing investment and operational costs, measures have to be taken to smooth the heat load. Here we present a dynamic model of the JT-60SA's Auxiliary Cold box (ACB) for preparing the acceptance tests of the refrigeration system planned in 2016 in Naka. The aim of this study is to simulate the pulsed load scenarios using different process controls. All the simulations have been performed with EcosimPro® and the associated cryogenic library: CRYOLIB.

Multifractality in edge localized modes in Japan Atomic Energy Research Institute Tokamak-60 Upgrade

International Nuclear Information System (INIS)

Bak, P.E.; Asakura, N.; Miura, Y.; Nakano, T.; Yoshino, R.

2001-01-01

The temporal losses of confinement during edge localized modes in the Japan Atomic Energy Research Institute Tokamak-60 Upgrade (JT-60U) show multifractal scaling and the spectra are generally smooth, but in some cases there are signs of discontinuous derivatives. Dynamics of the Sugama-Horton model, interpreted as edge localized modes, also display multifractal scaling. The spectra display singularities in the derivative, which can be interpreted as a phase transition. It is argued that the multifractal spectra of edge localized modes can be used to discriminate between different experimental discharges and validate edge localized mode models

Structure design of the central solenoid in JT-60SA

International Nuclear Information System (INIS)

Asakawa, Shuji; Tsuchiya, Katsuhiko; Kuramochi, Masaya; Yoshida, Kiyoshi

2009-09-01

The upgrade of JT-60U magnet system to superconducting coils (JT-60SA: JT-60 Super Advanced) has been decided by parties of Japanese government (JA) and European commission (EU) in the framework of the Broader Approach (BA) agreement. The magnet system for JT-60SA consists of a central solenoid (CS), equilibrium field(EF) coils, toroidal field(TF) coils. The central solenoid consists the four winding pack modules. In order to counteract the thermal contraction as well as the electric magnetic repulsion and attraction together with other forces generated in each module, it is necessary to apply pre-loading to the support structure of the solenoid and to pursue a structure which is capable of sustaining such loading. In the present report, the structural design of the supporting structure of the solenoid and the jackets of the modules is verified analytically, and the results indicate that the structural design satisfies the 'Codes for Fusion Facilities - Rules on Superconducting Magnet Structure -'. (author)

Resistive instabilities in reversed shear discharges and wall stabilization on JT-60U

International Nuclear Information System (INIS)

Takeji, S.; Tokuda, S.; Fujita, T.; Suzuki, T.; Isayama, A.; Ide, S.; Ishii, Y.; Kamada, Y.; Koide, Y.; Matsumoto, T.; Oikawa, T.; Ozeki, T.; Sakamoto, Y.

2001-01-01

Resistive instabilities and wall stabilization of ideal low toroidal mode number, n, kink modes are investigated in JT-60U reversed shear discharges. Resistive interchange modes with n=1 are found to appear in reversed shear discharges with large pressure gradient at the normalized beta, β N , of about unity or even lower. The resistive interchange modes appear as intermittent burst-like magnetohydrodynamic (MHD) activities and higher n≤3 modes are observed occasionally in higher β N regime. No clear degradation of the plasma stored energy is observed by the resistive interchange modes themselves. It is also found that resistive interchange modes can lead to major collapse owing to a coupling with tearing modes at the outer mode rational surface over the minimum safety factor. Stability analysis revealed that stability parameter of tearing modes, Δ' , at the outer mode rational surface is affected by the free-boundary condition. The result is consistent with the experimental evidence that major collapse tends to occur when plasma edge safety factor, q*, is near integer values. Stabilization of ideal low n kink modes by the JT-60U wall is demonstrated. Magnetohydrodynamic perturbations that are attributed to resistive wall modes are observed followed by major collapse in wall-stabilized discharges. (author)

High coupling performance of JT-60U ICRF antennas

International Nuclear Information System (INIS)

Saigusa, M.; Moriyama, S.; Fujii, T.; Kimura, H.; Sato, M.; Hosogane, N.; Nemoto, M.; Yamamoto, T.

1994-01-01

Sufficient coupling of an ICRF antenna for high power experiments was obtained even for a wide gap between the separatrix and the antenna in JT-60U. The loading resistances for an out-of-phase mode are over 4 Ω for a gap of 13 cm between the separatrix and the Faraday shield over the wide range of electron density from 1 x 10 19 to 5.5 x 10 19 m -3 . In particular, the loading resistances for an in-phase mode are about 5 Ω for a gap of 33 cm between the separatrix and the Faraday shield for the same plasma parameters. However, the heating response for the out-of phase mode is much stronger than that for the in-phase mode. (author). Letter-to-the-editor. 11 refs, 6 figs

Balance of ionization and recombination of carbon ions in high density peripheral plasmas of the JT-60 U tokamak

International Nuclear Information System (INIS)

Nakano, T.; Kubo, H.; Asakura, N.; Shimizu, K.

2009-01-01

In high density and low temperature peripheral plasmas of JT-60 U, i.e. detached divertor plasmas, C III and C IV lines were observed by a visible and VUV spectrometers in order to investigate dominant radiators, radiation power and particle balance between the radiators. An emission peak was found between the inner strike and the X-point. With increasing electron density, the emission peak moved to the X-point with a constant electron temperature of ∼7 eV. In the case the emission peak was located on the X-point, the dominant radiators in the emission peak were C 2+ and C 3+ , which contributed 30% and 60% to the total radiative power. It was found that C 3+ was produced by the ionization of C 2+ and the volume recombination of C 4+ at a similar rates. However, the loss flux of C 3+ was lower by two orders of magnitude than the C 3+ production flux, indicating that another loss mechanism such as transport loss around the X-point was significant.

Construction and testing of JT-60

International Nuclear Information System (INIS)

Kishimoto, H.; Aikawa, H.; Oikawa, A.; Miya, N.; Suzuki, K.; Ozeki, T.; Tokutake, T.; Kunieda, S.; Hiruta, K.; Hosoda, R.

1987-01-01

The JT-60 project is reviewed in terms of design, R and D, construction, commissioning and project management. Design features of JT-60 have been refined and renewed through periodic assessments. Engineering targets have been achieved by R and D efforts. Construction and commissioning have progressed on schedule with intensive project management and control. JT-60 obtained high performance and has entered into the experimental phase after completion of machine construction. (orig.)

The life test of a DC circuit breaker of tokamak device JT-60 for a nuclear fusion research

International Nuclear Information System (INIS)

Shimada, Ryuichi; Tani, Keiji; Kishimoto, Hiroshi; Tamura, Sanae; Yanabu, Satoru.

1979-01-01

In the Tokamak devices for nuclear fusion research, the construction of the current transformer circuits having plasma as the secondary circuit and the change of the primary circuit current are necessary for generating current in the plasma. This is considered to be fairly difficult in practice if conventional methods using capacitor discharge and iron core coils are employed. Considering such circumstances, it was decided for JT-60 to use an air-core current transformer coil and to employ the method of storing energy in the form of current in the coil inductance instead of a capacitor. For this reason, a DC circuit breaker is required to interrupt coil current. The authors improved an AV vacuum breaker, which had been developed as the vacuum breaker of longitudinal magnetic field type applying a magnetic field in parallel with an arc, to get the one for DC circuit for the purpose of applying it to JT-60. In this paper, the operational characteristic of the DC breaker is described, the construction and function of the life test circuit is explained, and the test results are reported. Finally, interruptions of 10,000 times at 20 kA were carried out. It is successful that the restrike of arc occurring during tens of milli-seconds after interruptions was improved to 0.05% or less for 10,000 times operations. Further, it was found that the generation of arc restrike can be reduced practically to zero with two breakers in series. (Wakatsuki, Y.)

Structural analysis of the JT-60SA cryostat vessel body

Energy Technology Data Exchange (ETDEWEB)

Botija, José, E-mail: jose.botija@ciemat.es [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Alonso, Javier; Fernández, Pilar; Medrano, Mercedes; Ramos, Francisco; Rincon, Esther; Soleto, Alfonso [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Davis, Sam; Di Pietro, Enrico; Tomarchio, Valerio [Fusion for Energy, JT-60SA European Home Team, 85748 Garching bei Munchen (Germany); Masaki, Kei; Sakasai, Akira; Shibama, Yusuke [JAEA – Japan Atomic Energy Agency, Naka Fusion Institute, Ibaraki 311-0193 (Japan)

2013-10-15

Highlights: ► Structural analysis to validate the JT-60SA cryostat vessel body design. ► Design code ASME 2007 “Boiler and Pressure Vessel Code. Section VIII”. ► First buckling mode: load multiplier of 10.644, higher than the minimum factor 4.7. ► Elastic and elastic–plastic stress analysis meets ASME against plastic collapse. ► Bolted fasteners have been analyzed showing small gaps closed by strong welding. -- Abstract: The JT-60SA cryostat is a stainless steel vacuum vessel (14 m diameter, 16 m height) which encloses the Tokamak providing the vacuum environment (10{sup −3} Pa) necessary to limit the transmission of thermal loads to the components at cryogenic temperature. It must withstand both external atmospheric pressure during normal operation and internal overpressure in case of an accident. The paper summarizes the structural analyses performed in order to validate the JT-60SA cryostat vessel body design. It comprises several analyses: a buckling analysis to demonstrate stability under the external pressure; an elastic and an elastic–plastic stress analysis according to ASME VIII rules, to evaluate resistance to plastic collapse including localized stress concentrations; and, finally, a detailed analysis with bolted fasteners in order to evaluate the behavior of the flanges, assuring the integrity of the vacuum sealing welds of the cryostat vessel body.

Design and performance tests of gas circulation heating of JT-60U vacuum vessel

International Nuclear Information System (INIS)

Yotsuga, M.; Masuzaki, T.; Sago, H.; Nishikane, M.; Uchikawa, T.; Iritani, Y.; Murakami, T.; Horiike, H.; Neyatani, Y.; Ninomiya, H.; Matsukawa, M.; Ando, T.; Miyachi, I.

1992-01-01

This paper reports that in the final stage of construction of the upgraded JT-60 device (JT-60U), baking tests of the vacuum vessel was performed. The vessel torus was heated-up to 300 degrees C by means of the nitrogen gas circulation system and electric heaters mounted on the outboard solid wall of the vessel. The design of the gas flow channels inside the double-wall structure of the vessel was done based on flow model tests, fluid analysis, and flow network analysis. The results of the baking tests were satisfactory. In maintaining 300 degrees C bake-out temperature, required heating power of the gas circulation system and outboard heaters was 520kW and 50kW, respectively. The temperature distribution over the vessel wall was within 300 ± 30 degrees C. It was also shown or suggested that heat-up and cool-down time is about 30 hours. The baking tests data have been reflected on operations for plasma experiments

Installation and pre-commissioning of the cryogenic system of JT-60SA tokamak

Science.gov (United States)

Hoa, C.; Michel, F.; Roussel, P.; Fejoz, P.; Girard, S.; Goncalves, R.; Lamaison, V.; Natsume, K.; Kizu, K.; Koide, Y.; Yoshida, K.; Cardella, A.; Portone, A.; Verrecchia, M.; Wanner, M.; Beauvisage, J.; Bertholat, F.; Gaillard, G.; Heloin, V.; Langevin, B.; Legrand, J.; Maire, S.; Perrier, J. M.; Pudys, V.

2017-02-01

The cryogenic system for the superconducting tokamak JT-60SA is currently being commissioned in Naka, Japan and shall be ready for operation in summer 2016. This contribution is part of the Broader Approach agreement between Japan and Europe. With an equivalent refrigeration capacity of about 9.5 kW at 4.5 K the cryogenic system will supply cryo-pump panels at 3.7 K, superconducting magnets and their structures at 4.4 K, high temperature superconducting current leads at 50 K and thermal shields between 80 K and 100 K. The system has been specifically designed to handle large pulse loads at 4.4 K during plasma operation. The mechanical and electrical assembly of the cryogenic system has been achieved within six months by October 2015. The main contractor Air Liquide Advanced Technology (AL-aT) have supplied eight parallel working screw compressors with a common oil removal and dryer system, a Refrigeration Cold Box and an Auxiliary Cold box with cold rotating machines. F4E has provided six GHe storage vessels and QST has provided the complete infrastructure and the facilities for the utilities. The paper gives an overview of the main design features, the infrastructure and the status of installation and pre-commissioning.

Design and characteristics of the drive mechanism for movable limiters of JT-60, (1)

International Nuclear Information System (INIS)

Takashima, Tetsuo; Morishita, Osamu; Yamamoto, Masahiro; Shimizu, Masatsugu; Ohta, Mitsuru

1976-10-01

Two fast-acting movable rail limiters will be installed in a large Tokamak JT-60 being designed in JAERI. The movable limiter consists of a drive mechanism, a vacuum seal, a bearing, and a molybdenum rail limiter. Design of the drive mechanism for the movable limiter and experimental results on the driving characteristics in full scale are described. (auth.)

Design of a negative-ion based NBI system for JT-60U

International Nuclear Information System (INIS)

Kuriyama, M.; Araki, M.; Inoue, T.; Kunieda, S.; Matsuoka, M.; Mizuno, M.; Ohara, Y.; Okumura, Y.; Oohara, H.; Watanabe, K.

1992-01-01

This paper reports on a negative-ion based NBI system which is planned as a key device on the JT-60U in the experiments of current drive and plasma core heating with high density plasmas. The NBI system will inject neutral beams of 500keV, 10MW for 10sec from a beamline with two ion sources. The neutral beam will be injected tangentially in the codirection. Each ion source is a modified volume production-type negative-ion source with cesium vapor. The acceleration current is 22A with deuterium beam, and the current density is 13mA/cm 2 . An operational pressure in the negative-ion generator is less than 0.5 Pa. A three-stage electro static acceleration system is adopted as the accelerator. The beamline length between the ion source and the injection port is 24m. The beamline consists of an ion source tank, neutralizer cells of 10m in length, an ion dump tank and a drift duct. The ion source tank contains large cryopumps to maintain the exit of the ion source sufficiently low. The ion dump tank contains ion deflecting coils, ion dumps for positive and negative ions, a calorimeter, cryopumps and beam scrapers. Residual ions are deflected by the combined magnetic fields produced by the deflecting coils and the stray field form the tokamak. The two sources are connected to an acceleration power supply of 500kV/64A/10sec, while the negative-ion generator power, the extraction voltage, and electron-suppression voltage are fed individually

Fast plasma shutdown by killer pellet injection in JT-60U with reduced heat flux on the divertor plate and avoiding runaway electron generation

International Nuclear Information System (INIS)

Yoshino, R.; Kondoh, T.; Neyatani, Y.; Itami, K.; Kawano, Y.; Isei, N.

1997-01-01

A killer pellet is an impurity pellet that is injected into a tokamak plasma in order to terminate a discharge without causing serious damage to the tokamak machine. In JT-60U neon ice pellets have been injected into OH and NB heated plasmas and fast plasma shutdowns have been demonstrated without large vertical displacement. The heat pulse on the divertor plate has been greatly reduced by killer pellet injections (KPI), but a low-power heat flux tail with a long time duration is observed. The total energy on the divertor plate increases with longer heat flux tail, so it has been reduced by shortening the tail. Runaway electron (RE) generation has been observed just after KPI and/or in the later phase of the plasma current quench. However, RE generation has been avoided when large magnetic perturbations are excited. These experimental results clearly show that KPI is a credible fast shutdown method avoiding large vertical displacement, reducing heat flux on the divertor plate, and avoiding (or minimizing) RE generation. (Author)

Stationary high confinement plasmas with large bootstrap current fraction in JT-60U

International Nuclear Information System (INIS)

Sakamoto, Y.; Fujita, T.; Ide, S.; Isayama, A.; Takechi, M.; Suzuki, T.; Takenaga, H.; Oyama, N.; Kamada, Y.

2005-01-01

This paper reports the results of the progress in stationary discharges with a large bootstrap current fraction in JT-60U towards steady-state tokamak operation. In the weak shear plasma regime, high-β p ELMy H-mode discharges have been optimized under nearly full non-inductive current drive conditions by the large bootstrap current fraction (f BS ∼ 45%) and the beam driven current fraction (f BD ∼ 50%), which was sustained for 5.8 s in the stationary condition. This duration corresponds to ∼26τ E and ∼2.8τ R , which was limited by the pulse length of negative-ion-based neutral beams. The high confinement enhancement factor H 89 ∼ 2.2 (HH 98y2 ∼ 1.0) was obtained and the profiles of current and pressure reached the stationary condition. In the reversed shear plasma regime, a large bootstrap current fraction (f BS ∼ 75%) has been sustained for 7.4 s under nearly full non-inductive current drive conditions. This duration corresponds to ∼16τ E and ∼2.7τ R . The high confinement enhancement factor H 89 ∼ 3.0 (HH 98y2 ∼ 1.7) was also sustained, and the profiles of current and pressure reached the stationary condition. The large bootstrap current and the off-axis beam driven current sustained this reversed q profile. This duration was limited only by the duration of the neutral beam injection

Characteristics of halo current in JT-60U

International Nuclear Information System (INIS)

Neyatani, Y.; Nakamura, Y.; Yoshino, R.; Hatae, T.

1999-01-01

Halo currents and their toroidal peaking factor (TPF) have been measured in JT-60U by Rogowski coil type halo current sensors. The electron temperature in the halo region was around 10 eV at 1 ms before the timing of the maximum halo current. The maximum TPF*I h /I p0 was 0.52 in the operational range of I p = 0.7 ∼ 1.8 MA, B T = 2.2 ∼ 3.5 T, including ITER design parameters of κ > 1.6 and q 95 = 3, which was lower than that of the maximum value of ITER data base (0.75). The magnitude of halo currents tended to decrease with the increase in stored energy just before the energy quench and with the line integrated electron density at the time of the maximum halo current. A termination technique in which the current channel remains stationary was useful to avoid halo current generation. Intense neon gas puffing during the VDE was effective for reducing the halo currents. (author)

Characteristics of halo current in JT-60U

International Nuclear Information System (INIS)

Neyatani, Y.; Nakamura, Y.; Yoshino, R.; Hatae, T.

2001-01-01

Halo currents and their toroidal peaking factor (TPF) have been measured in JT-60U by Rogowski coil type halo current sensors. The electron temperature in the halo region was around 10 eV at 1 ms before the timing of the maximum halo current. The maximum TPF *I h /I p0 was 0.52 in the operational range of I p =0.7∼1.8MA, B T =2.2∼3.5T, including ITER design parameters of κ>1.6 and q 95 =3, which was lower than that of the maximum value of ITER data base (0.75). The magnitude of halo currents tended to decrease with the increase in stored energy just before the energy quench and with the line integrated electron density at the time of the maximum halo current. A termination technique in which the current channel remains stationary was useful to avoid halo current generation. Intense neon gas puffing during the VDE was effective for reducing the halo currents. (author)

Deuterated-decaborane using boronization on JT-60U

Energy Technology Data Exchange (ETDEWEB)

Yagyu, Jun-ichi; Arai, Takashi; Kaminaga, Atsushi; Miyata, Katsuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Arai, Masaru [Kaihatsu Denki Co., Ltd., Tokyo (Japan)

2001-03-01

In JT-60U, boronization using hydride-decaborane (B{sub 10}H{sub 14}) vaporization has been conducted for the first wall conditioning. Compared to other discharge cleaning (DC), boronization is claimed to be efficient in reduction of oxygen impurities and hydrogen recycling in plasma. However, there are some problems in reduction of hydrogen included in boron film and stabilization of DC glow discharge during the boronization. To solve these problems, a new boronization method using deuterated-decaborane (B{sub 10}D{sub 14}) was adopted instead of the conventional hydride-decaborane. As a result, hydrogen content in the boron film decreased clearly and discharge conditioning shots, for decreasing hydrogen content in plasmas, after the boronization were reduced to 1/10 in comparison to the conventional process. Furthermore, DC glow discharge became stable, with only helium carrier gas, and it was possible to save 30 hours in maximum of the time necessary to boronization. It is shown that the boronization using deuterated-decaborane is very efficient and effective method for the first wall conditioning. (author)

Tritium distribution on plasma facing graphite tiles of JT-60U

International Nuclear Information System (INIS)

Tanabe, T.; Sugiyama, K.; Masaki, K.; Gotoh, Y.; Tobita, K.; Miya, N.

2003-01-01

Tritium distributions on the graphite divertor tiles, the dome units and the baffle plates of JT-60U were successfully measured. Poloidally, the highest tritium level was found at the dome top tiles and the outer baffle plates, where the plasma did not hit directly. On the other hand, although the toroidal tritium profiles on each tiles appeared uniform, detailed profiles in full toroidal direction clearly showed a periodic variation corresponding to the position of the magnetic field coils, indicating the ripple loss of high energy tritons as suggested by the OFMC code. Finally, the temperature increase owing to the plasma heat load was found to release the once retained tritium. (author)

Disruption generated secondary runaway electrons in present day tokamaks

International Nuclear Information System (INIS)

Pankratov, I.M.; Jaspers, R.

2000-01-01

An analysis of the runaway electron secondary generation during disruptions in present day tokamaks (JET, JT-60U, TEXTOR) was made. It was shown that even for tokamaks with the plasma current I approx 100 kA the secondary generation may dominate the runaway production during disruptions. In the same time in tokamaks with I approx 1 MA the runaway electron secondary generation during disruptions may be suppressed

Plasma regimes and research goals of JT-60SA towards ITER and DEMO

International Nuclear Information System (INIS)

Kamada, Y.; Ide, S.; Fujita, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; Barabaschi, P.; Lackner, K.; Ishida, S.; Bolzonella, T.

2011-01-01

The JT-60SA device has been designed as a highly shaped large superconducting tokamak with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) in order to satisfy the central research needs for ITER and DEMO. In the ITER- and DEMO-relevant plasma parameter regimes and with DEMO-equivalent plasma shapes, JT-60SA quantifies the operation limits, plasma responses and operational margins in terms of MHD stability, plasma transport and confinement, high-energy particle behaviour, pedestal structures, scrape-off layer and divertor characteristics. By integrating advanced studies in these research fields, the project proceeds 'simultaneous and steady-state sustainment of the key performances required for DEMO' with integrated control scenario development applicable to the highly self-regulating burning high-β high bootstrap current fraction plasmas.

Particle transport analysis in lower hybrid current drive discharges of JT-60U

International Nuclear Information System (INIS)

Nagashima, K.; Ide, S.; Naito, O.

1996-01-01

Particle transport is modified in lower hybrid current drive discharges of JT-60U. The density profile becomes broad during the lower hybrid wave injection and the profile change depends on the injected wave spectrum. Particle transport coefficients (diffusion coefficient and profile peaking factor) were evaluated using gas-puff modulation experiments. The diffusion coefficient in the current drive discharges is about three times larger than in the ohmic discharges. The profile peaking factor decreases in the current drive discharges and the evaluated values are consistent with the measured density profiles. (author)

Long-term erosion and re-deposition of carbon in the divertor region of JT-60U

International Nuclear Information System (INIS)

Gotoh, Y.; Tanabe, T.; Ishimoto, Y.; Masaki, K.; Arai, T.; Kubo, H.; Tsuzuki, K.; Miya, N.

2006-01-01

Erosion and redeposition profiles of carbon tiles used in the W-shaped divertor of JT-60U with all carbon plasma facing wall are studied. The inner divertor is mostly covered by carbon redeposited layers, while the outer divertor mostly eroded. In the dome region, the erosion dominates on the inner dome-wing, while redeposited on the outer dome-wing. The redeposited layers on the outer dome-wing show very clear columnar structures indicating local carbon transport form the outer divertor to the outer dome-wing. The weight gain by the redeposition extrapolated to the whole divertor area is 0.55 kg. Since the extrapolated total erosion is about 0.33 kg, the remaining 0.22 kg must be originated from the main chamber erosion. Significant amount of the redeposition is caused locally by multiple processes of erosion, ionization and prompt redeposition toward inboard direction owing to gyration along magnetic filed line. This inboard transport is one of the reasons for small redeposition on the plasma shadowed area of the W shaped divertor of JT-60U with pumping slots placed at the bottoms side

Energy confinement in JT-60 lower hybrid current driven plasmas

International Nuclear Information System (INIS)

Ushigusa, K.; Imai, T.; Naito, O.; Ikeda, Y.; Tsuji, S.; Uehara, K.

1990-01-01

The energy confinement in high power lower hybrid current driven (LHCD) plasmas has been studied in the JT-60 tokamak. At a plasma current of 1 MA, the diamagnetically estimated energy confinement time in LHCD plasmas has almost the same value as the confinement time in ohmically heated plasmas at n-bar e ∼ 1.0x10 19 m -3 . The confinement time of high power LHCD plasmas (P LH E varies as to P LH α n e β I p 0 with α + β ∼ -0.3. (author). Letter-to-the-editor. 12 refs, 5 figs

Heat structural problems in JT-60

International Nuclear Information System (INIS)

Takatsu, Hideyuki; Shimizu, Masaomi; Yamamoto, Masahiro; Nakamura, Hiroo; Miyauchi, Yasuyuki.

1980-01-01

The construction of JT-60 is in progress to study the behavior of hydrogen plasma. The D-T reaction does not occur in this device, therefore the considerations for neutron damage, tritium leakage and so on are not necessary. The long-pulse operation will be done, and the suppression of the production and mixing of impurity is considered in the design of the JT-60. The high temperature baking is possible, and the magnetic limiter is set. The vacuum container has the complex structure consists of 8 sector type thick rings and 8 U-shaped bellows, and has egg-shaped cross section. The main radius of the torus is about 3 m. The material of the vacuum container is INCONEL 625. The analyses of various stresses due to such as atmospheric pressure, eddy current and thermal expansion were made. It is also necessary to consider the thermal stress due to the leakage of neutral beam. The thermal input of about 20 MW per one discharge to the first wall is taken into consideration. The material of the first wall is molybdenum. (Kato, T.)

Review of JT-60U experimental results in 2003 and 2004

International Nuclear Information System (INIS)

2006-01-01

The results from the JT-60U experiments in 2003 and 2004 are reviewed. The steady-state advanced tokamak (AT) research has progressed with emphasis on long sustainment of high plasma parameters in comparison with the characteristic time scales, such as for current relaxation and variation on plasma-wall interaction. To achieve this, modification of the controls for the operation, heating and diagnostics systems have been done. As a result, ∼60 s current flat top at I p =0.7 MA and ∼30 s H-mode with I p =1.4 MA were obtained. The long pulse modification has opened a door into a new domain for JT-60U. High normalized beta (β N ) of 2.3 was maintained for 22.3 s in a high-β p H-mode plasma, which corresponds to 13.1 times current relaxation time (τ R ). A higher β N of 2.5 was also maintained for 16.5 s. In these plasmas, the confinement degradation due to increase in wall recycling on the longer time scale limited the sustainable duration for high β N , as well as reduction of NB heating power due to heat load limitation at the armor tiles. The change in the wall recycling was quantitatively investigated in long-pulse standard ELMy H-mode plasmas. Particle balance analysis indicated that the particle inventory at the first wall was globally saturated (global wall saturation) in the latter phase of the discharge by repeating several long pulse discharges. Development and investigation of the AT plasmas has been continued within the previous heating duration (≤ 10 s) as well, where higher NB power is available. Higher β N (∼3) was maintained for 6.2 s in high-β p H-mode plasmas in low q 95 regime. A stationary high-β p ELMy H-mode plasma was sustained for 5.8 s with high confinement of HH 98y2 - 1.0 (HH 98y2 is the confinement enhancement factor over the IPB98(y, 2) scaling) under nearly full non-inductive current driven by the large bootstrap current fraction (f BS - 45%) and the beam driven current fraction (f BD - 50%). This duration corresponds to �

JT-60U high performance regimes

International Nuclear Information System (INIS)

Ishida, S.

1999-01-01

High performance regimes of JT-60U plasmas are presented with an emphasis upon the results from the use of a semi-closed pumped divertor with W-shaped geometry. Plasma performance in transient and quasi steady states has been significantly improved in reversed shear and high- βp regimes. The reversed shear regime elevated an equivalent Q DT eq transiently up to 1.25 (n D (0)τ E T i (0)=8.6x10 20 m-3·s·keV) in a reactor-relevant thermonuclear dominant regime. Long sustainment of enhanced confinement with internal transport barriers (ITBs) with a fully non-inductive current drive in a reversed shear discharge was successfully demonstrated with LH wave injection. Performance sustainment has been extended in the high- bp regime with a high triangularity achieving a long sustainment of plasma conditions equivalent to Q DT eq ∼0.16 (n D (0)τ E T i (0)∼1.4x10 20 m -3 ·s·keV) for ∼4.5 s with a large non-inductive current drive fraction of 60-70% of the plasma current. Thermal and particle transport analyses show significant reduction of thermal and particle diffusivities around ITB resulting in a strong Er shear in the ITB region. The W-shaped divertor is effective for He ash exhaust demonstrating steady exhaust capability of τ He */τ E ∼3-10 in support of ITER. Suppression of neutral back flow and chemical sputtering effect have been observed while MARFE onset density is rather decreased. Negative-ion based neutral beam injection (N-NBI) experiments have created a clear H-mode transition. Enhanced ionization cross- section due to multi-step ionization processes was confirmed as theoretically predicted. A current density profile driven by N-NBI is measured in a good agreement with theoretical prediction. N-NBI induced TAE modes characterized as persistent and bursting oscillations have been observed from a low hot beta of h >∼0.1-0.2% without a significant loss of fast ions. (author)

Physics of strong internal transport barriers in JT-60U reversed-magnetic-shear plasmas

Energy Technology Data Exchange (ETDEWEB)

Hayashi, N; Takizuka, T; Sakamoto, Y; Fujita, T; Kamada, Y; Ide, S; Koide, Y [Japan Atomic Energy Agency, 801-1 Mukouyama, Naka, Ibaraki 311-0193 (Japan)

2006-05-15

The physics of strong internal transport barriers (ITBs) in JT-60U reversed-magnetic-shear (RS) plasmas has been studied through the modelling on the 1.5 dimensional transport simulation. The key physics to produce two scalings on the basis of the JT-60U box-type ITB database are identified. As for the scaling for the narrow ITB width proportional to the ion poloidal gyroradius, the following three physics are important: (1) the sharp reduction of the anomalous transport below the neoclassical level in the RS region, (2) the autonomous formation of pressure and current profiles through the neoclassical transport and the bootstrap current and (3) the large difference between the neoclassical transport and the anomalous transport in the normal-shear region. As for the scaling for the energy confinement inside ITB ({epsilon}{sub f}{beta}{sub p,core} {approx} 0.25, where {epsilon}{sub f} is the inverse aspect ratio at the ITB foot and {beta}{sub p,core} is the core poloidal beta value), the value of 0.25 is found to be a saturation value due to the MHD equilibrium. The value of {epsilon}{sub f}{beta}{sub p,core} reaches the saturation value, when the box-type ITB is formed in the strong RS plasma with a large asymmetry of the poloidal magnetic field, regardless of the details of the transport and the non-inductively driven current.

Design study of a negative-ion based NBI system for JT-60U

International Nuclear Information System (INIS)

Akino, Noboru; Araki, Masanori; Ebisawa, Noboru

1994-03-01

A high energy negative-ion based NBI system for JT-60U has been designed. The objective of the NBI system is to demonstrate mega-ampere level NB current drive and plasma core heating in a reactor-grade high density plasma. This is the first negative-ion based NBI system in the world. The required specifications of the NBI system are; a beam energy of 500 keV, an injection power of 10 MW, a beam pulse duration of 10 sec with a duty cycle of 1/60 and a beam species of deuterium or hydrogen. The neutral beam power of 10 MW is injected tangentially using one beam-line with two large negative-ion sources. The construction of the NBI system has been started, and will be operational in 1996. (author)

Conceptual design of divertor cassette handling by remote handling system for JT-60SA

International Nuclear Information System (INIS)

Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

2007-01-01

The JT-60SA aims to contribute and supplement ITER toward DEMO reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is prohibited. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor modules. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor module, which is 10 degrees wide in toroidal direction and weighs 500kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor module to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the module by a pallet installed from outside the VV. (author)

Conceptual design of divertor cassette handling by remote handling system of JT-60SA

International Nuclear Information System (INIS)

Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

2008-01-01

The JT-60SA aims to contribute and supplement ITER toward demonstration fusion reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is restricted. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor cassettes. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor cassette, which is 10 degrees wide in toroidal direction and weighs 500 kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor cassette to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the cassette by a pallet installed from outside the VV. (author)

Impact of arcing on carbon and tungsten. From the observations in JT-60U, LHD, and NAGDIS-II

International Nuclear Information System (INIS)

Kajita, Shin; Fukumoto, Masakatsu; Nakano, Tomohide; Tokitani, Masayuki; Masuzaki, Suguru; Ohno, Noriyasu; Takamura, Shuichi; Yoshida, Naoaki; Ueda, Yoshio

2012-11-01

This paper assesses the impact of arcing in fusion devices based on the observations in JT-60U, LHD, and the linear divertor simulator NAGDIS-II. In NAGDIS-II, the demonstration experiments of arcing/unipolar arcing have been conducted by simulating the transient heat load using a pulsed laser; it was found that the arcing can be easily initiated on the helium irradiated nanostructured tungsten. By measuring the field emission current property from the helium irradiated tungsten surface, the initiation conditions are discussed. From the detailed analysis of JT-60U tiles, it is found that arcing phenomena occurred on carbon baffle plates. From the observation of the arc trails recorded on the baffle plate, the amount of the eroded materials is discussed. The arcing seemed to occur frequently on inner baffles rather than the outer baffles. From LHD, it is shown that the arcing can be initiated on nanostructured tungsten even without transient events. The erosion of tungsten by arcing will become an important issue in a fusion reactor, where helium fluence is significantly increased. (author)

Baking technique of JT-60

Energy Technology Data Exchange (ETDEWEB)

Arai, Takashi; Masaki, Kei; Miyachi, Kengo [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1998-10-01

It needs to make low ultimate pressure to decrease impurities in a plasma in a nuclear fusion device. Baking technique is very effective method to decrease outgassing rate from first walls and inner surface of the vacuum vessel. However, in such large vacuum vessel, e.g. JT-60, the non-uniform heating and the deformation due to thermal expansion might be very important problems. The baking technique of JT-60 is described. Two baking systems were applied to JT-60 to solve the problem of non-uniform heating. One is a circulation system of the hot nitrogen gas which is located between the inner and outer skins of the double-skin structured vacuum vessel. The other is an electric heater which apply the thick of the vessel. In order to prevent the deformation, the difference of temperature was as small as possible. By both the baking at 300degC and the conditioning such as discharge cleaning, the ultimate pressure was achieved 7.9x10{sup -7} Pa. (author)

Baking technique of JT-60

International Nuclear Information System (INIS)

Arai, Takashi; Masaki, Kei; Miyachi, Kengo

1998-01-01

It needs to make low ultimate pressure to decrease impurities in a plasma in a nuclear fusion device. Baking technique is very effective method to decrease outgassing rate from first walls and inner surface of the vacuum vessel. However, in such large vacuum vessel, e.g. JT-60, the non-uniform heating and the deformation due to thermal expansion might be very important problems. The baking technique of JT-60 is described. Two baking systems were applied to JT-60 to solve the problem of non-uniform heating. One is a circulation system of the hot nitrogen gas which is located between the inner and outer skins of the double-skin structured vacuum vessel. The other is an electric heater which apply the thick of the vessel. In order to prevent the deformation, the difference of temperature was as small as possible. By both the baking at 300degC and the conditioning such as discharge cleaning, the ultimate pressure was achieved 7.9x10 -7 Pa. (author)

Investigation and measures to noise on spectroscopic measurement system in JT-60U

International Nuclear Information System (INIS)

Nagaya, Susumu; Kubo, Hirotaka; Sugie, Tatsuo; Onizawa, Masami; Kawai, Isao; Nakata, Hisao.

1997-11-01

Breakdown of a negative-ion-based neutral beam injection (N-NBI) has caused noise trouble to several systems. The control circuit of a spectroscopic measurement system had not well worked because of the noise. The noise has been measured by an optical-fiber isolation system during operation of JT-60U. The amplitude and the frequency were 15-18 V and 15 MHz respectively. The transmission noise has been reduced by putting ferrite cores to all cables connecting with the control circuits. As a result, the trouble with the spectroscopic measurement system has completely been solved. Adding condensers and resistors to the circuit was not effective to reduce the noise. (author)

Studies on structural analysis related to the design of the JT-60 vacuum vessel

International Nuclear Information System (INIS)

Takatsu, Hideyuki

1987-06-01

Studies on structural analysis of a vacuum vessel of tokamak-type fusion devices are presented. The present studies are proposals for the structural analysis procedures of the tokamak-type fusion devices and are composed of five parts, each of which covers the fundamental area required for the structural analysis and design; stress analysis, dynamic response analysis, fatigue evaluation, buckling analysis and seismic analysis. Special attention is paid to the critical component, bellows and the critical load, electromagnetic forces. A new finite element method modeling technique is proposed for the stress analysis of U-shaped bellows, where the bellows is replaced by an orthotropic plate having the same stiffness as the bellows. The applicability of the present modeling technique is confirmed by verification tests. Dynamic response and fatigue of the vacuum vessel are critical issues of the structural analysis and design of the tokamak-type fusion devices. Detailed dynamic response analyses of the JT-60 vacuum vessel are presented paying special attention to the dynamic behavior of the U-shaped bellows, where the above-mentioned modeling technique of the U-shaped bellows is applied. A fatigue evaluation method of the vacuum vessel under the dynamic electromagnetic forces is proposed, which utilizes the results of the detailed dynamic response analysis. In the present method, fatigue evaluation method for random loads is applied. Torsional fatigue strength of the welded bellows is experimentally evaluated aiming the application to the port of the fusion device and it is shown that the welded bellows reveals elastic buckling and spiral distortion under a small angle of tortion. Two formulae are proposed to evaluate the stress of the welded bellows under the forced angle of tortion. (author)

Qualification and preparatory activities for the manufacturing of 9 TF coils of the JT-60SA magnet

International Nuclear Information System (INIS)

Cucchiaro, Antonio; Polli, Gian Mario; Cocilovo, Valter; Drago, Giovanni; Cuneo, Stefano; Terzi, Franco; Peyrot, Marc; Phillips, Guy; Tomarchio, Valerio

2013-01-01

In the framework of the Broader Approach Agreement for the construction of the JT-60SA tokamak, ENEA is in charge to provide 9 of the 18 Toroidal Field (TF) coils. The 9 coils are being manufactured by ASG superconductors in Genoa under the supervision of ENEA in collaboration with the JT-60SA European home team. Prior the manufacturing, a preparatory activity has been carried out aimed at designing, constructing and setting-up the relevant components to be realized. In order to get confidence of some special manufacturing process, several qualification activities have been performed. In this paper an overview of the principal milestones reached during the preparatory phase and a description of the qualification activities with relevant test results are presented

Surface studies of tungsten erosion and deposition in JT-60U

International Nuclear Information System (INIS)

Ueda, Y.; Fukumoto, M.; Nishikawa, M.; Tanabe, T.; Miya, N.; Arai, T.; Masaki, K.; Ishimoto, Y.; Tsuzuki, K.; Asakura, N.

2007-01-01

In order to study tungsten erosion and migration in JT-60U, 13 W tiles have been installed in the outer divertor region and tungsten deposition on graphite tiles was measured. Dense local tungsten deposition was observed on a CFC tile toroidally adjacent to the W tiles, which resulted from prompt ionization and short range migration of tungsten along field lines. Tungsten deposition with relatively high surface density was found on an inner divertor tile around standard inner strike positions and on an outer wing tile of a dome. On the outer wing tile, tungsten deposition was relatively high compared with carbon deposition. In addition, roughly uniform tungsten depth distribution near the upper edge of the inner divertor tile was observed. This could be due to lift-up of strike point positions in selected 25 shots and tungsten flow in the SOL plasma

Characteristics of internal transport barrier in JT-60U reversed shear plasmas

International Nuclear Information System (INIS)

Sakamoto, Y.; Kamada, Y.; Ide, S.; Fujita, T.; Shirai, H.; Takizuka, T.; Koide, Y.; Fukuda, T.; Oikawa, T.; Suzuki, T.; Shinohara, K.; Yoshino, R.

2001-01-01

Characteristics of internal transport barrier (ITB) structure are studied and the active ITB control has been developed in JT-60U reversed shear plasmas. The following results are found. Outward propagation of the ITB with steep T i gradient is limited to the minimum safety factor location (ρ qmin ). However the ITB with reduced T i gradient can move to the outside of ρ qmin . Lower boundary of ITB width is proportional to the ion poloidal gyroradius at the ITB center. Furthermore the demonstration of the active control of the ITB strength based on the modification of the radial electric field shear profile is successfully performed by the toroidal momentum injection in different directions or the increase of heating power by neutral beams. (author)

Low energy, high power injection in JT-60 NBI

International Nuclear Information System (INIS)

Mizuno, Makoto; Dairaku, Masayuki; Horiike, Hiroshi

1988-05-01

JT-60 neutral beam injector (JT-60 NBI) is designed to inject 20 MW neutral hydrogen beam at energies of 70 ∼ 100 keV and the injection power decreases significantly at low energies (∼40 keV). For the extention of operation region aiming at the low density plasma heating and achieving H-mode by plasma periphery heating, increment of the injection power at low beam energies was required. The single-stage acceleration system was investigated in advance at the Prototype Injector Unit. From this result, the total injection power of 17 MW at 40 keV, 48 A per source was expected at the JT-60 NBI. This system was adopted in the JT-60 NBI from June, 1987 to July, 1987 and 17.6 MW neutral beam injection power was achieved. In the NB heating experiment, the H-mode transition phenomena was observed in JT-60 plasma. (author)

Control of divertor configuration in JT-60

International Nuclear Information System (INIS)

Yoshino, R.; Kukuchi, M.; Ninomiya, H.; Yoshida, H.; Tsuji, S.; Hosogane, N.; Seki, S.

1985-01-01

The control algorithm of JT-60 divertor configuration is presented. JT-60 has five types of poloidal magnetic field coil with each power supply in order to regulate the control objectives mentioned above. However, if one controls each objective by each coil current independently, there must inevitably occur large interaction between control objectives. Because the relation between control objectives and coil currents is complicated. This situation may be the same with a fusion reactor device. For making it possible to control each objective independently without causing large interaction, the authors adopt the noninteracting control algorithm. Hence, this report demonstrates the availability of this method to the control of JT-60 divertor configuration

Neural-net predictor for beta limit disruptions in JT-60U

International Nuclear Information System (INIS)

Yoshino, R.

2005-01-01

Prediction of major disruptions occurring at the β -limit for tokamak plasmas with a normal magnetic shear in JT-60U was conducted using neural networks. Since no clear precursors are generally observed a few tens of milliseconds before the β -limit disruption, a sub-neural network is trained to output the value of the β N limit every 2 ms. The target β N limit is artificially set by the operator in the first step to train a network with non-disruptive shots as well as disruptive shots, and then in the second step the target limit is modified using the β N limit output from the trained network. The adjusted target greatly improves the consistency between the input data and the output. This training, the 'self-teaching method', has greatly reduced the false alarm rate triggered for non-disruptive shots. To improve the prediction performance further, the difference between the output β N limit and the measured β N , and 11 parameters, are inputted to the main neural network to calculate the 'stability level'. The occurrence of a major disruption is predicted when the stability level decreases to the 'alarm level'. Major disruptions at the β -limit have been predicted by the main network with a prediction success rate of 80% at 10 ms prior to the disruption while the false alarm rate is lower than 4% for non-disruptive shots. This 80% value is much higher than that obtained for a network trained with a fixed target β N limit set to be the maximum β N observed at the start of a major disruption, lower than 10%. A prediction success rate of 90% with a false alarm rate of 12% at 10 ms prior to the disruption has also been obtained. This 12% value is about half of that obtained for a network trained with a fixed target β N limit

An overview on plasma disruption mitigation and avoidance in tokamak

International Nuclear Information System (INIS)

He Kaihui; Pan Chuanhong; Feng Kaiming

2002-01-01

Plasma disruption, which seems to be unavoidable in Tokamak operation, occurs very fast and uncontrolled. In order to keep Tokamak plasma from disruption and mitigate the disruption frequency, the research on Tokamak plasma major disruption constitutes one of the main topics in plasma physics. The phenomena and processes of the precursor, thermal quench, current quench, VDE, halo current and runaway electrons generation during plasma disruption are analyzed in detail and systematically based on the data obtained from current Tokamaks such as TFTR, JET, JT-60U and ASDEX-U, etc. The methods to mitigate and avoid disruption in Tokamak are also highlighted schematically. Therefore, it is helpful and instructive for plasma disruption research in next generation large Tokamak such as ITER-FEAT

Simulation of collisional effects on divertor pumping in JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Gleason-González, C., E-mail: cristian.gleason@kit.edu [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, Baden-Württemberg 76344 (Germany); Varoutis, S.; Luo, X. [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, Baden-Württemberg 76344 (Germany); Shimizu, K.; Nakano, T.; Hoshino, K.; Kawashima, H.; Asakura, N. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka, Ibaraki 311-0193 (Japan); Day, Chr. [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, Baden-Württemberg 76344 (Germany); Sakurai, S. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2016-11-01

Highlights: • The exhausted (sub-divertor) gas flows calculations in tokamaks by means of three approaches: ProVac3D, DIVGAS and NEUT2D. • Exhausted neutral gas flow is modeled for two scenarios of a simplified JT-60SA sub-divertor geometry. • A modelled scenario with a simplified setup is done by using two intrinsic collisionless approaches: ProVac3D and NEUT2D and a third approach: DIVGAS, which has been used without its collision module for a direct comparison with the other two. The solvers are cross-checked in terms of the reproduction of the transmission probability. • A second case study is based on the Scenario # 2 of JT-60SA, where the assessment of collisionality in the sub-divertor was done. The gas flow is simulated by using DIVGAS with and without consideration of collisions. • The results include the transmission probability in JT-60SA sub-divertor, the Knudsen number, which characterizes the collisionality of the flow, velocity profiles, pressure and temperature distributions. - Abstract: In this work, the exhausted neutral gas flow is modeled for two cases of a simplified JT-60SA sub-divertor geometry and compared via three different approaches, namely (i) a collisionless approach based on the ProVac3D code, (ii) the DSMC approach based on the DIVGAS code that can be run with and without consideration of particle collisions, and (iii) the NEUT2D approach which has been extensively used in the past for the JT-60 design. In a first case study, the transmission probability was calculated by the 3 approaches and very good agreement is found between NEUT2D-ProVac3D whereas discrepancies between DIVGAS and NEUT2D are found and further analyzed. In the second case, the assessment of collisions is done by means of DIVGAS. Simulations showed that the flow is found in the transitional regime with Kn numbers between 0.1 and 0.4. The DIVGAS collisionless case yielded lower values of temperature than the collisional one by factors of 0.5–0.8 in

Report on the design of JT-4

International Nuclear Information System (INIS)

Kitsunezaki, Akio; Seki, Shogo; Yokomizo, Hideaki; Matsuda, Toshiaki; Saito, Ryuta

1978-08-01

The present status of design of JT-4 tokamak is described. The objectives of JT-4 are shown graphically and the main parameters are tabulated. JT-4 is a tokamak of non-circular (ellipse or D-shape) plasma cross section with axisymmetric divertors at top and bottom of the plasma column. The principal purpose of JT-4 is to obtain high plasma beta values, desirably exceeding 5%, by strong secondary plasma heating and by impurities elimination. The experimental results obtained with JT-4 are essential in the design of future tokamaks and tokamak reactors with high efficiency and at reasonable cost. (author)

Quench propagation and quench detection in the TF system of JT-60SA

International Nuclear Information System (INIS)

Lacroix, Benoit; Duchateau, Jean-Luc; Meuris, Chantal; Ciazynski, Daniel; Nicollet, Sylvie; Zani, Louis; Polli, Gian-Mario

2013-01-01

Highlights: • The JT-60SA primary quench detection system will be based on voltage measurements. • The early quench propagation was studied in the JT-60SA TF conductor. • The impact of the conductor jacket on the hot spot criterion was quantified. • The detection parameters were investigated for different quench initiations. -- Abstract: In the framework of the JT-60SA project, France and Italy will provide to JAEA 18 Toroidal Field (TF) coils including NbTi cable-in-conduit conductors. During the tokamak operation, these coils could experience a quench, an incidental event corresponding to the irreversible transition from superconducting state to normal resistive state. Starting from a localized disturbance, the normal zone propagates along the conductor and dissipates a large energy due to Joule heating, which can cause irreversible damages. The detection has to be fast enough (a few seconds) to trigger the current discharge, so as to dump the stored magnetic energy into an external resistor. The JT-60SA primary quench detection system will be based on voltage measurements, which are the most rapid technology. The features of the detection system must be adjusted so as to detect the most probable quenches, while avoiding inopportune fast safety discharges. This requires a reliable simulation of the early quench propagation, performed in this study with the Gandalf code. The conductor temperature reached during the current discharge must be kept under a maximal value, according to the hot spot criterion. In the present study, a hot spot criterion temperature of 150 K was taken into account and the role of each conductor component (strands, helium and conduit) was analyzed. The detection parameters were then investigated for different hypotheses regarding the quench initiation

Acceleration mechanism of vertical displacement event and its amelioration in tokamak disruptions

International Nuclear Information System (INIS)

Nakamura, Yukiharu; Yoshino, Ryuji; Pomphrey, N.; Jardin, S.C.

1996-01-01

Vertical displacement events (VDEs), which are frequently observed in disruptive discharges of elongated tokamaks, are investigated using the Tokamak Simulation Code. We show that disruption events such as a sudden plasma pressure drop (β p collapse) and the subsequent plasma current quench (I p quench) can accelerate VDEs due to the adverse destabilizing effect of the resistive shell, which has previously been thought to stabilize VDEs. In a tokamak with a surrounding shell which is asymmetric with respect to the geometric midplane, the I p quench also causes an additional VDE acceleration due to the vertical imbalance of the attractive force. While the shell-geometry characterizes the VDE dynamics, the growth rate of VDEs depends strongly on the magnitude of the β p collapse, the speed of the I p quench and the n-index of the plasma equilibrium just before the disruption. An amelioration of I p quench-induced VDEs was experimentally established in the JT-60U tokamak by optimizing the vertical location of the plasma just prior to the disruption. The JT-60U vacuum vessel is shown to be suitable for preventing the β p collapse-induced VDE. (author)

Localized MHD activity near transport barriers in JT-60U and TFTR

International Nuclear Information System (INIS)

Manickam, J.

2001-01-01

Localized MHD activity observed in JT-60U and TFTR near transport barriers with their associated large pressure gradients is investigated. Stability analysis of equilibria modeling the experiments supports an identification of this MHD as being due to an ideal MHD n=1 instability. The appearance of the instability depends on the local pressure gradient, local shear in the q profile and the proximity of rational surfaces where q∼m/n and m and n are the poloidal and toroidal mode numbers respectively. The mode width is shown to depend on the local value of q, and is larger when q is smaller. In addition the role of the edge current density in coupling the internal mode to the plasma edge and of the energetic particles which can drive fishbone like modes is investigated. (author)

Application of diamond window for infrared laser diagnostics in a tokamak device

International Nuclear Information System (INIS)

Kawano, Yasunori; Chiba, Shinichi; Inoue, Akira

2004-01-01

Chemical vapor deposited diamond disks have been successfully applied as the vacuum windows for infrared CO 2 laser interferometry and polarimetry used in electron density measurement in the JT-60U tokamak. In comparison with the conventional zinc-selenide windows, the Faraday rotation component of diamond windows was negligible. This results in an improvement of the Faraday rotation measurement of tokamak plasma by polarimetry

Dynamic behavior of transport in normal and reversed shear plasmas with internal barriers in JT-60U

International Nuclear Information System (INIS)

Neudatchin, Sergi V.; Takizuka, Tomonori; Shirai, Hiroshi; Fujita, Takaaki; Isayama, Akihiko; Kamada, Yutaka; Koide, Yoshihiko

2001-12-01

Transport evolution in normal shear (NrS) and reversed shear (RS) JT-60U tokamak plasmas with internal transport barrier (ITB) is described as a combination of various fast and slow time scale processes. Abrupt in time (ms time scale) and wide in space (∼0.3 of minor radius) variations of electron and ion heat diffusivities χ e,i (δχ e,i ), which are called ITB-events and seen as simultaneous rise and decay of electron and ion temperatures in two spatial zones, are found for weak ITBs in both NrS and RS plasmas. Profiles of δχ e in RS plasmas with strong ITBs are usually localized near ITB foot inside smaller space region. The maximum of the heat flux variation is located near position of the minimum of safety factor in various RS plasmas, and variation is extended in positive shear region. Inward and outward heat pulse propagations created by the jump of χ e and the sawtooth-like crash are analyzed. Small values of χ e and the absence of heat pinch are found inside strong ITBs. Another non-local abrupt variations of χ e inside most of the plasma volume, including significant part of weak ITB inside RS zone of RS plasmas, are seen at the ELM-induced H-L transition and the L-H recovery. (author)

Preliminary design study of a steady state tokamak device

International Nuclear Information System (INIS)

Miya, Naoyuki; Nakajima, Shinji; Ushigusa, Kenkichi; and athors)

1992-09-01

Preliminary design study has been made for a steady tokamak with the plasma current of 10MA, as the next to the JT-60U experimental programs. The goal of the research program is the integrated study of steady state, high-power physics and technology. Present candidate design is to use superconducting TF and PF magnet systems and long pulse operation of 100's-1000's of sec with non inductive current drive mainly by 500keV negative ion beam injection of 60MW. Low activation material such as titanium alloy is chosen for the water tank type vacuum vessel, which is also the nuclear shield for the superconducting coils. The present preliminary design study shows that the device can meet the existing JT-60U facility capability. (author)

Data management facility for JT-60

International Nuclear Information System (INIS)

Ohasa, K.; Kurimoto, K.; Mochizuki, O.

1983-01-01

This study considers the Data Management Facility which is provided for unified management of various diagnostics data with JT-60 experiments. This facility is designed for the purpose of data access. There are about 30 kinds of diagnostic devices that are classified by diagnostic objects equipped for JT-60 facility. It gathers the diagnostic date about 10 Mega Byte per each discharge. Those diagnostic data are varied qualitatively and quantitatively by experimental purpose. Other fundamental information like discharge condition, adjustive value for diagnostic devices is required to process those gathered data

Alfven eigenmodes driven by Alfvenic beam ions in JT-60U

International Nuclear Information System (INIS)

Shinohara, K.; Kusama, Y.; Takechi, M.

2001-01-01

Instabilities with frequency chirping in the frequency range of Alfven eigenmodes have been found in the domain 0.1% h > bparallel /υ A ∼ 1 with high energy neutral beam injection in JT-60U. One instability with a frequency inside the Alfven continuum spectrum appears and its frequency increases slowly to the toroidicity induced Alfven eigenmode (TAE) gap on the timescale of an equilibrium change (∼ 200 ms). Other instabilities appear with a frequency inside the TAE gap and their frequencies change very quickly by 10-20 kHz in 1-5 ms. During the period when these fast frequency sweeping (fast FS) modes occur, abrupt large amplitude events (ALEs) often appear with a drop of neutron emission rate and an increase in fast neutral particle fluxes. The loss of energetic ions increases with a peak fluctuation amplitude of B-tilde θ /B θ . An energy dependence of the loss ions is observed and suggests a resonant interaction between energetic ions and the mode. (author)

Fast neutron-gamma discrimination on neutron emission profile measurement on JT-60U

International Nuclear Information System (INIS)

Ishii, K.; Okamoto, A.; Kitajima, S.; Sasao, M.; Shinohara, K.; Ishikawa, M.; Baba, M.; Isobe, M.

2010-01-01

A digital signal processing (DSP) system is applied to stilbene scintillation detectors of the multichannel neutron emission profile monitor in JT-60U. Automatic analysis of the neutron-γ pulse shape discrimination is a key issue to diminish the processing time in the DSP system, and it has been applied using the two-dimensional (2D) map. Linear discriminant function is used to determine the dividing line between neutron events and γ-ray events on a 2D map. In order to verify the validity of the dividing line determination, the pulse shape discrimination quality is evaluated. As a result, the γ-ray contamination in most of the beam heating phase was negligible compared with the statistical error with 10 ms time resolution.

Formation conditions for electron internal transport barriers in JT-60U plasmas

Energy Technology Data Exchange (ETDEWEB)

Fujita, T [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Fukuda, T [Osaka University, Suita, Osaka 565-0871 (Japan); Sakamoto, Y [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Ide, S [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Suzuki, T [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Takenaga, H [Japan Atomic Energy Research Institute, Naka Fusion Research Establishment, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Ida, K [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Idei, H [Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Shimozuma, T [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Fujisawa, A [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Ohdachi, S [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Toi, K [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan)

2004-05-01

The formation of electron internal transport barriers (ITBs) was studied using electron cyclotron (EC) heating in JT-60U positive shear (PS) and reversed shear (RS) plasmas with scan of neutral beam (NB) power. With no or low values of NB power and with a small radial electric field (E{sub r}) gradient, a strong, box-type electron ITB was formed in RS plasmas while a peaked profile with no strong electron ITBs was observed in PS plasmas within the available EC power. When the NB power and the E{sub r} gradient were increased, the electron transport in strong electron ITBs with EC heating in RS plasmas was not affected, while electron thermal diffusivity was reduced in conjunction with the reduction of ion thermal diffusivity, and strong electron and ion ITBs were formed in PS plasmas.

Improvement of the protection devices for JT-60U LHRF antenna system

International Nuclear Information System (INIS)

Suzuki, Sadaaki; Seki, Masami; Shinozaki, Shinichi; Sato, Fumiaki; Hiranai, Shinichi; Hasegawa, Koichi; Moriyama, Shinichi; Ishii, Kazuhiro

2007-09-01

In the experiments featuring lower hybrid range of frequency (LHRF) system in JT-60U, carbon grills were attached to the plasma-facing part of the antenna in order to avoid the damage by the excessive heat load from the plasma. However some electric discharge traces were found there in the observation after the experiments. To avoid such discharges, improvements of the arc detector and the protection interlock by visible picture detection were tackled. In the arc detector, the amplification circuit was improved in order to obtain shorter response time and higher resolution of optical detection. Moreover, in visible picture detection, a new function of RF-on/off control utilizing PC image processing was added to distinguish the light of the arc from one of the plasma. This report summarizes improvement of the protection interlock device in a LHRF heating system. (author)

Parametric thermo-hydraulic analysis of the TF system of JT-60SA during fast discharge

International Nuclear Information System (INIS)

Polli, Gian Mario; Lacroix, Benoit; Zani, Louis; Besi Vetrella, Ugo; Cucchiaro, Antonio

2013-01-01

Highlights: • We modeled the central clock-wise pancake of JT-60SA TF magnet at the EOB. • We simulated a quench followed by a fast discharge. • We evaluated the temperature and pressure rises in the nominal configuration. • We evaluated the effect of several parameter changes on the thermal-hydraulic response of the system. -- Abstract: The evolution of the conductor temperature and of the helium pressure of the central pancake of the TF superconducting magnet of the JT-60SA tokamak in a quench scenario are here discussed. The quench is triggered by a heat disturbance applied at the end of burning and followed by a fast safety discharge. A parametric study aimed at assessing the robustness of the calculation is also addressed with special regard to the voltage threshold, used to define the occurrence of the quench, and to the time delay, that cover all the possible delays in the fast discharge after quench detection. Finally, due to sensitivity analyses the influences of different parameters were assessed: the material properties of the strands (RRR, copper fraction), the magnitude and the spatial length of the triggering disturbance and the magnetic field distribution. The numerical evaluations were performed in the framework of the Broader Approach Agreement in collaboration with CEA, ENEA and the JT-60SA European Home Team using the 1D code Gandalf [1

Numerical simulation of a high-brightness lithium ion gun for a Zeeman polarimetry on JT-60U

International Nuclear Information System (INIS)

Kojima, Atsushi; Kamiya, Kensaku; Fujita, Takaaki; Kamada, Yutaka; Iguchi, Harukazu

2007-01-01

A lithium ion gun is under construction for a lithium beam Zeeman polarimetry on JT-60U. The performance of the prototype ion gun has been estimated by the numerical simulation taking the space charge effects into account. The target values of the ion gun are the beam energy of 30 keV, the beam current of 10 mA and the beam divergence angle within 0.13 degrees. The low divergence of 0.13 degrees is required for the geometry of the Zeeman polarimetry on JT-60U where the observation area is 6.5 m away from the neutralizer. The numerical simulation needs to be carried out for the design study because the requirement of the divergence angle is severe for the development of the high-brightness ion gun. The simulation results show the beam loss of 50% caused by the clash to the electrode such as the cathode and the neutralizer. Moreover, the beam transport efficiency from the neutralizer to the observation area is low due to the broadening of the divergence angle. The total beam efficiency is about 5%. Extracted beam profile affects the beam focusing and the efficiency. The peaked profile achieves better efficiency than the hollow one. As a result, beam current of 1 mA is obtained at the observation area by the simulation for the prototype ion gun. (author)

Heating experiments of JT-60

International Nuclear Information System (INIS)

1987-01-01

In JT-60, after the finish of the first stage Joule experiment, the heating facilities were installed, and the heating experiment was started in August, 1986. As to neutral beam injection, the beam injection experiment at the maximum rating 20 MW carried out, and also as to RF, the injection experiment up to 1.4 MW was carried out in both ion cyclotron and low band hybrid waves. The results worthy of special mention in the heating experiment were the success in the current drive up to 1.7 MA at maximum using low band hybrid waves and the improvement of plasma confinement characteristics obtained by the compound heating of NBI and RF. In this paper, the main results of these heating experiments and their significance are explained. The JT-60 is the testing facilities for attaining the critical plasma condition by additionally heating the plasma which is generated by Joule electric discharge with NBI and RF heatings. The experimental operation cycle of the JT-60 consists of the unit cycle of two weeks, and the number of days in operation is nine days. The temperature of heated plasma rose to 70 million deg C in the 20 MW NBI heating. Hereafter, the improvement of confinement time by increasing the stored energy of plasma is attempted. (Kako, I.)

Advanced control scenario of high-performance steady-state operation for JT-60 superconducting tokamak

International Nuclear Information System (INIS)

Tamai, H.; Kurita, G.; Matsukawa, M.; Urata, K.; Sakurai, S.; Tsuchiya, K.; Morioka, A.; Miura, Y.M.; Kizu, K.; Kamada, Y.; Sakasai, A.; Ishida, S.

2004-01-01

Plasma control on high-β N steady-state operation for JT-60 superconducting modification is discussed. Accessibility to high-β N exceeding the free-boundary limit is investigated with the stabilising wall of reduced-activated ferritic steel and the active feedback control of the in-vessel non-axisymmetric field coils. Taking the merit of superconducting magnet, advanced plasma control for steady-state high performance operation could be expected. (authors)

Validation of special processes for the integration activities of the JT-60SA TF coils manufactured in Italy

Energy Technology Data Exchange (ETDEWEB)

Polli, Gian Mario, E-mail: gianmario.polli@enea.it [ENEA, UT-FUS, Via E. Fermi 45, Frascati (Italy); Cucchiaro, Antonio; Cocilovo, Valter [ENEA, UT-FUS, Via E. Fermi 45, Frascati (Italy); Drago, Giovanni; Pesenti, Paolo; Cuneo, Stefano; Terzi, Franco [ASG Superconductors, Corso Perrone 73 r, Genova (Italy); Phillips, Guy; Tomarchio, Valerio [JT-60SA European Home Team, 85748 Garching bei Munchen (Germany)

2015-10-15

Highlights: • Insertion. • Casing welding. • Casing embedding. - Abstract: In the framework of the Broader Approach Agreement for the construction of the JT-60SA tokamak, ENEA provides 9 of the 18 toroidal field (TF) coils of the JT-60SA magnet system. The 9 coils are being manufactured by ASG superconductors in Genoa under the supervision of ENEA in collaboration with the JT-60SA European home team. The manufacturing is composed of two main steps: one concerning winding pack assembly and impregnation, and the other devoted to the integration into the casing structure and associated final coil preparation. This paper describes the results of the validation activities set-up for the integration phase. Specifically, welding of casing components has been retained particularly critical for at least three reasons: (i) during welding the WP may be damaged by the intense heating; (ii) distortion caused by heating may determine incorrect coil geometry and then field errors; and (iii) flaws may reduce structural strength and then the overall lifetime of the machine. Similarly, final embedding has been demonstrated on a 1 m long mock-up of the coil. Main results and lessons learned are here described.

Discharge optimization in JT-60

International Nuclear Information System (INIS)

Ninomiya, H.; Hosogane, N.; Kikuchi, M.; Yoshino, R.; Seki, S.; Kurihara, K.; Kimura, T.; Shimada, R.; Matsukawa, M.

1986-01-01

For the optimization of the feedback control gains of the plasma control system in JT-60, the plasma modelling by the regression analysis, the matrix transfer function analysis and the simulation study are performed. The experimental results of plasma control are well consistent with these estimations and the usefulness of a modelling by the regression analysis, the matrix transfer function analysis and the simulation study is experimentally confirmed. It is also shown that the regression analysis is useful for development of the sensor algorithm of plasma shape and location of separatrix line in a feedback control system. Some topics are also presented about plasma engineering obtained in JT-60: possibility to suppress the uncontrollability of plasma density, αI/sub p/ control for plasma position and volt-sec consumption

Development of the centrifugal pellet injector for JT-60U

Energy Technology Data Exchange (ETDEWEB)

Kizu, Kaname; Hiratsuka, Hajime; Ichige, Hisashi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

2001-03-01

For core fueling of JT-60U plasmas, a repetitive pellet injector which centrifugally accelerates D{sub 2} cubic pellets using a straight rod has been developed. This centrifugal pellet injector can eject trains of up to 40 cubic pellets at frequencies of 1-10 Hz and velocities of 0.3-1.0 km/s. The average pellet mass is 3.6x10{sup 20} atoms/pellet below 0.7 m/s. Key techniques for the development were a mesh structured acceleration component for removing gas sublimated from the pellet and a funnel with an appropriate angle connected just behind the acceleration chamber for introducing the pellet to plasma without destruction. Using the mesh structured components, the horizontal angular distribution of pellets ejected became narrow, because irregular pellet motion caused by sublimated gas was reduced. To investigate the performance of the injector, pellet injection experiments from the low magnetic field side (LFS) were conducted using ohmic heating plasmas. Central fueling and enhanced fueling rate have been observed. D{alpha} intensity around the divertor region was reduced in a pellet injection plasma compared to gas puffing, indicating low recycling rate was maintained with the pellet injection. (author)

Development of the centrifugal pellet injector for JT-60U

International Nuclear Information System (INIS)

Kizu, Kaname; Hiratsuka, Hajime; Ichige, Hisashi

2001-03-01

For core fueling of JT-60U plasmas, a repetitive pellet injector which centrifugally accelerates D 2 cubic pellets using a straight rod has been developed. This centrifugal pellet injector can eject trains of up to 40 cubic pellets at frequencies of 1-10 Hz and velocities of 0.3-1.0 km/s. The average pellet mass is 3.6x10 20 atoms/pellet below 0.7 m/s. Key techniques for the development were a mesh structured acceleration component for removing gas sublimated from the pellet and a funnel with an appropriate angle connected just behind the acceleration chamber for introducing the pellet to plasma without destruction. Using the mesh structured components, the horizontal angular distribution of pellets ejected became narrow, because irregular pellet motion caused by sublimated gas was reduced. To investigate the performance of the injector, pellet injection experiments from the low magnetic field side (LFS) were conducted using ohmic heating plasmas. Central fueling and enhanced fueling rate have been observed. Dα intensity around the divertor region was reduced in a pellet injection plasma compared to gas puffing, indicating low recycling rate was maintained with the pellet injection. (author)

Power and particle control in JT-60SA to support and supplement ITER and DEMO

International Nuclear Information System (INIS)

Sakurai, Shinji

2007-01-01

JT-60 is planned to be modified as a fully superconducting coil tokamak (JT-60 Super Advanced, JT-60SA). Divertor targets are water-cooled to handle heat flux up to 15 MW/m 2 . JT-60SA allows exploitation of high beta regime with stabilizing shell covered with ferritic plates and internal resistive wall mode (RWM) stabilizing coils. A remote handling system is equipped to maintain in-vessel components even for high dose rate due to a substantial annual neutron production. Divertor cassettes are introduced to be maintained by a remote handling. In the present design, a monoblock type carbon fibre composite (CFC) divertor target will be used to withstand a heat load of ∼15 MW/m 2 . CFC divertor targets and other bolted armor tiles will be mounted on the divertor cassette. All of the plasma facing components including the first wall armor are water-cooled to handle heat load during 100s or more. Divertor heat load and pumping efficiency for an ITER-like configuration has been evaluated, using 2D plasma fluid (SOLDOR) and neutral Monte-Carlo (NEUT2D) code. The pumping speed of 50 m 3 /s is specified at an albedo for neutrals in front of the in-vessel cryopanel. In the simulation for the divertor with a V -shaped corner' like as that in ITER, the plasma detachment occurs near the outer-strike point within the 'V-shaped corner', as well as near the inner-strike point, which results in low peak heat flux density 5.8 MW/m 2 for the case with additional gas puff of 5x10 21 /s compared to 11.4 MW/m 2 for the case without 'V-shaped corner'. (author)

Edge safety factor at the onset of plasma disruption during VDEs in JT-60U

International Nuclear Information System (INIS)

Sugihara, Masayoshi; Lukash, Victor; Khayrutdinov, Rustam; Neyatani, Yuzuru

2004-01-01

Detailed examinations of the value of the edge safety factor (q a ) at the onset of thermal quench (TQ) during intentional vertical displacement event (VDE) experiments in JT-60U are carried out using two different reconstruction methods, FBI/FBEQU and DINA. The results from the two methods are very similar and show that the TQ occurs when the q a value is in the range between 1.5 and 2. This result suggests that the predictive simulations for VDEs should be performed within this range of q to examine the subsequent differences in the halo currents, plasma movement and other plasma behaviour during the current quench

Divertor characteristics and control on the W-shaped divertor with pump of JT-60U

International Nuclear Information System (INIS)

Hosogane, N.; Kubo, H.; Higashijima, S.

1999-01-01

Roles of the inner leg pumping and the private dome, which are special features of the W-shaped divertor of JT-60U, have been investigated. The following observations were made: The inner leg pumping functions well in attached states or partially detached states with weak X-point MARFE where the inner particle recycling is enhanced. A combination of main gas puff and inner leg pump is effective in reduction of intrinsic carbon impurity. Geometrical effects of the private dome on transport of hydrocarbons in the private flux region was confirmed by spectroscopic measurements of CD-band intensity profile and impurity transport simulation code using experimental data. (author)

Behaviour of tritium in the vacuum vessel of JT-60U

International Nuclear Information System (INIS)

Kobayashi, K.; Miya, N.; Ikeda, Y.; Torikai, Y.; Saito, M.; Alimov, V.

2015-01-01

The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D 2 (92.8 %) - T 2 (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily

Behaviour of tritium in the vacuum vessel of JT-60U

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, K.; Miya, N.; Ikeda, Y. [JT-60 Safety Assessment Group, JAEA, Mukoyama (Japan); Torikai, Y. [Hydrogen Isotope Research Center, University of Toyama, Gofuku (Japan); Saito, M.; Alimov, V. [ITER Project Management Group, JAEA, Mukoyama (Japan)

2015-03-15

The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D{sub 2} (92.8 %) - T{sub 2} (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily.

Conceptual design of a new supervisory control system for JT-60SA

International Nuclear Information System (INIS)

Totsuka, Toshiyuki; Sakata, Shinya

2009-05-01

The functions of JT-60 discharge control computer system and the data processing computer system will be integrated into a new JT-60SA supervisory control system to improve the operational efficiency of the JT-60 control computer system. In this report, we first show the necessary requirements for the new JT-60SA supervisory control system that should have high cost performance and maintainability. Next, overall system image of the new JT-60SA supervisory control system is presented and the necessary functions and the issues to be solved in the development are shown. Finally, the necessary manpower for this development and performance of the computer hardware, and the expected reduction of maintenance cost of the computer system are described. (author)

Retention of Hydrogen Isotopes in Divertor Tiles Used in JT-60U

International Nuclear Information System (INIS)

Hirohata, Y.; Shibahara, T.; Tanabe, T.; Oya, Y.; Arai, T.; Gotoh, Y.; Masaki, K.; Yagyu, J.; Oyaidzu, M.; Okuno, K.; Nishikawa, M.; Miya, N.

2005-01-01

Retention characteristics of deuterium and hydrogen retained in graphite tiles placed in the divertor region of JT-60U were investigated by thermal desorption spectroscopy (TDS). The deuterium retained in the near surface of all graphite tiles was mostly replaced by hydrogen due to exposure to hydrogen plasma at the final stage operations, resulting in main deuterium retention in the deeper region. The dominant species desorbed from the divertor tiles were H 2 , HD, D 2 and CH 4 . The smallest retention of hydrogen isotopes (H+D) was observed in the outer divertor tile which was eroded with maximum of 20 μm depth. The amount of H+D retained in the inner divertor tiles covered by the re-deposited layers increased with the thickness of the re-deposited layers. Hydrogen isotopes concentration ((H+D)/C) in the re-deposited layers was ∼0.02, which was much smaller than those observed in JET and other devices

Development of fast opening magnetic valve for JT-60 pellet injector

International Nuclear Information System (INIS)

Hiratsuka, Hajime; Kawasaki, Kouzo; Takatsu, Hideyuki; Miyo, Yasuhiko; Yoshioka, Yuji; Ohta, Kazuya; Shimizu, Masatsugu; Onozuka, Masanori; Uchikawa, Takashi; Iwamoto, Syuichi; Hashiri, Noboru

1989-01-01

A pneumatic four-pellet injector (JT-60 pellet injector) has been constructed for JT-60 in May, 1988. A fast opening magnetically driven propellant gas injection valve has been developed for JT-60 pellet injector. This valve can accelerate four cylindrical pellets, two 3.8 mm diameter by 3.8 mm and two 2.7 mm diameter by 2.7 mm, to greater than 1.6 km/s with propellent gas of up to 50 bar. It is now successfully in use in JT-60, contributing to plasma studies. In this paper the outline of a newly developed fast opening magnetic valve and the results of performance tests are presented. (author). 6 figs.; 1 tab

Vacuum leak test technique of JT-60

International Nuclear Information System (INIS)

Kaminaga, Atsushi; Arai, Takashi; Kodama, Kozo; Sasaki, Noboru; Saidoh, Masahiro

1998-01-01

Since a vacuum vessel of JT-60 is very large (167 m 3 ) and is combined with many components, such as magnetic coils, neutral beam injection systems and RF heating systems, etc., the position of leak testing exceeds 700. The two kind of techniques for vacuum leak test used in JT-60 has been described. Firstly the probe helium gas can be fed remotely in the three-dimensionally sectioned 54 regions of the JT-60 torus. The leak test was very rapidly performed by using this method. Secondly the helium detector system has been modified by the additional installation of the cryopump, which reduced the background level of the deuterium gas. The sensitivity of vacuum leak test with the cryopump was two orders of magnitude larger than that of without it. The examples of the performed vacuum leak test are stated. The vacuum leaks during experiments were 9 times. They were caused by thermal strain and plasma discharge. The vacuum leaks just after maintenance are 36 times which mainly caused by mis-installation. (author)

Tokamaks (Second Edition)

Energy Technology Data Exchange (ETDEWEB)

Stott, Peter [JET, UK (United Kingdom)

1998-10-01

The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

Tokamaks (Second Edition)

International Nuclear Information System (INIS)

Stott, Peter

1998-01-01

The first edition of John Wesson's book on tokamaks, published in 1987, established itself as essential reading for researchers in the field of magnetic confinement fusion: it was an excellent introduction for students to tokamak physics and also a valuable reference work for the more experienced. The second edition, published in 1997, has been completely rewritten and substantially enlarged (680 pages compared with 300). The new edition maintains the aim of providing a simple introduction to basic tokamak physics, but also includes discussion of the substantial advances in fusion research during the past decade. The new book, like its predecessor, is well written and commendable for its clarity and accuracy. In fact many of the chapters are written by a series of co-authors bringing the benefits of a wide range of expertise but, by careful editing, Wesson has maintained a uniformity of style and presentation. The chapter headings and coverage for the most part remain the same - but are expanded considerably and brought up to date. The most substantial change is that the single concluding chapter in the first edition on 'Experiments' has been replaced by three chapters: 'Tokamak experiments' which deals with some of the earlier key experiments plus a selection of recent small and medium-sized devices, 'Large experiments' which gives an excellent summary of the main results from the four large tokamaks - TFTR, JET, JT60/JT60U and DIII-D, and 'The future' which gives a very short (possibly too short in my opinion) account of reactors and ITER. This is an excellent book, which I strongly recommend should have a place - on the desk rather than in the bookshelf - of researchers in magnetic confinement fusion. (book review)

Present construction status and future plan of JT-60

Energy Technology Data Exchange (ETDEWEB)

Iso, Yasuhiko; Yoshikawa, Masaji [Japan Atomic Energy Research Inst., Tokai, Ibaraki. Tokai Research Establishment

1982-04-01

As for the critical plasma test facility JT-60, its detailed design was initiated in 1975, and the manufacture of the main body components was started in April, 1978, through the trial manufacture and development of important components. The manufacture of all the system equipment in factories is in progress. Especially, the essential components of main body have been almost completed, and ready to be installed in the building. The building is under construction, and a part of the equipment was brought in and installed since the beginning of 1982 fiscal year. JT-60 is composed of heating system, toroidal magnetic field coil power supply, poloidal magnetic field coil power supply, total control system, associated system for main body and various measuring equipment as well as the main body which consists of vacuum vessel, toroidal and poloidal magnetic field coils, base, primary cooling system, movable limiter and evacuating system. In this report, first the significance of JT-60 development and the outline of JT-60 are described, and the present construction status is reported. The specifications of the essential equipment of the main body and heating system and the measuring system configuration are listed in tables. The aims in JT-60 design are to make plasma generating pulse width longer, to obtain higher beta value in circular cross-section plasma, to install the magnetic limiter, and to realize compound second stage heating.

Design of Plasma Facing Components for Superconducting Modification of JT-60

International Nuclear Information System (INIS)

Shinji Sakurai; Kei Masaki; Yusuke-Kudo Shibama; Hiroshi Tamai; Makoto Matsukawa; Cordier, J.J.

2006-01-01

JT-60 is planning to modify the machine as a fully superconducting coil tokamak (JT-60 Super Advanced, the former JT-60SC and NCT) to establish scientific and technological bases for an economically and environmentally attractive DEMO reactor. It will be also a satellite tokamak in a part of broader approach for ITER. It is designed for high beta (betaN = 3.5-5.5) and steady-state research in a break-even class DD plasma for 100 s or longer. Nominal plasma parameters are I p =5.5 MA, B t =2.7 T, R=3.01 m, a=1.14 m with double-null configuration. An ITER-like single-null configuration with I p =3.5 MA, B t =2.6 T can be also operated. In order to study the ITER-relevant high confinement plasma with high density, designed plasma heating power was enhanced from 25 MW to 41 MW for 100 s through the design review with EU and Japan. The heat flux onto outer divertor target exceeds 10 MW/m 2 with moderate radiative fraction of 50-60% in single-null configuration. Therefore, the ITER-like mono-block CFC target will be adopted to aim at power handling of 15 MW/m 2 . A cooling water system should be reinforced 3 times from original design for double null divertor with high coolant flow velocity of ∼10 m/s. The peak heat flux onto the neutral beam armor for perpendicular injected positive NB is evaluated to be 2 MW/m 2 , which needs to be actively water-cooled. A bolt-fixed CFC tile was tested at the heat flux of 1-3 MW/m 2 and will be applied to the NB armor. In order to improve plasma beta value by enhancing wall stabilization effect, passive-stabilizing plates, which are electrically and mechanically connected in poloidal and toroidal direction, will be installed near the plasma surface (r wall /a=1.1-1.3) at the outboard side. Stabilizing plate has double-wall ribbed structure and can be operated at 573 K with heating nitrogen gas instead of cooling water between double walls. It has crank-type support legs to allow thermal expansion at high temperature operation. The

Physics issues of high bootstrap current tokamaks

International Nuclear Information System (INIS)

Ozeki, T.; Azumi, M.; Ishii, Y.

1997-01-01

Physics issues of a tokamak plasma with a hollow current profile produced by a large bootstrap current are discussed based on experiments in JT-60U. An internal transport barrier for both ions and electrons was obtained just inside the radius of zero magnetic shear in JT-60U. Analysis of the toroidal ITG microinstability by toroidal particle simulation shows that weak and negative shear reduces the toroidal coupling and suppresses the ITG mode. A hard beta limit was observed in JT-60U negative shear experiments. Ideal MHD mode analysis shows that the n = 1 pressure-driven kink mode is a plausible candidate. One of the methods to improve the beta limit against the kink mode is to widen the negative shear region, which can induce a broader pressure profile resulting in a higher beta limit. The TAE mode for the hollow current profile is less unstable than that for the monotonic current profile. The reason is that the continuum gaps near the zero shear region are not aligned when the radius of q min is close to the region of high ∇n e . Finally, a method for stable start-up for a plasma with a hollow current profile is describe, and stable sustainment of a steady-state plasma with high bootstrap current is discussed. (Author)

Quench detection of fast plasma events for the JT-60SA central solenoid

International Nuclear Information System (INIS)

Murakami, Haruyuki; Kizu, Kaname; Tsuchiya, Katsuhiko; Kamiya, Koji; Takahashi, Yoshikazu; Yoshida, Kiyoshi

2012-01-01

Highlights: ► Pick-up coil method is used for the quench detection of JT-60SA magnet system. ► Disk-shaped pick-up coils are inserted in CS module to compensate inductive voltage. ► Applicability of pick-up coil is evaluated by two dimensional analysis. ► Pick-up coil is applicable whenever disruption, mini collapse and other plasma event. - Abstract: The JT-60 is planned to be modified to a full-superconducting tokamak referred to as the JT-60 Super Advanced (JT-60SA). The maximum temperature of the magnet during its quench might reach the temperature of higher than several hundreds Kelvin that will damage the superconducting magnet itself. The high precision quench detection system, therefore, is one of the key technologies in the superconducting magnet protection system. The pick-up coil method, which is using voltage taps to detect the normal voltage, is used for the quench detection of the JT-60SA superconducting magnet system. The disk-shaped pick-up coils are inserted in the central solenoid (CS) module to compensate the inductive voltage. In the previous study, the quench detection system requires a large number of pick-up coils. The reliability of quench detection system would be higher by simplifying the detection system such as reducing the number of pick-up coils. Simplifying the quench detection system is also important to reduce the total cost of the protection system. Hence the design method is improved by increasing optimizing parameters. The improved design method can reduce the number of pick-up coils without reducing the sensitivity of detection; consequently the protection system can be designed with higher reliability and lower cost. The applicability of the disk-shaped pick-up coil for quench detection system is evaluated by the two dimensional analysis. In the previous study, however, the analysis model only took into account the CS, EF (equilibrium field) coils and plasma. Therefore, applicability of the disk-shaped pick-up coil for

Design optimization of JT-60SU for steady-state advanced operation

International Nuclear Information System (INIS)

Ushigusa, K.; Kurita, G.; Toyoshima, N.

2001-01-01

Design optimization of JT-60SU has been done for a steady-state advanced operation. A transport code simulation indicates that a fully non-inductive reversed shear plasmas with fractions of 70% of the bootstrap current and 30% of beam driven current can be sustained for more than 1,000s without any additional control. Investigations have been progressed on MHD stability, vertical positional stability and dynamics of the vertical displacement events. Significant progress has been achieved in the R and D of Nb 3 Al superconducting wires, low induced activation material (Fe-Cr-Mn steel). A design improvement has been made in TF coils to reduce a local stress on radial disk. Dynamic behaviors of the tokamak machine have been analyzed at emergency events such as an earthquake. (author)

Maximum β limited by ideal MHD ballooning instabilites in JT-60

International Nuclear Information System (INIS)

Seki, Shogo; Azumi, Masashi

1986-03-01

Maximum β limited by ideal MHD ballooning instabilities is investigated on divertor configurations in JT-60. Maximum β against ballooning modes in JT-60 has strong dependecy on the distribution of the safety factor over the magnetic surfaces. Maximum β is ∼ 2 % for q 0 = 1.0, while more than 3 % for q 0 = 1.5. These results suggest that the profile control of the safety factor, especially on the magnetic axis, is attractive to the higher β operation in JT-60. (author)

Plasma equilibrium control during slow plasma current quench with avoidance of plasma-wall interaction in JT-60U

Science.gov (United States)

Yoshino, R.; Nakamura, Y.; Neyatani, Y.

1997-08-01

In JT-60U a vertical displacement event (VDE) is observed during slow plasma current quench (Ip quench) for a vertically elongated divertor plasma with a single null. The VDE is generated by an error in the feedback control of the vertical position of the plasma current centre (ZJ). It has been perfectly avoided by improving the accuracy of the ZJ measurement in real time. Furthermore, plasma-wall interaction has been avoided successfully during slow Ip quench owing to the good performance of the plasma equilibrium control system

Design of tangential viewing phase contrast imaging for turbulence measurements in JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Tanaka, K., E-mail: ktanaka@nifs.ac.jp [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Department of Advanced Energy Engineering, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Coda, S. [EPFL–SPC, Lausanne (Switzerland); Yoshida, M.; Sasao, H.; Kawano, Y.; Imazawa, R.; Kubo, H.; Kamada, Y. [National Institutes for Quantum and Radiological Science and Technology, Naka, Ibaraki 311-0193 (Japan)

2016-11-15

A tangential viewing phase contrast imaging system is being designed for the JT-60SA tokamak to investigate microturbulence. In order to obtain localized information on the turbulence, a spatial-filtering technique is applied, based on magnetic shearing. The tangential viewing geometry enhances the radial localization. The probing laser beam is injected tangentially and traverses the entire plasma region including both low and high field sides. The spatial resolution for an Internal Transport Barrier discharge is estimated at 30%–70% of the minor radius at k = 5 cm{sup −1}, which is the typical expected wave number of ion scale turbulence such as ion temperature gradient/trapped electron mode.

Performance verification tests of JT-60SA CS model coil

Energy Technology Data Exchange (ETDEWEB)

Obana, Tetsuhiro, E-mail: obana.tetsuhiro@LHD.nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Murakami, Haruyuki [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Takahata, Kazuya; Hamaguchi, Shinji; Chikaraishi, Hirotaka; Mito, Toshiyuki; Imagawa, Shinsaku [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Kizu, Kaname; Natsume, Kyohei; Yoshida, Kiyoshi [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2015-11-15

Highlights: • The performance of the JT-60SA CS model coil was verified. • The CS model coil comprised a quad-pancake wound with a Nb{sub 3}Sn CIC conductor. • The CS model coil met the design requirements. - Abstract: As a final check of the coil manufacturing method of the JT-60 Super Advanced (JT-60SA) central solenoid (CS), we verified the performance of a CS model coil. The model coil comprised a quad-pancake wound with a Nb{sub 3}Sn cable-in-conduit conductor. Measurements of the critical current, joint resistance, pressure drop, and magnetic field were conducted in the verification tests. In the critical-current measurement, the critical current of the model coil coincided with the estimation derived from a strain of −0.62% for the Nb{sub 3}Sn strands. As a result, critical-current degradation caused by the coil manufacturing process was not observed. The results of the performance verification tests indicate that the model coil met the design requirements. Consequently, the manufacturing process of the JT-60SA CS was established.

Identity physics experiment on internal transport barriers in JT-60U and JET

Energy Technology Data Exchange (ETDEWEB)

De Vries, P C; Beurskens, M N A; Brix, M; Giroud, C; Hawkes, N C; Parail, V [EURATOM/UKAEA Association, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Sakamoto, Y; Fujita, T; Hayashi, N; Matsunaga, G; Oyama, N; Shinohara, K; Suzuki, T; Takechi, M [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Litaudon, X; Joffrin, E [CEA, IRFM, F-13108 St-Paul-Lez-Durance (France); Crombe, K [Department of Applied Physics, Ghent University, Rozier 44, 9000 Gent (Belgium); Mantica, P [Istituto di Fisica del Plasma, EURATOM/ENEA-CNR Association, Milano (Italy); Salmi, A [Association Euratom-Tekes, Helsinki University of Technology, PO Box 4100 (Finland); Strintzi, D, E-mail: Peter.de.Vries@jet.u [National Technical University of Athens, EURATOM Association, GR-15773, Athens (Greece)

2009-12-15

A series of experiments have been carried out in 2008 at JT-60U and JET to find common characteristics and explain differences between internal transport barriers (ITBs). The identity experiments succeeded in matching the profiles of most dimensionless parameters at the time ITBs were triggered. Thereafter the q-profile development deviated due to differences in non-inductive current density profile, affecting the ITB. Furthermore, the ITBs in JET were more strongly influenced by the H-mode pedestal or edge localized modes. It was found to be difficult to match the plasma rotation characteristics in both devices. However, the wide range of Mach numbers obtained in these experiments shows that the rotation has little effect on the triggering of ITBs in plasmas with reversed magnetic shear. On the other hand the toroidal rotation and more specifically the rotational shear had an impact on the subsequent growth and allowed the formation of strong ITBs.

Properties of internal transport barrier formation in JT-60U

International Nuclear Information System (INIS)

Sakamoto, Yoshiteru; Suzuki, T.; Ide, S.

2003-01-01

The dependence of the ion thermal diffusivity (χ i ) on the radial electric field (E r ) shear has been investigated in JT-60U plasmas. In positive magnetic shear (PS) plasmas, χ i in the core region generally increases with the heating power, similar to the L mode at low heating power. However, as a result of the intensive central heating, which is relevant to the enhancement of the E γ shear, a weak internal transport barrier (ITB) is formed, and χ i in the core region starts to decrease. Corresponding to a further increase of the heating power, a strong ITB is formed and χ i is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of χ i is observed in any of heating regimes. The electron thermal diffusivity (χ e ) is strongly correlated with χ i in PS and RS plasmas. There exists a threshold in the effective E γ shear to change the state from a weak to a strong ITB. It is found that the threshold of the effective E γ shear in the case of a PS plasma depends on the poloidal magnetic field at the ITB. There are multiple levels of reduced transport in the strong ITB for RS plasmas. (author)

Properties of internal transport barrier formation in JT-60U

International Nuclear Information System (INIS)

Sakamoto, Y.; Suzuki, T.; Ide, S.

2003-01-01

The dependence of the ion thermal diffusivity (χ i ) on the radial electric field (E r ) shear has been investigated in JT-60U plasmas. In positive magnetic shear (PS) plasmas, χ i in the core region generally increases with the heating power, similar to the L mode at low heating power. However, as a result of the intensive central heating, which is relevant to the enhancement of the E r shear, a weak internal transport barrier (ITB) is formed, and χ i in the core region starts to decrease. Corresponding to a further increase of the heating power, a strong ITB is formed and χ i is reduced substantially. In the case of reversed magnetic shear (RS) plasmas, on the other hand, no power degradation of χ i is observed in any of the heating regimes. The electron thermal diffusivity (χ e ) is strongly correlated with χ i in PS and RS plasmas. There exists a threshold in the effective E r shear to change the state from a weak to a strong ITB. It is found that the threshold of the effective E r shear in the case of a PS plasma depends on the poloidal magnetic field at the ITB. There are multiple levels of reduced transport in the strong ITB for RS plasmas. (author)

Prototype ion source for JT-60 neutral beam injectors

International Nuclear Information System (INIS)

Akiba, M.

1981-01-01

A prototype ion source for JT-60 neutral beam injectors has been fabricated and tested. Here, we review the construction of the prototype ion source and report the experimental results about the source characteristics that has been obtained at this time. The prototype ion source is now installed at the prototype unit of JT-60 neutral beam injection units and the demonstration of the performances of the ion source and the prototype unit has just started

Carbon deposition and hydrogen retention in tokamak

International Nuclear Information System (INIS)

Tanabe, Tetsuo

2006-01-01

The results of measurements on co-deposition of hydrogen isotopes and wall materials, hydrogen retention, redeposition of carbon and deposition of hydrogen on PMI of JT-60U are described. From above results, selection of plasma facing material and ability of carbon wall is discussed. Selection of plasma facing materials in fusion reactor, characteristics of carbon materials as the plasma facing materials, erosion, transport and deposition of carbon impurity, deposition of tritium in JET, results of PMI in JT-60, application of carbon materials to PFM of ITER, and future problems are stated. Tritium co-deposition in ITER, erosion and transport of carbon in tokamak, distribution of tritium deposition on graphite tile used as bumper limiter of TFTR, and measurement results of deposition of tritium on the Mark-IIA divertor tile and comparison between them are described. (S.Y.)

Simple multijunction launcher with oversized waveguides for lower hybrid current drive on JT-60U

International Nuclear Information System (INIS)

Ikeda, Y.; Naito, O.; Seki, M.; Kondoh, T.; Ide, S.; Anno, K.; Fukuda, H.; Ikeda, Y.; Kitai, T.; Kiyono, K.; Sawahata, M.; Shinozaki, S.; Suganuma, K.; Suzuki, N.; Ushigusa, K.

1994-01-01

A multijunction technique with oversized waveguides has been developed for the lower hybrid current drive launcher on JT-60U. The launcher consists of 4 (toroidal)x4 (poloidal) multijunction modules. RF power in the module is divided toroidally into 12 sub-waveguides at a junction point through an oversized waveguide. This method simplifies the structure of the multijunction launcher with a large number of subwaveguides. A maximum power density up to 25 MW m -2 has been achieved with a low reflection coefficient of less than 2%. The coupling and current drive efficiency are well explained by the designed wave spectra without taking account of higher modes in the oversize waveguides. Thus, the simple multijunction launcher has been demonstrated to excite expected wave spectra with high power handling capability. ((orig.))

Ion heating up to 1 MeV range with higher harmonic ICRF wave on JT-60U

International Nuclear Information System (INIS)

Nemoto, M.; Kusama, Y.; Hamamatsu, K.; Kimura, H.; Fujii, T.; Moriyama, S.; Saigusa, M.; Afanassiev, V.I.

1997-01-01

The properties of protons under accleration by an ion cyclotron range of frequency (ICRF) waves with the second to fourth hydrogen harmonics have been investigated in the JT-60U tokamak at the Japan Atomic Energy Research Institute (JAERI). Protons have been accelerated up to 1 MeV in the presence of an ICRF wave of fixed frequency, neutral beams (NB), and a fixed toroidal magnetic field which is scanned through several plasma discharges. The tail temperature of the protons, which is evaluated in the range 0.32-0.86 MeV, has been observed to increase in the second to third harmonics, however increase of the tail temperature in the third to fourth harmonics has not been observed clearly. Furthermore, the dependence of tail temperature on the harmonic number has been found to be in qualitative agreement with results from a simulation code analysis based upon the one-dimensional Fokker-Planck equation coupled with the kinetic wave equation. Experimental values for the stored energy of the accelerated ions have shown, however, that the response of stored energy to changes in absorbed ICRF power is much stronger than the response to changes in harmonic number. Also, the incremental energy confinement times for heating discharges matching the third and fourth harmonics (3 ω CH) and 4 ω CH) of hydrogen have been observed to be less than half that for those matching the second harmonic. It has been found that suppression of the absorbed ICRF power accompanied with the occurence of cavity resonance in the 3ω CH and 4ω CH heating discharges reduces the stored energy of the accelerated ions and the incremental energy confinement time. (Author)

LHCD current profile control experiments towards steady state improved confinement on JT-60U

International Nuclear Information System (INIS)

Ide, S.; Naito, O.; Oikawa, T.; Fujita, T.; Kondoh, T.; Seki, M.; Ushigusa, K.

2001-01-01

In JT-60U lower hybrid current drive (LHCD) experiments, a reversed magnetic shear configuration that was accompanied by the internal transport barriers was successfully maintained by means of LHCD almost in the full current drive quasi-steady state for 4.7 s. The normalized beta was kept near 1 and the neutron emission rate was almost steady as well indicating no accumulation of impurities into the plasma. Diagnostics data showed that all the profiles of the electron and ion temperatures, the electron density and the current profile were almost unchanged during the LHCD phase. Moreover, capability of LHCD in H-mode plasmas has been also investigated. It was found that the lower hybrid waves can be coupled to an H-mode edge plasma even with the plasma wall distance of about 14 cm. The maximum coupling distance was found to depend on the edge recycling. (author)

Local transport analysis of L-mode plasmas in JT-60 tokamak

International Nuclear Information System (INIS)

Hirayama, Toshio; Kikuchi, Mitsuru; Shirai, Hiroshi; Shimizu, Katsuhiro; Yagi, Masatoshi; Koide, Yoshihiko; Ishida, Shinichi; Azumi, Masafumi.

1991-03-01

Local heat transport has been studied in auxiliary heated JT-60 plasmas with emphasis on understanding the deteriorated confinement observed in L-mode plasmas. The systematic experiment and analysis have been carried out in L-mode phase of divertor (single null, lower X-point), and limiter discharges with hydrogen neutral beam heating into hydrogen plasmas, based on sets of consistent experimental data including ion temperature profiles from CXR measurements. The deterioration in the energy confinement time with increasing the auxiliary heating power, so-called the power scaling, is mainly due to the degradation in ion energy transport. The confinement improvement as the plasma current increases is followed by both improvement in ion and electron transport properties. It is found that the ion thermal diffusivity has an approval dependence on the density. High ion temperature (T i (0) ≤ 12 keV) L-mode plasmas are attained at high β p up to 3.5. The centrally peaked ion temperature is significantly due to the improvement in ion transport property, which is reduced to the level of the electron thermal diffusivities. (author)

Recent developments in the JT-60 data processing system

International Nuclear Information System (INIS)

Matsuda, T.; Saitoh, N.; Tsugita, T.; Oshima, T.; Sakata, S.; Sato, M.; Koiwa, M.; Watanabe, K.

1999-01-01

The JT-60 data processing system was originally a large computer complex system including a lot of micro-computers, several mini-computers, and a mainframe computer. Recently, several improvements have been made to the original system to modernize the system. Many sub-systems composed of aged mini-computers have been replaced with workstations by utilizing recent progress in computer and network technologies. The system can handle ∝300 MB of raw data per discharge, which is three times larger than the amount in the original system. These improvements have been applied to develop element technologies necessary to the remote participation in JT-60 experiments. A remote diagnostics control and monitoring system and a computer system for access to JT-60 data from the Internet are used together with video conferencing systems for a real-time communication. In 1996, the remote participation based on them was successfully demonstrated in collaboration with Japan Atomic Energy Research Institute, Los Alamos National Laboratory, and Princeton Plasma Physics Laboratory. (orig.)

Development of a compact W-shaped pumped divertor in JT-60U

Energy Technology Data Exchange (ETDEWEB)

Sakurai, S.; Hosogane, N.; Masaki, K.; Kodama, K.; Sasajima, T.; Kishiya, K.; Takahashi, S.; Shimizu, K.; Akino, N.; Miyo, Y.; Hiratsuka, H.; Saidoh, M. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Inoue, M.; Umakoshi, T.; Onozuka, M.; Morimoto, M. [Mitsubishi Heavy Industries, Wadasaki-cho, Hyogo-ku, Kobe-shi, 642 (Japan)

1998-09-01

In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current. (orig.) 11 refs.

Studies on first wall and plasma wall interaction in JT-60

International Nuclear Information System (INIS)

Nakamura, Hiroo

1988-12-01

This paper describes studies on first wall and plasma wall interaction in JT-60. Main results are as follows; (1) To select JT-60 first wall material, various RandD were done in FY1975 ∼ 1976. Mo was selected as first wall materials of limiters and divertor plates because of its reliability under a high heat flux condition. (2) Development of low-Z material has been done to reduce impurity problem of Mo first wall. As a result, titanium carbide (TiC) was selected as a coating material on the Mo. High heat load testing has been done for TiC coated Mo limiter same as JT-60. This material can survive under the condition of 1 kW/cm 2 x 10 s, expected in JT-60 limiter design. (3) To reduce high heat load on the divertor plate, separatrix swing is proposed. Optimum frequency of the sweeping is evaluated to be 2 Hz in JT-60. For a discharge with heating power of 30 MW and duration time of 10 s, in addition to the separatrix swing, remote radiative cooling in the divertor region is necessary. Moreover, calculations of erosion thickness have been done for stainless steel, Mo, graphite, TiC and silicon caibide under high heat flux during plasma disruption. (4) In divertor experiments in JT-60, divertor functions on particle, heat load and impurity controls have been demonstrated. In elctron density of 6 x 10 19 m -3 , particle fueling rate of 20 MW NB heating (3 Pa m 3 /s) can be exhausted by divertor pumping system. Effectiveness of remote radiative cooling is demonstrated under the condition of 20 MW NB heating power. Also, separatrix swing is demonstrated to reduce heat load on the divertor plate. Total radiation in main plasma is 5 ∼ 10% of total absorbed power. (author) 120 refs

Work and safety managements for on-site installation, commissioning, tests by EU of quench protection circuits for JT-60SA

International Nuclear Information System (INIS)

Yamauchi, Kunihito; Okano, Jun; Shimada, Katsuhiro; Ohmori, Yoshikazu; Terakado, Tsunehisa; Matsukawa, Makoto; Koide, Yoshihiko; Kobayashi, Kazuhiro; Ikeda, Yoshitaka; Fukumoto, Masahiro; Kushita, Kouhei N.

2016-03-01

The superconducting Satellite Tokamak machine “JT-60SA” under construction in Naka Fusion Institute is an international collaborative project between Japan Atomic Energy Agency (JAEA) as the Implementing Agency (IA) of Japan (JA) and Fusion for Energy (F4E) as the IA of Europe (EU). The contributions for this project are based on the supply of components, and thus European manufacturer shall conduct the installation, commissioning and tests on Naka site under the general supervision by F4E via the designated institute in each EU nation. This means that JAEA had an issue to manage the works by European workers and their safety although there is no direct contract. This report describes the approaches for the work and safety managements, which were agreed with EU after the negotiation, and the completed on-site works for Quench Protection Circuits (QPC) as the first experience for EU in JT-60SA project. (author)

Reduced transport and ER shearing in improved confinement regimes in JT-60U

International Nuclear Information System (INIS)

Shirai, H.; Kikuchi, M.; Takizuka, T.

2001-01-01

The global confinement and the local transport properties of improved core confinement plasmas in JT-60U have been studied in connection with E r shear formation. The improved core confinement mode with ITB, the internal transport barrier, is roughly classified into 'parabolic' type ITBs and 'box' type ITBs. The parabolic type ITB has the reduced thermal diffusivity, χ, in the core region; however, the E r shear, dE r /dr, is not so strong. The box type ITB has a very strong E r shear at the thin ITB layer and the χ value decreases to the level of neoclassical transport there. The estimated ExB shearing rate, ω ExB , becomes almost the same as the linear growth rate of the drift microinstability, γ L , at the ITB layer in the box type ITB. Experiments of hot ion mode plasmas during the repetitive L-H-L transition shows that the thermal diffusivity clearly depends on the E r shear and the strong E r shear contributes to the reduced thermal diffusivity. (author)

Reduced transport and Er shearing in improved confinement regimes in JT-60U

International Nuclear Information System (INIS)

Shirai, H.; Kikuchi, M.; Takizuka, T.

1999-01-01

The global confinement and the local transport properties of improved core confinement plasmas in JT-60U were studied in connection with E r shear formation. In the improved core confinement mode with internal transport barriers (ITBs), these are roughly classified into 'parabolic type' ITBs and 'box type' ITBs. The parabolic type ITB has a reduced thermal diffusivity χ in the core region; however, the E r shear, dE r /dr, is not as strong. The box type ITB has a very strong E r shear at the thin ITB layer and χ decreases to the level of neoclassical transport there. The estimated E x B shearing rate, ω ExB , becomes almost the same as the linear growth rate of the drift microinstability, γ L , at the ITB layer in the box type ITB. Experiments with hot ion mode plasmas during the repetitive L-H-L transition showed that the thermal diffusivity clearly depends on the E r shear and the strong E r shear contributes to the reduced thermal diffusivity. (author)

Ferritic insertion for reduction of toroidal magnetic field ripple on JT-60U

International Nuclear Information System (INIS)

Shinohara, K.; Sakurai, S.; Ishikawa, M.; Tsuzuki, K.; Suzuki, Y.; Masaki, K.; Naito, O.; Kurihara, K.; Suzuki, T.; Koide, Y.; Fujita, T.; Miura, Y.

2007-01-01

Ferritic steel tiles (FSTs) have been installed to improve the energetic ion confinement by reducing a toroidal magnetic field ripple. Aiming at cost-effective installation, orbit-following calculations of energetic ions were carried out for a design of the installation of ferritic steel on the JT-60U by using the fully three dimensional magnetic field orbit-following Monte-Carlo (F3D OFMC) code, which had been developed for ferritic insert experiments on the JFT-2M and can treat the complex magnetic field structure produced by ferritic inserts. The installed FSTs add a non-linear magnetic field on magnetic sensors for plasma control and an equilibrium calculation. The code for real-time control has been modified to take into account the magnetic field by the FSTs. The plasma operation was successfully resumed after usual conditioning processes and real-time plasma control was successfully carried out. The heat load measurement indicates the improved confinement of energetic ions. These results are important for practical application of the ferritic steel, which is a leading candidate of a structural material on a DEMO reactor

Reference design of the power supply system for the resistive-wall-mode control in JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Ferro, Alberto, E-mail: alberto.ferro@igi.cnr.it [Consorzio RFX, C.so Stati Uniti 4, 35127 Padova (Italy); Gaio, Elena [Consorzio RFX, C.so Stati Uniti 4, 35127 Padova (Italy); Novello, Luca [Fusion for Energy, Broader Development of Fusion Department, Boltzmannstr 2, 85748 Garching (Germany); Matsukawa, Makoto; Shimada, Katsuhiro; Kawamata, Yoichi; Takechi, Manabu [Japan Atomic Energy Agency, Naka Fusion Institute, 801-1 Mukoyama, Naka, Ibaraki 311-019 (Japan)

2015-10-15

Highlights: • In JT-60SA, a power supply system (RWM-PS) will feed 18 coils to control the RWMs. • One power amplifier per coil will follow an arbitrary real-time reference. • Very fast dynamics is required (current bandwidth: 3 kHz; latency: 50 μs). • The requirements of the RWM-PS are updated and design solutions discussed. • The reference design of the RWM-PS is based on H-bridges operated at 20 – 30 kHz. - Abstract: The mission of JT-60SA, the satellite Tokamak under construction in Naka (Japan), includes the attainment of steady-state high-beta plasmas. For this purpose, an active control system based on 18 in-vessel sector coils (SC) is foreseen to suppress the resistive wall modes (RWM). Each coil will be independently fed by a dedicated converter, rated for 300 A and 240 V, which has to produce the required current/voltage following in real time the reference provided by the JT-60SA MHD Controller. To minimize the current rating, these converters shall be sufficiently fast to avoid an excessive growth of the RWM. This requires a very high dynamic performance, largely beyond that of standard industrial applications. This paper firstly reports the latest results of the studies on the requirements of the RWM active control system. Then, the reference design of the power supply system is presented, including the ac/dc conversion stage, the fast converters and the control section. The advantages of the proposed scheme are discussed and the main electrical parameters are presented.

Hydrogen isotope distributions and retentions in the inner divertor tile of JT-60U

Energy Technology Data Exchange (ETDEWEB)

Oya, Y. [Radioisotope Center, University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-0032 (Japan)]. E-mail: yoya@ric.u-tokyo.ac.jp; Hirohata, Y. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Tanabe, T. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Shibahara, T. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Kimura, H. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Oyaidzu, M. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Arai, T. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Masaki, K. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Gotoh, Y. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Okuno, K. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Miya, N. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Hino, T. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Tanaka, S. [Graduate School of Engineering, University of Tokyo, Tokyo 113-8656 (Japan)

2005-11-15

Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature.

Hydrogen isotope distributions and retentions in the inner divertor tile of JT-60U

International Nuclear Information System (INIS)

Oya, Y.; Hirohata, Y.; Tanabe, T.; Shibahara, T.; Kimura, H.; Oyaidzu, M.; Arai, T.; Masaki, K.; Gotoh, Y.; Okuno, K.; Miya, N.; Hino, T.; Tanaka, S.

2005-01-01

Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature

A Review of Fusion and Tokamak Research Towards Steady-State Operation: A JAEA Contribution

Directory of Open Access Journals (Sweden)

Mitsuru Kikuchi

2010-11-01

Full Text Available Providing a historical overview of 50 years of fusion research, a review of the fundamentals and concepts of fusion and research efforts towards the implementation of a steady state tokamak reactor is presented. In 1990, a steady-state tokamak reactor (SSTR best utilizing the bootstrap current was developed. Since then, significant efforts have been made in major tokamaks, including JT-60U, exploring advanced regimes relevant to the steady state operation of tokamaks. In this paper, the fundamentals of fusion and plasma confinement, and the concepts and research on current drive and MHD stability of advanced tokamaks towards realization of a steady-state tokamak reactor are reviewed, with an emphasis on the contributions of the JAEA. Finally, a view of fusion energy utilization in the 21st century is introduced.

Design of the movable limiters for JT-60

International Nuclear Information System (INIS)

Takashima, Tetsuo; Yamamoto, Masahiro; Nakamura, Hiroo; Ohkubo, Minoru; Ohta, Mitsuru

1976-07-01

Two fast-acting movable rail limiters will be used in JT-60 to suppress skin effect of the plasma current with a large radius. They travel safely through a stroke of about 1 m for 0.1 sec in the build-up phase of plasma current. The movable limiter system consists of a driving mechanism, a vacuum seal, a bearing used at high temperatures in a vacuum, a molybdenum rail limiter weighing 200 kg and its auxiliary members. Many problems are involved in construction of the system because the design specifications exceed the present technology. Described are design of the movable limiter system for JT-60 and problems in the mechanical, electrical and vacuum aspects. (auth.)

The design study of the JT-60SU device. No. 3. The superconductor-coils of JT-60SU

International Nuclear Information System (INIS)

Ushigusa, Kenkichi; Mori, Katsuharu; Nakagawa, Syouji

1997-03-01

The superconducting coil systems and the cryogenic system for the JT-60 Super Upgrade (JT-60SU) has been designed. Both Nb 3 Al and NbTi as a superconducting wire material are employed in the toroidal coils (D-shaped 18 coils) to realize a high field magnet with a low cost. Significant reduction of the coil weight (150 tons/coil) without losing the coil rigidity has been achieved by connecting two toroidal coils with shear panels. Validity of this design is confirmed by the detailed structural analysis and thermohydraulic analysis. The poloidal coil system consists of 4 central solenoid coils with (NbTi) 3 Sn and 6 outer equilibrium field coils with NbTi. This system has an enough capability to supply the flux of 170Vs to produce a 10MA discharge with 200s of flat-top and to make various plasma configurations. The construction procedure of the poloidal coil system is also established under the constraint of the JT-60 site. Two sets of race-track shaped superconducting coils mounted on the top of the machine is designed to compensate the error field inside the vessel by supplying helical (m=2/n=1) magnetic field. By using cryogenic system with a 36kW of cooling capacity, the total cold weight of around 4000tons can be cooled down to 4.5K within one month, and steady heat load of 6.5kW and transient heat load of 9.0MJ can be removed within 30 minutes of discharge repetition rate. (author)

Radiation loss and global energy balance of ohmically heated divertor discharge in JT-60 tokamak

International Nuclear Information System (INIS)

Koide, Yoshihiko; Yamada, Kimio; Yoshida, Hidetoshi; Nakamura, Hiroo; Niikura, Setsuo; Tsuji, Shunji

1986-03-01

Divertor experiment in JT-60 with a small divertor chamber has been successfully performed up to 1.6 MA discharge. Several divertor effects were experimentally confirmed as follows. Radiation loss in main plasma saturates with the increase of plasma current and its ratio to the input power is about 20 % at 1.5 MA. The rest of input power is exhausted into the divertor chamber and a half of it is dissipated as the radiation loss. Impurity accumulation is not observed during a few sec without internal MHD activity and gross impurity confinement time is several hundred msec. (author)

Waveguide circuit for LHRF heating in 'JT-60'

International Nuclear Information System (INIS)

Uehara, Kazuya; Saegusa, Mikio; Mizuno, Takenori; Sano, Keigo; Hara, Mitsuru; Oishi, Isamu; Kanai, Takao.

1985-01-01

As the heating method for attaining the critical condition in the critical plasma experiment apparatus 'JT-60' in the Japan Atomic Energy Research Institute, in addition to Joule heating, as the second heating method, neutral beam injection heating and high frequency heating have been adopted. For this high frequency heating, several tens to 200 MHz band of ICRF heating, several hundreds MHz to several GHz band of LHRF heating and several tens to 200 GHz band of ECR heating were considered, and in the JT-60, 100 MHz band (ICRF) and 2 GHz band (LHRF) have been adopted. Furukawa Electric Co., Ltd. has engaged in the development and manufacture of the waveguides of transmission system used for this high frequency heating through NEC Corp. This high frequency heating is to heat plasma by injecting high frequency radio waves into plasma proper, and reaches 10 MW for the whole high frequency heating. The system efficiently transmitting the radio waves of large power from a Klystron as a high frequency source to the JT-60 is the transmission system. The outline of the waveguides of the 2 GHz band transmission system and the individual performance of respective waveguides are reported. (Kako, I.)

Recent ion cyclotron range of frequencies experiments in JT-60U

International Nuclear Information System (INIS)

Kimura, H.; Fujii, T.; Saigusa, M.; Moriyama, S.; Sato, M.; Nemoto, M.; Kondoh, T.; Hamamatsu, K.

1995-01-01

Recent results on the minority ion second harmonic heating on JT-60U are presented. Maximum coupled power reached 6.4MW. Good antenna-plasma coupling capability and a small fraction (less than 10%) of an incremental radiation loss to r.f. power are confirmed. Power absorption rate increases with increasing r.f. power and is saturated around unity at r.f. powers higher than 3MW. The sawtooth stabilization by minority ion second harmonic heating was realized over a wide parameter range, i.e. I P =0.9MA-4MA, q 95 =2.3-8.6, n e =(1.3-5)x10 19 m -3 and P IC ≥2.2MW. A figure of merit V P left angle n e right angle /P tot for efficiency of the sawtooth stabilization is about 50% higher than those in other devices where fundamental resonance minority ion heating is employed. The longest stable period reached 2.33s. Attainable sawtooth-free periods scale with the resistive diffusion time. It was found that the energy confinement is further improved by 25% during the reheating phase after the giant sawtooth crash. The electron temperature profile became more peaked at the improved confinement phase. Those phenomena were observed only in low q discharges (q 95 ≤2.9). ((orig.))

Core density fluctuations in reverse magnetic shear plasmas with internal transport barrier on JT-60U

International Nuclear Information System (INIS)

Nazikian, R.; Shinohara, K.; Yoshino, R.; Fujita, T.; Shirai, H.; Kramer, G.T.

1999-01-01

First measurements of the radial correlation length of density fluctuations in JT-60U plasmas with internal transport barrier (ITB) is reported. The measurements are obtained using a newly installed correlation reflectometer operating in the upper X-mode. Before transport barrier formation in the low beam power current ramp-up phase of the discharge, reflectometer measurements indicate density fluctuation levels n-tilde/n∼0.1-0.2% and radial correlation lengths 2-3 cm (k r p i ≤0.5) in the central plasma region (r/a r p i ∼3. However, fluctuation levels are considerably higher than measured near the magnetic axis. Reflectometer measurements obtained at the foot of the ITB also indicate high fluctuation levels compared to measurements in the central region of the discharge. (author)

JT-60 database system, 2

International Nuclear Information System (INIS)

Itoh, Yasuhiro; Kurihara, Kenichi; Kimura, Toyoaki.

1987-07-01

The JT-60 central control system, ''ZENKEI'' collects the control and instrumentation data relevant to discharge and device status data for plant monitoring. The former of the engineering data amounts to about 3 Mbytes per shot of discharge. The ''ZENKEI'' control system which consists of seven minicomputers for on-line real-time control has little performance of handling such a large amount of data for physical and engineering analysis. In order to solve this problem, it was planned to establish the experimental database on the Front-end Processor (FEP) of general purpose large computer in JAERI Computer Center. The database management system (DBMS), therefore, has been developed for creating the database during the shot interval. The engineering data are shipped up from ''ZENKEI'' to FEP through the dedicated communication line after the shot. The hierarchical data model has been adopted in this database, which consists of the data files with tree structure of three keys of system, discharge type and shot number. The JT-60 DBMS provides the data handling packages of subroutines for interfacing the database with user's application programs. The subroutine packages for supporting graphic processing and the function of access control for security of the database are also prepared in this DBMS. (author)

Long-pulse hybrid scenario development in JT-60U

International Nuclear Information System (INIS)

Oyama, N.; Isayama, A.; Matsunaga, G.; Suzuki, T.; Takenaga, H.; Sakamoto, Y.; Nakano, T.; Kamada, Y.; Ide, S.

2009-01-01

The performance and sustained duration of long-pulse discharges for the 'ITER hybrid scenario' have been improved in JT-60U. The modification of power supply systems for three perpendicular neutral beam (NB) injections provides a long period of central NB heating up to 30 s, which is important for keeping the internal transport barrier (ITB). The peaked density profile in the core plasma can be maintained even when the density at the pedestal increased in the latter phase of the discharge due to the increase in the divertor recycling. Then, the peaked pressure profile attributed to the ITB can be kept constant through the discharge with the peaked power deposition profile. In these long-pulse discharges, MHD activity with toroidal mode number n = 1 is observed even when neoclassical tearing modes (NTMs) are suppressed. When the amplitude of the mode in the peripheral region becomes large, the pedestal pressure is degraded. The mode amplitude is sensitive to the toroidal magnetic field (or edge safety factor) and heating power. After the adjustment of the toroidal magnetic field so as to reduce the mode amplitude, a high normalized beta (β N ) of 2.6 and a high thermal confinement enhancement factor (H H98(y,2) > 1) are sustained for 25 s (∼14τ R , where τ R is the current diffusion time) under the ITER relevant small toroidal rotation condition. The peaked pressure profile in low safety factor plasma (safety factor at 95% flux surface q 95 ∼ 3.2) is stable against NTMs up to β N ∼ 3. A high β N H H98(y,2) of 2.6 gives a high G-factor ( β N H H98(y,2) /q 95 2 ) of 0.25 and a peaked pressure profile gives a large bootstrap current fraction (f BS > 0.43).

Thermal properties of redeposition layers in the JT-60U divertor region

International Nuclear Information System (INIS)

Ishimoto, Y.; Gotoh, Y.; Arai, T.; Masaki, K.; Miya, N.; Oyama, N.; Asakura, N.

2006-01-01

Thermal properties of the redeposition layer on the inner plate of the W-shaped divertor of JT-60U have been measured with laser flash method so as to estimate transient heat loads onto the divertor. Morphology analysis of the redeposition layer was conducted with a scanning electron microscope. Measurement of a redeposition layer sample of more than 200 μm thick, which had been produced near the most frequent striking point, showed following results: (1) the bulk density of the redeposition layer is about half of that of carbon fiber composite material; (2) the specific heat of the layer is roughly equal to that of the isotropic graphite; (3) the thermal conductivity of the redeposition layer is two orders of magnitude smaller than that of the carbon fiber composite. This low thermal conductivity of the redeposition layer is considered to be caused by a low graphitization degree of the redeposition layer. The difference between the divertor heat loads and the loss of the plasma stored energy becomes smaller taking account of thermal properties of the redeposition layer on the inner divertor, whereas estimated heat loads due to the ELMs is still larger than the loss. This is probably caused by the poloidal distribution of the thermal properties

Development of multilayer piezoelectric actuator valve for JT-60

International Nuclear Information System (INIS)

Miyo, Yasuhiko; Hiratsuka, Hajime; Masui, Hiroshi; Hosogane, Nobuyuki; Miya, Naoyuki

2001-11-01

In order to improve the gas injection valve used for the operation of JT-60, a new type of valve (multilayer piezoelectric actuator valve) was developed. The conventional valve (bimorph piezoelectric valve) has been used for 15 years since the beginning of experimental operation in April, 1985. However, it came to be hard to keep the performance of the valve because of the deterioration of the driving source, i.e. piezoelectric element. Developments of the new valve were carried out based on experiences through experimental operations in JT-60. Requirements for the design are: (1) to be hard structure for making a sheet leak, (2) to allow a repair work at atmosphere side without an air vent of the vacuum vessel, (3) to be more smaller and lighter compared with the conventional one, and (4) to have a high maintenance efficiency by utilizing of the commercial piezoelectric elements and power supplies. The newly developed valve was examined with various tests such as gas flow characteristic test, high magnetic field proof test, high temperature proof test and gas flow rate test for aged deterioration. Results, confirm that the performance of the valve is applicable for JT-60 operations. (author)

Conceptual radiation shielding design of superconducting tokamak fusion device by PHITS

International Nuclear Information System (INIS)

Sukegawa, Atsuhiko M.; Kawasaki, Hiromitsu; Okuno, Koichi

2010-01-01

A complete 3D neutron and photon transport analysis by Monte Carlo transport code system PHITS (Particle and Heavy Ion Transport code System) have been performed for superconducting tokamak fusion device such as JT-60 Super Advanced (JT-60SA). It is possible to make use of PHITS in the port streaming analysis around the devices for the tokamak fusion device, the duct streaming analysis in the building where the device is installed, and the sky shine analysis for the site boundary. The neutron transport analysis by PHITS makes it clear that the shielding performance of the superconducting tokamak fusion device with the cryostat is improved by the graphical results. From the standpoint of the port streaming and the duct streaming, it is necessary to calculate by 3D Monte Carlo code such as PHITS for the neutronics analysis of superconducting tokamak fusion device. (author)

Full power in the European tokamak

International Nuclear Information System (INIS)

Lallia, P.P.; Hugon, M.

1987-01-01

A new research campaign begins at Jet (Abingdon, UK). At this occasion, authors review and compare the performances of the three big Tokamaks that are currently in competition: Jet, JT60 and TFTR, insisting upon the European one. Conditions of ignition are reviewed together and energy losses are specified. Magnetic configurations used in tokamaks are shown, together with the technological responses brought these last years

Hydrogen isotope behavior in the first wall of JT-60U after deuterium plasma operation

International Nuclear Information System (INIS)

Oya, Y.; Tanabe, T.; Oyaidzu, M.; Shibahara, T.; Sugiyama, K.; Yoshikawa, A.; Onishi, Y.; Hirohata, Y.; Ishimoto, Y.; Yagyu, J.; Arai, T.; Masaki, K.; Okuno, K.; Miya, N.; Tanaka, S.

2007-01-01

Retention of hydrogen isotopes in the carbon (isotropic graphite) first wall tiles of JT-60U was studied by secondary ion mass spectrometry and thermal desorption spectroscopy. The surface morphology and erosion/deposition profiles of the tiles were characterized using scanning electron microscope and X-ray photoelectron spectroscopy. The upper area is mainly eroded, while the bottom area of the inboard wall is dominated by deposition. In contrast to the divertor area, hydrogen isotope retention in the eroded wall area was generally larger than that in the deposition dominated area. Measured near surface concentrations of hydrogen isotopes in the wall tiles, as well as the D/H ratios, were a little higher than those in the divertor area. This indicates direct implantation of high-energy D from NBI into the first wall. The lower temperature of the first wall relative to the divertor tiles would reduce desorption and/or replacement of implanted D by subsequent D or H impingement

Divertor pumping system with NBI cryopump for JT-60

International Nuclear Information System (INIS)

Akino, Noboru; Kuriyama, Masaaki; Ohga, Tokumichi; Seki, Hiroshi; Tanai, Yutaka

1998-08-01

The pumping system for JT-60 W-shape divertor with the NBI cryopump have been developed. The pumping speed achieved in the divertor region was 13-15 m 3 /s for deuterium gas with three units of the NBI cryopumps. In a simulation experiment of helium ash exhaust through the divertor, pumping of a mixed gas of helium and deuterium has been demonstrated using the NBI cryosorption pumps covered with an argon condensed layer. Control of neutral particle pressure in the divertor region became possible by having remodeled an aperture of the existing fast shutter, which is installed between the JT-60 vacuum vessel and NBI beam-line, to be regulated. (author)

Review of JT-60 experiment (March 1986)

International Nuclear Information System (INIS)

1986-11-01

Results of JT-60 experiment with ohmic heating in March 1986 are summarized. A maximum plasma current of 2 MA, an average plasma density of 5.7 x 10 19 m -3 and energy confinement time of 0.4 - 0.5 sec were obtained. Detailed characteristics of ohmic plasmas are discussed. (author)

Research and development of the JAERI large tokamak (JT-60), (4)

International Nuclear Information System (INIS)

Takashima, Tetsuo; Shimizu, Masatsugu; Ohta, Mitsuru; Minaguchi, Tadayoshi; Maeda, Hideto.

1978-01-01

A pair of fast-acting movable rail limiters are to be installed in the vacuum chamber of JT-60 to suppress skin current in the plasma column. They should travel across the vacuum chamber over a stroke of about 1 m in 0.1 sec in the build-up phase of the plasma current. Each movable limiter system consists of a drive mechanism, a vacuum seal, a bearing usable at high temperatures in a vacuum, a molybdenum rail limiter head and its auxiliary members. Various engineering problems are involved in constructing such a system because the design specifications outlined above exceed the present technology. A full-scale movable limiter, therefore, was designed, constructed and then put to mechanical, electrical and vacuum-technological tests. The model features a hydraulic drive mechanism with servovalves to control the oil flow. A special vacuum seal allowing a movement at high speeds was developed. It consists of welded bellows jointed together and connected to a pantograph at the joints. It allows uniform expansion of each bellows at high speeds. Molybdenum disulphide with 20% Ta is chosen as the most suitable bearing material after conducting tests on various bearing materials. The overall test of the model showed that its specifications were met with satisfactory reliability and reproducibility. Furthermore, the endurance test demonstrated that it functioned reliably over 50,000 times of operation. (author)

JT-60 configuration parameters for feedback control determined by regression analysis

Energy Technology Data Exchange (ETDEWEB)

Matsukawa, Makoto; Hosogane, Nobuyuki; Ninomiya, Hiromasa (Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment)

1991-12-01

The stepwise regression procedure was applied to obtain measurement formulas for equilibrium parameters used in the feedback control of JT-60. This procedure automatically selects variables necessary for the measurements, and selects a set of variables which are not likely to be picked up by physical considerations. Regression equations with stable and small multicollinearity were obtained and it was experimentally confirmed that the measurement formulas obtained through this procedure were accurate enough to be applicable to the feedback control of plasma configurations in JT-60. (author).

JT-60 configuration parameters for feedback control determined by regression analysis

International Nuclear Information System (INIS)

Matsukawa, Makoto; Hosogane, Nobuyuki; Ninomiya, Hiromasa

1991-12-01

The stepwise regression procedure was applied to obtain measurement formulas for equilibrium parameters used in the feedback control of JT-60. This procedure automatically selects variables necessary for the measurements, and selects a set of variables which are not likely to be picked up by physical considerations. Regression equations with stable and small multicollinearity were obtained and it was experimentally confirmed that the measurement formulas obtained through this procedure were accurate enough to be applicable to the feedback control of plasma configurations in JT-60. (author)

Atomic and molecular processes in JT-60U divertor plasmas

Energy Technology Data Exchange (ETDEWEB)

Takenaga, H.; Shimizu, K.; Itami, K. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1997-01-01

Atomic and molecular data are indispensable for the understanding of the divertor characteristics, because behavior of particles in the divertor plasma is closely related to the atomic and molecular processes. In the divertor configuration, heat and particles escaping from the main plasma flow onto the divertor plate along the magnetic field lines. In the divertor region, helium ash must be effectively exhausted, and radiation must be enhanced for the reduction of the heat load onto the divertor plate. In order to exhaust helium ash effectively, the difference between behavior of neutral hydrogen (including deuterium and tritium) and helium in the divertor plasma should be understood. Radiation from the divertor plasma generally caused by the impurities which produced by the erosion of the divertor plate and/or injected by gas-puffing. Therefore, it is important to understand impurity behavior in the divertor plasma. The ions hitting the divertor plate recycle through the processes of neutralization, reflection, absorption and desorption at the divertor plates and molecular dissociation, charge-exchange reaction and ionization in the divertor plasma. Behavior of hydrogen, helium and impurities in the divertor plasmas can not be understood without the atomic and molecular data. In this report, recent results of the divertor study related to the atomic and molecular processes in JT-60U were summarized. Behavior of neural deuterium and helium was discussed in section 2. In section 3, the comparisons between the modelling of the carbon impurity transport and the measurements of C II and C IV were discussed. In section 4, characteristics of the radiative divertor using Ne puffing were reported. The new diagnostic method for the electron density and temperature in the divertor plasmas using the intensity ratios of He I lines was described in section 5. (author)

Density fluctuation measurement at edge and internal transport barriers in JT-60U

International Nuclear Information System (INIS)

Oyama, N; Bruskin, L G; Takenaga, H; Shinohara, K; Isayama, A; Ide, S; Sakamoto, Y; Suzuki, T; Fujita, T; Kamada, Y; Miura, Y

2004-01-01

A new analytical method using a combination of the O-mode reflectometer and a time-dependent two-dimensional full-wave simulation code has been developed for the quantitative evaluation of density fluctuations in JT-60U. Two statistical parameters of the reflectometer signals, fluctuation index (F) and elongation factor (χ), are introduced as measures of the fluctuation amplitude (γ) and the width of the poloidal wave number spectrum (k θ0 ). This method is applied to the edge transport barrier (ETB) and internal transport barrier (ITB). At the transition to the ELM free H-mode phase, analysis suggests that the density fluctuation level reduced from 1.9-3.2% to 0.29-0.44%, while the value of k θ0 changed from 1.6-2.0 to 0.77-0.81 cm -1 in the ETB region. On the other hand, the amplitude of the density fluctuation was evaluated as 1.0-2.0% at the ITB region, even after the formation of the box type ITB. Instead, when a pellet was injected into the plasma with a box type ITB as an external perturbation, a remarkable change in the frequency spectrum was observed. Analysis suggests a reduction in the density fluctuation level to 0.4-0.6% after the pellet injection

Magnetic confinement experiment. I: Tokamaks

International Nuclear Information System (INIS)

Goldston, R.J.

1995-08-01

Reports were presented at this conference of important advances in all the key areas of experimental tokamak physics: Core Plasma Physics, Divertor and Edge Physics, Heating and Current Drive, and Tokamak Concept Optimization. In the area of Core Plasma Physics, the biggest news was certainly the production of 9.2 MW of fusion power in the Tokamak Fusion Test Reactor, and the observation of unexpectedly favorable performance in DT plasmas. There were also very important advances in the performance of ELM-free H- (and VH-) mode plasmas and in quasi-steady-state ELM'y operation in JT-60U, JET, and DIII-D. In all three devices ELM-free H-modes achieved nTτ's ∼ 2.5x greater than ELM'ing H-modes, but had not been sustained in quasi-steady-state. Important progress has been made on the understanding of the physical mechanism of the H-mode in DIII-D, and on the operating range in density for the H-mode in Compass and other devices

Fast dynamics of Type I ELM and transport of ELM pulse in JT-60U

International Nuclear Information System (INIS)

Oyama, N.

2002-01-01

The mitigation of the large ELM heat load on the divertor target is one of the most important issues to be overcome on ITER. Since the ELM heat load strikes the divertor target not as a time-averaged load but as an instantaneous heat pulse, the evaluation of both ELM energy, and the time scale of the collapse and transport is very important. In JT-60U, the detailed dynamic behaviors of the collapse were measured using O-mode reflectometer. The duration of the collapse was within 0.35 ms and the lost pedestal density was recovered quickly within 0.5 ms. The collapse reached 10 cm inside the separatrix, which corresponds to twice the pedestal width of 5 cm. Dedicated edge density measurements on high- and low-field side revealed the poloidal asymmetry of the collapse of density pedestal for the first time. The measurement of SOL flow and heat load to the divertor target by using SOL Mach probe and IRTV showed that convective transport of the SOL plasma gave large contribution to the ELM heat deposition process. (author)

Structural evaluation of a compact, semi-closed W-shaped divertor system for JT-60U

Energy Technology Data Exchange (ETDEWEB)

Onozuka, M.; Morimoto, M.; Inoue, M.; Umakoshi, T.; Shimizu, K. [Mitsubishi Heavy Ind., Ltd., Yokohama (Japan). Adv. Reactor and Nucl. Fuel Cycle Eng. Dept.; Sakurai, S.; Hosogane, N.; Masaki, K. [Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki-pref. 311-0193 (Japan)

1999-03-01

A compact, semi-closed W-shaped divertor system has been designed, fabricated and installed in the JT-60U to replace the open divertor system. The new system consists of inclined divertors, a dome and baffles. To meet the structural requirements, a segmented structure with an electrically insulated flexible gas seal was applied. Using FEM codes, the system`s structural integrity was confirmed for the plasma disruptions by electromagnetic and structural analyses, which take into account the effect of halo currents. Substantial reduction of induced electromagnetic forces is attained in the divertor system due to the electrical insulation used for the gas seal structure. In addition, because of its segmented structure, the induced electromagnetic forces on each component unit are found to be limited. The maximum stress intensities and their ranges are obtained within allowable values. Thermal stress arising from the temperature difference between the divertor system and the vacuum vessel during the baking operation is also satisfactory. Furthermore, thermal and thermal stress analyses showed that the plasma facing components have sufficient structural integrity. (orig.) 18 refs.

Structural evaluation of a compact, semi-closed W-shaped divertor system for JT-60U

International Nuclear Information System (INIS)

Onozuka, M.; Morimoto, M.; Inoue, M.; Umakoshi, T.; Shimizu, K.

1999-01-01

A compact, semi-closed W-shaped divertor system has been designed, fabricated and installed in the JT-60U to replace the open divertor system. The new system consists of inclined divertors, a dome and baffles. To meet the structural requirements, a segmented structure with an electrically insulated flexible gas seal was applied. Using FEM codes, the system's structural integrity was confirmed for the plasma disruptions by electromagnetic and structural analyses, which take into account the effect of halo currents. Substantial reduction of induced electromagnetic forces is attained in the divertor system due to the electrical insulation used for the gas seal structure. In addition, because of its segmented structure, the induced electromagnetic forces on each component unit are found to be limited. The maximum stress intensities and their ranges are obtained within allowable values. Thermal stress arising from the temperature difference between the divertor system and the vacuum vessel during the baking operation is also satisfactory. Furthermore, thermal and thermal stress analyses showed that the plasma facing components have sufficient structural integrity. (orig.)

Power injection performance of the LH antenna tipped with carbon grills in JT-60U

International Nuclear Information System (INIS)

Ishii, Kazuhiro; Seki, Masami; Shinozaki, Shinichi; Hasegawa, Koichi; Hiranai, Shinichi; Suzuki, Sadaaki; Sato, Fumiaki; Moriyama, Shinichi; Yokokura, Kenji

2007-07-01

The lower hybrid (LH) antenna in JT-60U has interaction with plasmas because it should be close to them in order to inject effectively radio frequency (RF) power into them. As a result, it has been a serious problem that the antenna mouth made of stainless steels was damaged due to excessive heat loads of plasmas and RF breakdowns. To solve the problem, a heat-resistant LH antenna was developed tipping carbon grills with fairly high heat resistance on the antenna mouth, and therefore reduction in damages on the mouth was expected. Power injection into plasmas was firstly performed with the heat-resistant antenna. RF conditioning was done carefully in the initial phase because RF breakdown due to outgassing from the grills might be occurred. After sufficient degassing was done through RF conditioning, RF power of about 1.6 MW x 10 sec injection was successfully injected to plasmas. Moreover it was demonstrated that it had comparably high plasma current drive capability (about 1.6 x 10 19 A/W/m 2 ), required as a current drive LH antenna. (author)

Lower hybrid current drive in tokamak plasmas

International Nuclear Information System (INIS)

Ushigusa, Kenkichi

1999-03-01

Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar e - 10 20 m -3 , ALCATOR-C) and the highest current drive efficiency (η CD = 3.5x10 19 m -2 A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)

Developmental prototype for replacement of JT-60 timing system

International Nuclear Information System (INIS)

Akasaka, H.; Kawamata, Y.; Yonekawa, I.

2004-01-01

The present CAMAC based timing system has been used for synchronizing sequential events of the discharge and the data collection of the interesting JT-60U experiment plasma phenomena. However, a more flexible and sophisticated state-of-the-art timing system now is required to realize advanced plasma control with minimal maintenance costs. In this context, the versa module Europe (VME-bus) system with a high-speed data communication network using reflective memory (RM) modules and user-friendly application software based on MATLAB TM tools has been selected to develop the new prototype timing system. In the ZENKEI, the supervisory control system of the JT-60, the supervisory timing system provides the 50-μs master clock pulses, the various timing signal preparation logic, which is built into the digital signal processing (DSP) module in conjunction with the discharge sequence event signals, and the 6.2 MB/s high-speed communication data link provided by the RM module. Except the clock pulse generator (CPG) module, no other special timing module is necessary for this new timing system. The timing signal is prepared by software logic in conjunction with sequential events and the preset timer, is transferred to the subsystems through the RM module, where it is synchronized to the 50-μs clock pulses. The timing system of the subsystems also consists of hardware similar in structure to the ZENKEI timing system. The fundamental timing system configuration, the necessary functions, and the preliminary test results of the prototype system are reported in this presentation

Runaway acceleration during magnetic reconnection in tokamaks

International Nuclear Information System (INIS)

Helander, P; Eriksson, L-G; Andersson, F

2002-01-01

In this paper, the basic theory of runaway electron production is reviewed and recent progress is discussed. The mechanisms of primary and secondary generation of runaway electrons are described and their dynamics during a tokamak disruption is analysed, both in a simple analytical model and through numerical Monte Carlo simulation. A simple criterion for when these mechanisms generate a significant runaway current is derived, and the first self-consistent simulations of the electron kinetics in a tokamak disruption are presented. Radial cross-field diffusion is shown to inhibit runaway avalanches, as indicated in recent experiments on JET and JT-60U. Finally, the physics of relativistic post-disruption runaway electrons is discussed, in particular their slowing down due to emission of synchrotron radiation, and their ability to produce electron-positron pairs in collisions with bulk plasma ions and electrons

Heat deposition on the first wall due to ICRF-induced loss of fast ions in JT-60U

International Nuclear Information System (INIS)

Kusama, Y.; Tobita, K.; Kimura, H.; Hamamatsu, K.; Fujii, T.; Nemoto, M.; Saigusa, M.; Moriyama, S.; Tani, K.; Koide, Y.; Sakasai, A.; Nishitani, T.; Ushigusa, K.

1995-01-01

In JT-60U, the heat deposition on the first wall due to the ICRF-induced loss of fast ions was investigated by changing the position of the resonance layer in the ripple-trapping region. A heat spot appears on the first wall of the same major radius as the resonance layer of the ICRF waves. The broadening of the heat spot in the major radius direction is consistent with that of the resonance layer due to the Doppler broadening. The heat spot is considered to be formed by the ICRF-induced ripple-trapped loss of fast ions. Although the total ICRF-induced loss power to the heat spot is as low as 2% of the total ICRF power, the additional heat flux will become a new issue because of the localized heat deposition on the first wall. ((orig.))

Dynamic response of the JT-60 vacuum vessel under the electromagnetic forces

International Nuclear Information System (INIS)

Takatsu, H.; Shimizu, M.; Ohta, M.

1982-01-01

Dynamic response analyses of the JAERI Tokamak 60 (JT-60) vacuum vessel were carried out under three kinds of saddle-like electromagnetic forces. In the analysis, the dynamic response of the bellows was obtained by dividing it into three components; the first, caused by the forced deflection due to the displacement of an adjacent rigid ring; the second, caused by inertia force; and the third, caused by a saddle-like electromagnetic force. Eigenvalue analyses showed that the 20th mode is a typical rotation mode of the rigid ring around the major radius with a natural frequency of 46.3 Hz. From the results of the dynamic response analyses, the maximum displacement response of the rigid ring was 3.1 mm and remarkable dynamic response was observed in the case of plasma disruption with a time constant of 1 ms. In cases of start-up of the plasma current and plasma disruption with a time constant of 50 ms, the rigid ring vibrates quasi-statically. It is clear that the dynamic behavior of the vacuum vessel is governed mainly by the saddle-like electromagnetic force, with a smaller effect of the inverse saddle-like electromagnetic force on the dynamic response of the vacuum vessel. (orig.)

Diagnostic system for passive charge-exchange particle measurements on JT-60

International Nuclear Information System (INIS)

Nemoto, Masahiro; Tobita, Kenji; Kusama, Yoshinori; Takeuchi, Hiroshi

1988-01-01

In order to measure energy distributions of the charge-exchange neutral particles in the JT-60 experiments, a compact size electrostatic energy analyzer which the measurable energy range is from 1 keV to 100 keV is developed successfully. Compactness of an analyzer is accomplished by setting an accelerator between a gas stripping cell and a deflector of 45deg injection type. The calibration of the analyzer was carried out owing to confirm the capability of energy analysis and stripping efficiency. This analyzer was applied to measure the energy distribution in additionally heated plasmas in JT-60. The usefullness of the analyzer was confirmed. (author)

Density limit in JT-60

International Nuclear Information System (INIS)

Kamada, Yutaka; Hosogane, Nobuyuki; Hirayama, Toshio; Tsunematsu, Toshihide

1990-05-01

This report studies mainly the density limit for a series of gas- and pellet-fuelled limiter discharges in JT-60. With the pellet injection into high-current/low-q (q(a)=2.3∼2.5) discharges, the Murakami factor reaches up to 10∼13 x 10 19 m -2 T -1 . The values are about factors of 1.5∼2.0 higher than those for usual gas-fuelled discharges. The pellet injected discharges have high central density, whereas the electron density in the outer region (a/2 abs and n e 2 (r=50 cm) x Z eff (r=50 cm). (author)

JT-60 database system, 1

International Nuclear Information System (INIS)

Kurihara, Kenichi; Kimura, Toyoaki; Itoh, Yasuhiro.

1987-07-01

Naturally, sufficient software circumstance makes it possible to analyse the discharge result data effectively. JT-60 discharge result data, collected by the supervisor, are transferred to the general purpose computer through the new linkage channel, and are converted to ''database''. Datafile in the database was designed to be surrounded by various interfaces. This structure is able to preserve the datafile reliability and does not expect the user's information about the datafile structure. In addition, the support system for graphic processing was developed so that the user may easily obtain the figures with some calculations. This paper reports on the basic concept and system design. (author)

Ion beam dump for JT-60 NBI

International Nuclear Information System (INIS)

Kuriyama, Masaaki; Horiike, Hiroshi; Matsuda, Shinzaburo; Morita, Hiroaki; Shibanuma, Kiyoshi

1981-10-01

The design of the active cooling type ion beam dump for JT-60 NBI which receives the total beam power of 5.6 MW for 10 sec continuously is described. It is composed of array of many finned tubes which is made of oxygen free copper with 0.2% silver content. The safety margin against thermal and mechanical troubles is estimated by the heat transfer and the thermal stress calculation. (author)

Prospective performances in JT-60SA towards the ITER and DEMO relevant plasmas

International Nuclear Information System (INIS)

Tamai, H.; Fujita, T.; Kikuchi, M.

2006-01-01

JT-60SA, the former JT-60SC and NCT, a superconducting tokamak positioned as the satellite machine of ITER, collaborating with Japan and EU fusion community, aims at contribution to ITER and DEMO through the demonstration of advanced plasma operation scenario and the plasma applicability test with advanced materials. After the discussions in JA-EU Satellite Tokamak Working Group in 2005, the increased heating power, higher heat removal capacity for the plasma facing components, improvement of the radiation shielding, the remote handling system for the maintenance of in-vessel components, and related equipment are planed to be additionally installed. With such full equipment towards the increased heating power of 41 MW (34 MW-NBI and 7 MW-ECH) with 100 s, the prospective plasma performances, analysed by the equilibrium and transport analysis codes, are rather improved in the view point of the contribution to ITER and DEMO relevant research. Accessibility for higher heating power in a higher density region enables the lower normalized Larmor radius and normalized collision frequency close to the reactor relevant plasma with the ITER-similar configuration of single null divertor plasma with the aspect ratio of A = 3.1, elongation of k95 = 1.7, triangularity of d95 (q95) in the plasma current of I p = 3.5 MA, toroidal magnetic field of B T = 2.59 T and the major radius of Rp=3.16 m. The increases in the electron temperature, beam driven and bootstrap current fraction by the increase of the power of Negative ion based NBI (10 MW) reduce the loop voltage and contribute to increase the maximum plasma current of ITER similar shape. Full non-inductive current drive controllability is also extended into the high density and high plasma current operation towards high beta plasma. Flexibility in aspect ratio and shape parameter is kept the same as NCT, i.e. a double null divertor plasma with A = 2.6, k95 = 1.83, d95 = 0.57, I p = 5.5 MA, B T = 2.72 T, and R p = 3.01 m which

Profile formation and sustainment of autonomous tokamak plasma with current hole configuration

International Nuclear Information System (INIS)

Hayashi, N.; Takizuka, T.; Ozeki, T.

2005-01-01

We have investigated the profile formation and sustainment of tokamak plasmas with the current hole (CH) configuration by using 1.5D time-dependent transport simulations. A model of the current limit inside the CH on the basis of the Axisymmetric Tri-Magnetic-Islands equilibrium is introduced into the transport simulation. We found that a transport model with the sharp reduction of anomalous transport in the reversed-shear (RS) region can reproduce the time evolution of profiles observed in JT-60U experiments. The transport becomes neoclassical-level in the RS region, which results in the formation of profiles with internal transport barrier (ITB) and CH. The CH plasma has an autonomous property because of the strong interaction between a pressure profile and a current profile through the large bootstrap current fraction. The ITB width determined by the neoclassical-level transport agrees well with that measured in JT-60U. The energy confinement inside the ITB agrees with the scaling based on the JT-60U data. The scaling means the autonomous limitation of energy confinement in the CH plasma. The plasma with the large CH is sustained with the full current drive by the bootstrap current. The plasma with the small CH and the small bootstrap current fraction shrinks due to the penetration of inductive current. This shrink is prevented and the CH size can be controlled by the appropriate external current drive (CD). The CH plasma is found to respond autonomically to the external CD. (author)

Development of a linear motion antenna for the JT-60SA ECRF system

International Nuclear Information System (INIS)

Moriyama, Shinichi; Kobayashi, Takayuki; Isayama, Akihiko; Hoshino, Katsumichi; Suzuki, Sadaaki; Hiranai, Shinichi; Yokokura, Kenji; Sawahata, Masayuki; Terakado, Masayuki; Hinata, Jun; Wada, Kenji; Sato, Yoshikatsu

2013-01-01

Highlights: ► Development of an antenna featuring linear motion (LM) concept for long pulse electron cyclotron range of frequency (ECRF) heating and current drive in JT-60SA is in progress. ► A mock-up using a metallic sliding bearing with solid lubricant was fabricated. ► A vacuum pumping test with mass analyzer showed evidence of some hydrocarbons during shaft motion. ► Injection beam profile in toroidal beam scan was checked by low power measurement with mock-up. ► Current drive characteristics with the LM antenna for typical experimental scenarios of JT-60SA have been investigated by calculation. -- Abstract: Development of an antenna that features the linear motion (LM) concept for long-pulse electron-cyclotron range of frequency heating and current drive for the JT-60SA is in progress. Combining a linearly movable first mirror and a fixed curved second mirror allows the injection-beam angle to be controlled. Cooling water is fed through the drive shaft for the first mirror and through the fixed support for the second mirror. The shaft support structure uses a metallic sliding bearing with a solid lubricant. The sliding bearing supports combined linear and rotational motion, whereas a conventional ball bearing supports either linear or rotational motion. Therefore, the sliding bearing offers the advantage of reducing the support-structure volume, which is important in the design of the relatively narrow port duct of the JT-60SA. Recently, the sliding bearing has been installed into the mockup. Results of a vacuum test with a mass analyzer indicate the presence of hydrocarbons during shaft motion. The injection-beam profile obtained from a toroidal beam scan is checked against low-power measurements taken on the mockup. Finally, for typical JT-60SA experimental scenarios, heating- and current-drive characteristics of the LM antenna are investigated theoretically

Relationship between particle and heat transport in JT-60U plasmas with internal transport barrier

International Nuclear Information System (INIS)

Takenaga, H.

2002-01-01

Relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated for the first time in reversed shear (RS) and high-β p ELMy H-mode (weak positive shear) plasmas of JT-60U for understanding of compatibility of improved energy confinement and effective particle control such as exhaust of helium ash and reduction in impurity contamination. In the RS plasma, no helium and carbon accumulation inside the ITB is observed even with highly improved energy confinement. In the high-β p plasma, both helium and carbon density profiles are flat. As the ion temperature profile changes from parabolic- to box-type, the helium diffusivity decreases by a factor of about 2 as well as the ion thermal diffusivity in the RS plasma. The measured soft X-ray profile is more peaked than that calculated by assuming the same n AR profile as the n e profile in the Ar injected RS plasma with the box-type profile, suggesting accumulation of Ar inside the ITB. Particle transport is improved with no change of ion temperature in the RS plasma, when density fluctuation is drastically reduced by a pellet injection. (author)

Design study of the vertical field power supply for JT-60

International Nuclear Information System (INIS)

Yabuno, Kohei; Tani, Keiji; Shimada, Ryuichi; Kishimoto, Hiroshi; Yoshida, Hidetoshi

1977-09-01

The results of a basic design study of the vertical field power supply for JT-60 (JAERI large tokamak) are described. The objective of the study is to evaluate several types of power supply circuits for fast excitation and control of the vertical field. A design requirement is to produce a rapidly increasing vertical field within accuracy of +-5% around the proper field strength required to center the plasma in the vacuum vessel. The plasma current is assumed to increase at the rate of about 100 MA/sec. To meet the requirement, a maximum voltage of 15 kV is necessary in the current build-up time, while generally relatively low voltage is necessary after the current flattop is reached. A hybrid power supply which consists of a dc power source (a thyristor converter) and an inductive energy storage system is proposed. The maximum voltage of the dc power source is determined as 4 kV from the voltage required in the current flattop time. This is sufficient also in the current build-up time if the dc power source is used together with the inductive energy storage system. (auth.)

Triton burnup study using scintillating fiber detector on JT-60U

Energy Technology Data Exchange (ETDEWEB)

Harano, Hideki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1997-09-01

The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV {alpha} particles behave. The {alpha} particles` behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton`s behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector`s directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R{sub p}-scan and n{sub e}-scan experiments have been performed. The R{sub p}-scan experiment indicates that the triton`s transport was increased as the ripple amplitude over the triton became larger. In the n{sub e}-scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs.

Triton burnup study using scintillating fiber detector on JT-60U

International Nuclear Information System (INIS)

Harano, Hideki

1997-09-01

The DT fusion reactor cannot be realized without knowing how the fusion-produced 3.5 MeV α particles behave. The α particles' behavior can be simulated using the 1 MeV triton. To investigate the 1 MeV triton's behavior, a new type of directional 14 MeV neutron detector, scintillating fiber (Sci-Fi) detector has been developed and installed on JT-60U in the cooperation with LANL as part of a US-Japan collaboration. The most remarkable feature of the Sci-Fi detector is that the plastic scintillating fibers are employed for the neutron sensor head. The Sci-Fi detector measures and extracts the DT neutrons from the fusion radiation field in high time resolution (10 ms) and wide dynamic range (3 decades). Triton burnup analysis code TBURN has been made in order to analyze the time evolution of DT neutron emission rate obtained by the Sci-Fi detector. The TBURN calculations reproduced the measurements fairly well, and the validity of the calculation model that the slowing down of the 1 MeV triton was classical was confirmed. The Sci-Fi detector's directionality indicated the tendency that the DT neutron emission profile became more and more peaked with the time progress. In this study, in order to examine the effect of the toroidal field ripple on the triton burnup, R p -scan and n e -scan experiments have been performed. The R p -scan experiment indicates that the triton's transport was increased as the ripple amplitude over the triton became larger. In the n e -scan experiment, the DT neutron emission showed the characteristic changes after the gas puffing injection. It was theoretically confirmed that the gas puffing was effective for the collisionality scan. (J.P.N.) 127 refs

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

Development of virtual private network for JT-60SA CAD integration

International Nuclear Information System (INIS)

Oshima, Takayuki; Fujita, Takaaki; Seki, Masami; Kawashima, Hisato; Hoshino, Katsumichi; Shibanuma, Kiyoshi; Verrecchia, M.; Teuchner, B.

2010-01-01

The CAD models will be exchanged and integrated at Naka for JT-60SA, a common computer network efficiently connected between Naka site and the Garching site is needed to be established. Virtual Private Network (VPN) was introduced with 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. It was found that the introduction of the Wide Area File Service (WAFS) could solve the issue of the data transmission time and enhance the usability of the VPN for design integration in JT-60SA. (author)

Quality control of the software in the JT-60 computer control system

International Nuclear Information System (INIS)

Isaji, Nobuaki; Kurihara, Kenichi; Kimura, Toyoaki

1990-07-01

The JT-60 Control System should be improved corresponding to the experimental requirements. In order to keep the integrity of the system even in the modification the concept of quality control (QC) was introduced in the software development. What we have done for QC activity are (1) to establish standard procedures of the software development, (2) to develop support tools for grasping the present status of the program structure, and (3) to develop a document system, and a source program management system. This paper reports these QC activities and their problems for the JT-60 control system. (author)

Lower hybrid current drive in tokamak plasmas

Energy Technology Data Exchange (ETDEWEB)

Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1999-03-01

Past ten years progress on Lower Hybrid Current Drive (LHCD) experiments have demonstrated the largest non-inductive current (3.6 MA, JT-60U), the longest current sustainment (2 hours, TRIAM-1M), non-inductive current drive at the highest density (n-bar{sub e} - 10{sup 20}m{sup -3}, ALCATOR-C) and the highest current drive efficiency ({eta}{sub CD} = 3.5x10{sup 19} m{sup -2}A/W, JT-60). These results indicate that LHCD is one of the most promising methods to drive non-inductive current in the present tokamak plasmas. This paper presents recent experimental results on LHCD experiments. Basic theories of LH waves, the wave propagation and the current drive are briefly summarized. The main part of this paper describes several important results and their physical pictures on recent LHCD experiments; 1) the experimental set-up, 2) the current drive efficiency, 3) the control of current profile and MHD activities, 4) the global energy confinement, 5) the global power flow, 6) fast electron behavior, 7) interaction between LH waves and thermal/fast ions, 8) combination with other CD method. (author)

Effect of ripple-induced transport on H-mode performance in tokamaks

International Nuclear Information System (INIS)

Parail, V.; Vries, P. de; Lonnroth, J.; Kiviniemi, T.; Johnson, T.; Loarte, A.; Saibene, G.; Hatae, T.; Kamada, Y.; Konovalov, S.; Oyama, N.; Shinohara, K.; Tobita, K.; Urano, H.

2005-01-01

A number of experiments have shown that ripple-induced transport influences performance of ELMy H-modes in the tokamak. A noticeable difference in confinement, ELM frequency and amplitude was found between JET (with ripple amplitude δ∼0.1%) and JT-60U (with δ∼1%) in otherwise identical discharges. It was previously shown in JET experiments with enhanced ripple that a gradual increase in the ripple amplitude first leads to a modest improvement in plasma confinement, which is followed by the degradation of edge pedestal and further transition to the L-mode regime if δ increases further. The DIII-D team recently reported a marginal increase in confinement in experiments with an edge transport enhanced by the externally driven resonant magnetic perturbation. Numerical predictive modelling of the dynamics of ELMy H-mode JET plasma relevant to a JET/JT-60U similarity experiment has been conducted taking into account ripple-induced ion transport, which was computed using the orbit following code ASCOT. This predictive modelling reveals that, depending on plasma parameters, ripple amplitude and localisation (the latter depending on the toroidal coil design), this additional transport can either improve global plasma confinement or reduce it. These controlled ripple losses might be used as an effective tool for ELM mitigation and may provide an explanation for the difference between JET and JT-60U observed in the similarity experiments. A detailed comparison between ripple- induced transport and the alternative method of ELM mitigation by an externally driven edge magnetic perturbation is discussed. The fact that ripple losses mainly increase ion transport, while a stochastic magnetic layer increases electron transport indicates that it might be beneficial to use a combination of both methods in future experiments. This work was funded partly by the United Kingdom Engineering and Physical Sciences Research Council and by the European Communities under the contract of

Performance of RF power and phase control on JT-60 LHRF heating system

International Nuclear Information System (INIS)

Fujii, T.; Ikeda, Y.; Imai, T.; Honda, M.; Kiyono, K.; Maebara, S.; Saigusa, M.; Sakamoto, K.; Sawahata, M.; Seki, M.

1987-01-01

The performance of RF power and phase control on the JT-60 LHRFD heating system are presented. The JT-60 LHRF heating system has three units of huge RF source with a total output of 24 MW, each unit consisting of eight amplifier chains. A high power klystron generating 1 MW for 10 s at 2 GHz is used in each chain. Automatic gain control is employed to regulate the output power not only against gain fluctuations in the chain but also against the unstable plasma load without any output circulator for the klystron

Diagnostic planning in JT-60 project

International Nuclear Information System (INIS)

Matoba, Tohru; Suzuki, Yasuo; Funahashi, Akimasa; Itagaki, Tokiyoshi

1977-08-01

The diagnostic plans of JT-60 were made along with design of the main machine. Basic requirements of the diagnostic program are (1) multiple measurement of respective plasma parameters, (2) efficient usage of the discharge, (3) capable data acquisition system, (4) high reliability of the diagnostic equipments, and (5) systematic development of new diagnostic techniques. Dimensions of the diagnostic ports were determined in detailed design of the vacuum vessel, anticipating the possible diagnostic methods. The proposed diagnostic systems and the plans are shown in table and figures respectively. Problems in the diagnostics are also described. (auth.)

Technical aspects and manufacturing methods for JT-60SA toroidal field coil casings

International Nuclear Information System (INIS)

Rossi, Paolo; Cucchiaro, A.; Brolatti, G.; Cocilovo, V.; Ginoulhiac, G.; Polli, G.; Gabriele, M.; Di Muzio, F.; Philips, G.; Tomarchio, V.

2014-01-01

Highlights: • A contract between ENEA and Walter Tosto started on July 2012 for the construction of 18 TF coil casings for JT-60SA. • Design and manufacturing of mock-ups representative of straight and curved legs of the casings have been completed. • Final design of the casings has been completed and manufacturing activities have already started and are ongoing. • The completion of the first three casings will be completed within the end of 2013 and the production of all the 18 casings is foreseen by the end of 2015. - Abstract: JT-60SA is a superconducting tokamak machine to be assembled in Naka site, Japan, designed to contribute to the early realization of fusion energy by supporting the exploitation of ITER and research toward DEMO. In the frame of the Broader Approach Agreement a contract between ENEA and Walter Tosto (Chieti, Italy) started on July 2012 for the construction of 18 TF coil casings for JT-60SA. Two different sets of 9 casings each will be progressively delivered, from 2013 to the end of 2015, to ASG Superconductors (Genoa, Italy) and to Alstom (Belfort, France), where the integration of the winding pack into the casing will be carried out. Each TF coil casing (height 7.5 m and width 4.5 m) consists of four main components: one “Straight Leg Outboard” and one “Curved Leg Outboard” both with their own covers, “Straight Leg Inboard” and “Curved Leg Inboard”. The casing components are segmented in forgings and plates made of FM316LNL. The straight leg outboard is composed of two wings welded to a central core and two elbows welded at the ends with a cooling channel installed inside. Elbows of straight leg outboard are segmented in two half-elbows machined from 1 rough forging and welded to the central core made by plate. Welding of wings to the central core is performed in EBW (electron beam welding) and the straight part is welded to the elbows by NGTIG (TIG narrow gap) process. The curved leg outboard is composed of two

Technical aspects and manufacturing methods for JT-60SA toroidal field coil casings

Energy Technology Data Exchange (ETDEWEB)

Rossi, Paolo, E-mail: paolo.rossi@enea.it [ENEA, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Cucchiaro, A.; Brolatti, G.; Cocilovo, V.; Ginoulhiac, G.; Polli, G. [ENEA, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Gabriele, M.; Di Muzio, F. [Walter Tosto, Via Erasmo Piaggio, 66100 Chieti (Italy); Philips, G.; Tomarchio, V. [JT-60SA European Home Team, Boltzmannstrasse 2, D-85748 Garching (Germany)

2014-10-15

Highlights: • A contract between ENEA and Walter Tosto started on July 2012 for the construction of 18 TF coil casings for JT-60SA. • Design and manufacturing of mock-ups representative of straight and curved legs of the casings have been completed. • Final design of the casings has been completed and manufacturing activities have already started and are ongoing. • The completion of the first three casings will be completed within the end of 2013 and the production of all the 18 casings is foreseen by the end of 2015. - Abstract: JT-60SA is a superconducting tokamak machine to be assembled in Naka site, Japan, designed to contribute to the early realization of fusion energy by supporting the exploitation of ITER and research toward DEMO. In the frame of the Broader Approach Agreement a contract between ENEA and Walter Tosto (Chieti, Italy) started on July 2012 for the construction of 18 TF coil casings for JT-60SA. Two different sets of 9 casings each will be progressively delivered, from 2013 to the end of 2015, to ASG Superconductors (Genoa, Italy) and to Alstom (Belfort, France), where the integration of the winding pack into the casing will be carried out. Each TF coil casing (height 7.5 m and width 4.5 m) consists of four main components: one “Straight Leg Outboard” and one “Curved Leg Outboard” both with their own covers, “Straight Leg Inboard” and “Curved Leg Inboard”. The casing components are segmented in forgings and plates made of FM316LNL. The straight leg outboard is composed of two wings welded to a central core and two elbows welded at the ends with a cooling channel installed inside. Elbows of straight leg outboard are segmented in two half-elbows machined from 1 rough forging and welded to the central core made by plate. Welding of wings to the central core is performed in EBW (electron beam welding) and the straight part is welded to the elbows by NGTIG (TIG narrow gap) process. The curved leg outboard is composed of two

Progress of data processing system in JT-60 utilizing the UNIX-based workstations

International Nuclear Information System (INIS)

Sakata, Shinya; Kiyono, Kimihiro; Oshima, Takayuki; Sato, Minoru; Ozeki, Takahisa

2007-07-01

JT-60 data processing system (DPS) possesses three-level hierarchy. At the top level of hierarchy is JT-60 inter-shot processor (MSP-ISP), which is a mainframe computer, provides communication with the JT-60 supervisory control system and supervises the internal communication inside the DPS. The middle level of hierarchy has minicomputers and the bottom level of hierarchy has individual diagnostic subsystems, which consist of the CAMAC and VME modules. To meet the demand for advanced diagnostics, the DPS has been progressed in stages from a three-level hierarchy system, which was dependent on the processing power of the MSP-ISP, to a two-level hierarchy system, which is decentralized data processing system (New-DPS) by utilizing the UNIX-based workstations and network technology. This replacement had been accomplished, and the New-DPS has been started to operate in October 2005. In this report, we describe the development and improvement of the New-DPS, whose functions were decentralized from the MSP-ISP to the UNIX-based workstations. (author)

Driving mechanism of SOL plasma flow and effects on the divertor performance in JT-60U

International Nuclear Information System (INIS)

Asakura, N.

2002-01-01

SOL plasma flow plays an important role in the plasma transport along the field lines, and influences control of the divertor plasma and impurity ions. Recently, mechanisms producing the SOL flow such as drifts produced by electric field and pressure gradient are pointed out. In JT-60U, three reciprocating Mach probes were installed at the high-field-side (HFS) baffle, low-field-side (LFS) midplane and just below the X-point. The measurements of the SOL flow and plasma profiles both at the HFS and LFS, for the first time, found out the SOL flow pattern and its driving mechanism. 'Flow reversal' was found near the separatrix of the HFS and LFS. Radial profiles of the SOL flow were similar to those calculated numerically using the UEDGE code with the plasma drifts included. SOL particle fluxes towards the HFS and LFS divertors were, for the first time, evaluated. Important physics issues for the divertor design and operation, such as in-out asymmetries of the heat and particle fluxes, and control of impurity ions with intense gas puff and divertor pump (puff and pump), were investigated. (author)

Enhanced performance and control issues in JT-60U long pulse discharges

International Nuclear Information System (INIS)

Sakamoto, Y

2005-01-01

Recent experimental results are reported on control issues involved in long timescales and enhanced performance in JT-60U. The control issues in neoclassical tearing mode (NTM) suppression in the weak shear plasma regime include background optimization through decreasing β p (L q /L p ) at the rational surface and active stabilization of NTMs using ECCD. By optimizing β p (L q /L p ), a condition of β N ∼ 2.5 was sustained for 10 times the current profile relaxation time and one of β N ∼ 2.4 with q min ∼ 1.5 was sustained for 2.8 times the current profile relaxation time, with nearly full non-inductive current drive. In addition, a condition of β N ∼ 3 was sustained for 5.5 s through stabilization of NTMs using ECCD, and an EC driven current nearly equal to the bootstrap current was required for complete stabilization. In the reversed shear plasma regime, the issue is the existence of the steady state solution with a large f BS value. By controlling the pressure gradient at the internal transport barrier through toroidal rotation to avoid the disruption, a large f BS value of approximately 75% was sustained for 2.7 times the current profile relaxation time, with nearly full non-inductive current drive, and a steady-state solution with a large f BS value is confirmed. The control issues for the edge pedestal and edge localized modes (ELMs) are control of the pedestal pressure and the energy loss through ELMs. The pedestal pressure increases by >40% through the change in toroidal rotation. The type of ELM can be controlled by toroidal rotation from type-I to grassy

Development of TiC coated wall materials for JT-60

International Nuclear Information System (INIS)

Abe, T.; Murakami, Y.; Obara, K.; Hiroki, S.; Nakamura, K.; Inagawa, K.

1985-01-01

Development of titanium carbide (TiC, 20 μm thick) coated wall materials has been carried out for JT-60. Application of TiC coatings onto molybdenum and Inconel 625 substrates requires a deposition temperature below 950 0 C and 600 0 C respectively, because recrystallization of molybdenum and age hardening of Inconel 625 occur above these temperatures. Through this process of coating we develop a new type plasma CVD(TP-CVD method) for molybdenum and a new type PVD(HCD-ARE method) for Inconel 625 which could successfully reduce the deposition temperature to 900 0 C and 500 0 C, respectively. The TiC coated wall samples were characterized by AES, ESCA, X-ray diffractometer, EPMA, SEM, metalography, tensile tests, thermal shock tests, and other techniques. As a result of the above measurements, it was demonstrated that the characteristics of those TiC coated walls satisfy the requirements arising from JT-60 operation conditions. (orig.)

JT-60 plasma control system

International Nuclear Information System (INIS)

Kurihara, K.

1988-01-01

JT-60 plasma control can be performed by the supervisory controller, the measurement system and actuators such as the poloidal field coil power supplies, gas injectors, neutral beam injection (NBI) heating system and radio frequency (RF) heating system. One of the most important characteristics of this system is a perfect digital control one composed of mini-computers, fast array processors and CAMAC modules, and it has large flexibility and few troubles to adjust the system. This system started to be operated in April 1985, after the six-year-long design, construction and testing, and have been operated and improved many times for two years. In this paper, the final system specification and its performance are presented aiming at the technological aspect of hardware and software. In addition, and experienced troubles are also presented. (author)

Numerical and experimental analysis of eddy currents induced in tokamak machines

International Nuclear Information System (INIS)

Takahashi, T.; Takahashi, G.; Kazawa, Y.; Suzuki, Y.

1977-01-01

This paper deals with eddy current phenomena in Tokamak machines. A numerical method is presented which will permit eddy currents to be calculated. Examples of numerical results and a discussion of the JT-60 are shown. Calculations are checked by measurements in basic models

JT-60 negative ion beam NBI apparatus. Present state of its construction and initial experimental results

Energy Technology Data Exchange (ETDEWEB)

Kuriyama, Masaaki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1997-02-01

The NBI (Neutral Beam Injection) apparatus used for negative ion at first in the world, has an aim to actually prove heating and electric current drive with high density plasma at the JT-60 and to constitute physical and technical bases for selection and design of heating apparatus of ITER (International Thermal Nuclear Fusion Experimental Reactor). Construction of 500 KeV negative ion NBI apparatus for the JT-60 started to operate on 1993 was completed at March, 1996. On the way, at a preliminary test on forming and acceleration of the negative ion beam using a portion of this apparatus, 400 KeV and 13.5 A/D of the highest deuterium negative ion beam acceleration in the world was obtained successfully, which gave a bright forecasting of the plasma heating and electric current drive experiment using the negative ion NBI apparatus. After March, 1996, some plans to begin beam incident experiment at the JT-60 using the negative ion NBI apparatus and to execute the heating and electric current drive experiment at the JT-60 under intending increase of beam output are progressed. (G.K.)

Optical design for divertor Thomson scattering system for JT-60SA

International Nuclear Information System (INIS)

Kajita, Shin; Enokuchi, Akito; Hatae, Takaki; Itami, Kiyoshi; Hamano, Takashi; Kado, Shinichiro; Ohno, Noriyasu; Takeyama, Norihide

2014-01-01

Highlights: •A detailed designing for collection optical system of divertor Thomson scattering system in JT-60SA is conducted. •The assessment of the density and temperature errors of the measurement system is conducted. •It is shown that the measurement could be done with the temperature error of 50% when the density was 10 20 m −3 . •The availability of the laser transmission mirrors for the measurement system is discussed. •Several guidelines to improve the measurement system are discussed. -- Abstract: Optical design for divertor Thomson scattering system in JT-60SA has been conducted. The measurement system will use a Nd:YAG laser at 1064 nm, and scattered photons are collected by a collection optical system. The collection optics consists of primary mirror, secondary mirror, relay optics, and fiber collection optics. The laser transmission mirror and collection optics were designed to be installed in a slender lower port of JT-60SA. The assessment of the measurement errors in temperature was conducted for the designed collection optical system. Because of spatial limitation, the solid angle from the measurement points would be small especially for the measurement points in high field side, and consequently, the temperature errors in the high field side would be considerably large. The effects of several improvements on the error are discussed. Moreover, an assessment for the in-vessel laser transmission metallic mirrors is conducted for the present design

Vacuum pumping system for the JT-60 radio-frequency heating system

International Nuclear Information System (INIS)

Yokokura, Kenji; Ikeda, Yoshitaka; Imai, Tuyoshi; Suganuma, Kazuaki; Nagashima, Takashi

1988-01-01

The basic design requirements set up for the JT-60 radio-frequency heating system included: (1) rapid pumping of gas released upon application of a radio-frequency power to maintain the pressure in the launchers at 10 -2 - 10 -3 Pa or less, (2) incorporation of a gas analysis system that can operate under a strong field and high pressure (>10 -2 Pa) to permit remote controlled data collection and processing, and (3) low cost, multiple functions and high reliability. The vacuum pumping system, consisting of three units for low hybrid radio-frequency (LHRF) and one unit for ion cyclotron radio-frequency (ICRF), is connected to each launcher provided at the four ports of JT-60. The LHRF unit is composed of a main pump, an alumina joint for electrical insulation from the launcher, a metallic gate valve for isolation from the JT-60 vacuum region, and various vacuum gauges. Only a turbo-molecular pump is used for the ICRF system because a large-scale differential pumping is not required. A gas measuring system is incorporated which consists of a mass filter, personal computer, turbo-molecular pump, and variable flow valve equipped with an APG control. This system is designed to identify and make use of gas impurities released during the launcher aging process. The control system employed consists of a personal computer, interlock control board, data logger and other devices such as vacuum gages. (Nogami, K.)

Development of database for the divertor recycling in JT-60U and its analysis

Energy Technology Data Exchange (ETDEWEB)

Takizuka, Tomonori; Shimizu, Katsuhiro; Hayashi, Nobuhiko; Asakura, Nobuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Arakawa, Kazuya [Komatsu, Ltd., Tokyo (Japan)

2003-05-01

We have developed a database for the divertor recycling in JT-60U plasmas. This database makes it possible to investigate behaviors of the neutral-particle flux in plasmas and the ion flux to divertor plates under a condition for core-plasma parameters, such as electron density and heating power. The correlation between the electron density and the heating power is not strong in this database, and parameter scans for the density and the power in wide ranges are realized. On the basis of this database, we have analyzed the ion flux to divertor plates. The divertor-plate ion flux amplified by the recycling grows nonlinearly with the increase of the electron density n{sub e}. Its averaged dependence is a linear growth ({approx}n{sub e}{sup 1.0}) at the low density, and becomes a nonlinear growth ({approx}n{sub e}{sup 1.5}) at the high density. The spread of dependence from the averaged one is very large. This spread is caused mainly by complex physical characteristics of divertor plasmas, though it is little dependent on the heating power. The behavior of ion flux depends strongly on divertor configurations and divertor-plate/first-wall conditions. It is confirmed that the bifurcated transition takes place from the low-recycling divertor plasma at the low density to the high-recycling divertor plasma at the high density. The density at the transition is nearly proportional to the 1/4 power of the heating power. (author)

Stabilization of neoclassical tearing modes by electron cyclotron current drive in JT-60U

International Nuclear Information System (INIS)

Isayama, A.; Oyama, N.; Urano, H.; Suzuki, T.; Takechi, M.; Hayashi, N.; Nagasaki, K.; Kamada, Y.; Ide, S.; Ozeki, T.

2007-01-01

Results of active control of neoclassical tearing modes (NTMs) by electron cyclotron current drive (ECCD) in JT-60U are described. Growth of an NTM with poloidal mode number m = 3 and toroidal mode number n = 2 has been suppressed by ECCD inside the sawtooth inversion radius in the co-direction, showing the possibility of the coexistence of sawtooth oscillations and a small-amplitude m/n = 3/2 NTM without large confinement degradation. Stabilization of an m/n = 2/1 NTM by ECCD at the mode rational surface has been demonstrated with a small ratio of the current density driven by the electron cyclotron (EC) wave to the local bootstrap current density (∼ 0.5). In addition, dependence of the stabilization effect on ECCD location has been investigated in detail. It has been found that an m/n = 2/1 NTM can be completely stabilized with the misalignment of the ECCD location less than about half of the full island width, and that the m/n = 2/1 NTM is destabilized with the misalignment comparable to the full island width. Time-dependent, self-consistent simulation of magnetic island evolution using the TOPICS code has shown that the stabilization and destabilization of an m/n = 2/1 NTM are well reproduced with the same set of coefficients of the modified Rutherford equation. The TOPICS simulation has also clarified that EC wave power required for complete stabilization can be significantly reduced by narrowing the ECCD deposition width

Manufacturing of central control system of 'JT-60' a plasma feasibility experiment device

International Nuclear Information System (INIS)

Kondo, Ikuo; Kimura, Toyoaki; Murai, Katsuji; Iba, Daizo; Takemaru, Koichi.

1984-01-01

For constructing a critical-plasma-experiment apparatus JT-60, it was necessary to develop a new control system which enables to operate safely and smoothly a large scale nuclear fusion apparatus and to carry out efficient experiment. For the purpose, the total system control facility composed of such controllers as CAMAC system, timing system and protective interlock panel with multi-computer system as the core was developed. This system generalizes, keeps watch on and controls the total facilities as the key point of the control system of JT-60, and allows flexible operation control corresponding to the diversified experimental projects. At the same time, it carries out the fast real-time control of high temperature, high density plasma. In this paper, the system constitution, function and the main contents of development of the total system control facility are reported. JT-60 is constructed to attain the critical plasma condition as the premise of nuclear fusion reactors and to scientifically verify controlled nuclear fusion. Plasma expe riment will be started in April, 1985. The real-time control of plasma for carrying out high beta operation is planned, intending to develop future economical practical reactors. (Kako, I.)

Web-based Java application to advanced JT-60 Man-Machine Interfacing System for remote experiments

International Nuclear Information System (INIS)

Totsuka, Toshiyuki; Suzuki, Yoshio; Sakata, Shinya; Oshima, Takayuki; Iba, Katsuyuki

2008-01-01

Since remote participation in ITER experiments is planned, it is expected to demonstrate that the JT-60SA experiment is controlled from a Japanese remote experiment center located in Rokkasho-mura, Aomori-ken, Japan as a part of the ITER-BA project. Functions required for this experiment are monitoring of the discharge sequence status, handling of the discharge parameter, checking of experiment data, and monitoring of plant data, all of which are included in the existing JT-60 Man-Machine Interfacing System (MMIF). The MMIF is now only available to on-site users at the Naka site due to network safety. The motivation for remote MMIF is prompted by the issue of developing and achieving compatibility with network safety. The Java language has been chosen to implement this task. This paper deals with details of the JT-60 MMIF for the remote experiment that has evolved using the Java language

The influence of the analog-to-digital conversion error on the JT-60 plasma position/shape feedback control system

International Nuclear Information System (INIS)

Yoshida, Michiharu; Kurihara, Kenichi

1995-12-01

In the plasma feedback control system (PFCS) and the direct digital controller (DDC) for the poloidal field coil power supply in the JT-60 tokamak, it is necessary to observe signals of all the poloidal field coil currents. Each of the signals, originally measured by a single sensor, is distributed to the PFCS and DDC through different cable routes and different analog-to-digital converters from each other. This produces the conversion error to the amount of several bits. Consequently, proper voltage from feedback calculation cannot be applied to the coil, and hence the control performance is possibly supposed to deteriorate to a certain extent. This paper describes how this error makes an influence on the plasma horizontal position control and how to improve the deteriorated control performance. (author)

Effects of low central fuelling on density and ion temperature profiles in reversed shear plasmas on JT-60U

Energy Technology Data Exchange (ETDEWEB)

Takenaga, H; Ide, S; Sakamoto, Y; Fujita, T [Japan Atomic Energy Agency, Naka Ibaraki 311-0193 (Japan)], E-mail: takenaga.hidenobu@jaea.go.jp

2008-07-15

Effects of low central fuelling on density and ion temperature profiles have been investigated using negative ion based neutral beam injection and electron cyclotron heating (ECH) in reversed shear plasmas on JT-60U. Strong internal transport barrier (ITB) was maintained in density and ion temperature profiles, when central fuelling was decreased by switching positive ion based neutral beam injection to ECH after the strong ITB formation. Similar density and ion temperature ITBs were formed for the low and high central fuelling cases during the plasma current ramp-up phase. Strong correlation between the density gradient and the ion temperature gradient was observed, indicating that particle transport and ion thermal transport are strongly coupled or the density gradient assists the ion temperature ITB formation through suppression of drift wave instabilities such as ion temperature gradient mode. These results support that the density and ion temperature ITBs can be formed under reactor relevant conditions.

Effects of low central fuelling on density and ion temperature profiles in reversed shear plasmas on JT-60U

Science.gov (United States)

Takenaga, H.; Ide, S.; Sakamoto, Y.; Fujita, T.; JT-60 Team

2008-07-01

Effects of low central fuelling on density and ion temperature profiles have been investigated using negative ion based neutral beam injection and electron cyclotron heating (ECH) in reversed shear plasmas on JT-60U. Strong internal transport barrier (ITB) was maintained in density and ion temperature profiles, when central fuelling was decreased by switching positive ion based neutral beam injection to ECH after the strong ITB formation. Similar density and ion temperature ITBs were formed for the low and high central fuelling cases during the plasma current ramp-up phase. Strong correlation between the density gradient and the ion temperature gradient was observed, indicating that particle transport and ion thermal transport are strongly coupled or the density gradient assists the ion temperature ITB formation through suppression of drift wave instabilities such as ion temperature gradient mode. These results support that the density and ion temperature ITBs can be formed under reactor relevant conditions.

Review of JT-60 experiment (April-June, 1985)

International Nuclear Information System (INIS)

1986-11-01

Initial ohmic heating experiments in JT-60 were performed for a three month period of April-June 1985. A maximum plasma current of 1.6 MA was obtained for both divertor and limiter discharges. Low-q discharges of q eff = 2.5 and high density discharges of 4.8 x 10 19 m -3 were obtained in the divertor configuration. Typical divertor actions, i.e. particle exhaust, heat exhaust, impurity reduction and remote radiative cooling were demonstrated. (author)

ELM frequency dependence on toroidal rotation in the grassy ELM regime in JT-60U

International Nuclear Information System (INIS)

Oyama, N; Kamada, Y; Isayama, A; Urano, H; Koide, Y; Sakamoto, Y; Takechi, M; Asakura, N

2007-01-01

A systematic study of the effect of the level of toroidal plasma rotation at the top of the ion temperature pedestal ( T i ped ) on the edge localised mode (ELM) characteristics in JT-60U has been performed. The level of toroidal plasma rotation was varied by using different combinations of tangential and perpendicular neutral beam injection (NBI). In the grassy ELM regime at high triangularity (δ) and high safety factor (q), the ELM frequency clearly increased up to 1400 Hz, when counter (ctr) plasma rotation was increased. The response of the ELM frequency was independent of poloidal beta (β p ) in the range 0.84 p 0.53. Even in non-rotating plasma with balanced-NBIs, a high ELM frequency of ∼400 Hz was observed without a large energy loss. When the frequency of the plasma rotation in the co-direction of the plasma current became higher than ∼1 kHz, type I ELMs with a frequency of ∼20 Hz was observed. The achieved pedestal pressure and plasma confinement were similar both in plasmas with type I ELMs and in plasmas with grassy ELMs. The energy loss due to grassy ELMs was evaluated from the reduction in the electron temperature, and the ratio of the energy loss to the pedestal stored energy was less than 1%

Results of tests of the X2274 high power tetrode in a JT-60 110 to 130 MHz ICRH amplifier

International Nuclear Information System (INIS)

Remsen, D.B.; Loring, C.M.; McNees, S.G.; Moriyama, S.; Ogawa, Y.; Anno, K.; Fujii, T.; Terakado, M.; Kogure, S.; Nagashima, T.; Ohta, M.

1990-09-01

This paper reports the results of tests of the newly developed Varian EIMAC X2274 in the JAERI JT-60 ICRH system at pulse lengths up to 6 seconds at 131 MHz. It is our belief that these tests achieved the highest long pulse, or CW, power that has ever been delivered by a single power grid tube at frequencies above 100 MHz. Varian's EIMAC X2274, developed in conjunction with General Atomics and the US Department of Energy, uses an improved pyrolytic graphite grid configuration which provides significantly better vhf performance than the grids of the X2242 tetrode which was tested in this system in 1989. The EIMAC X2274 combines the improved grids with a new anode design which reduces the required water flow approximately 50% and increases the maximum anode dissipation 80%. All tests were performed at 131 MHz, the system's highest operating frequency. Tests of both prototype EIMAC X2274s produced essentially identical results. The basic objectives of these tests were: to demonstrate that the system with the EIMAC X2274 can reliably produce 1.5 MW at 130 MHz at 5 second pulse lengths for the JT-60U tokamak and to collect data for use in the design of future high power ICRH systems. In these tests the tube and system produced up to 1.7 MW at pulse lengths up to 5.4 seconds: i.e, the EIMAC X2274 in this system can easily meet Objective 1. The remainder of this paper shows that Objective 2 has been fulfilled. In addition to the high power tests, operational range tests were performed under variable VSWR conditions. Unlike the EIMAC X2242 tests were rf current heating of the screen grid limited output power, system parameters, rather than tube parameters, limited the output power in the high power tests. Operational range tests were conducted at output power levels chosen to be well within the system's anode cooling capability

Beta-limit of a large tokamak with a circular cross-section

International Nuclear Information System (INIS)

Tsunematsu, Toshihide; Takeda, Tatsuoki; Kurita, Gen-ichi; Azumi, Masafumi; Matsuura, Toshihiko; Gruber, R.; Troyon, F.

1982-01-01

The dependence of stabilizing effect of a conducting shell on a poloidal beta value (βsub(p)) is investigated as to instabilities with low toroidal mode numbers (n = 1 and 2) for a tokamak with a circular cross-section such as JT-60. The n = 1 mode is completely stabilized by the conducting shell which is located at a practically possible position and the critical position of the shell becomes closer to the plasma surface with increasing βsub(p). The stabilizing effect on the n = 2 mode is remarkable for higher βsub(p) when the shell is placed sufficiently close to the plasma surface but the shell far from the plasma surface has hardly an effect on the stability property of a higher βsub(p) plasma. It is concluded that critical β of about 2% is attainable even in a standard circular tokamak such as JT-60 and higher β value is also expected by taking advantage of the closely located conducting shell. (author)

Development and contribution of rf heating and current drive systems to long pulse, high performance experiments in JT-60U

International Nuclear Information System (INIS)

Moriyama, Shinichi; Seki, Masami; Terakado, Masayuki; Shimono, Mitsugu; Ide, Shunsuke; Isayama, Akihiko; Suzuki, Takahiro; Fujii, Tsuneyuki

2005-01-01

To contribute to high performance long pulse (∼65 s) experiments in JT-60U, the target of the electron cyclotron (EC) operation in long pulse is 0.6 MW for 30 s with four gyrotrons, though 10 MJ (2.8 MW and 3.6 s) was achieved in high power operation before 2003. One of the critical issues for the long pulse operation is detuning due to decay in beam current of the gyrotron. This decay comes from the cathode cooling by continuous electron emission. As a countermeasure for this issue, active adjustments for the heater current and anode voltage during the pulse have successfully extended the duration of a good oscillation condition for the gyrotron. As a result, 0.4 MW for 16 s with one gyrotron to the dummy load and for 8.7 s to the plasma have been achieved up to now

Development of a new discharge control system utilizing UNIX workstations and VME-bus systems for JT-60

Energy Technology Data Exchange (ETDEWEB)

Akasaka, Hiromi; Sueoka, Michiharu; Takano, Shoji; Totsuka, Toshiyuki; Yonekawa, Izuru; Kurihara, Kenichi; Kimura, Toyoaki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

2002-01-01

The JT-60 discharge control system, which had used HIDIC-80E 16 bit mini-computers and CAMAC systems since the start of JT-60 experiment in 1985, was renewed in March, 2001. The new system consists of a UNIX workstation and a VME-bus system, and features a distributed control system. The workstation performs message communication with a VME-bus system and controllers of JT-60 sub-systems and processing for discharge control because of its flexibility to construction of a new network and modifications of software. The VME-bus system performs discharge sequence control because it is suitable for fast real time control and flexible to the hardware extension. The replacement has improved the control function and reliability of the discharge control system and also has provided sufficient performance necessary for future modifications of JT-60. The new system has been running successfully since April 2001. The data acquisition speed was confirmed to be twice faster than the previous one. This report describes major functions of the discharge control system, technical ideas for developing the system and results of the initial operation in detail. (author)

Balance of the stored energies sustained by the internal and edge transport barriers and effects of ELMs and L-H transitions in JT-60U

International Nuclear Information System (INIS)

Kamada, Y.; Yoshida, M.; Sakamoto, Y.; Koide, Y.; Oyama, N.; Urano, H.; Kamiya, K.; Suzuki, T.; Isayama, A.

2009-01-01

To understand key physics processes determining radial profiles of the kinetic plasma parameters in the advanced tokamak operation scenarios, correlations between the edge transport barrier (ETB) and the internal transport barrier (ITB) have been studied in the JT-60U tokamak device. It has been found that the edge pedestal poloidal beta, β p -ped, increases almost linearly with the total poloidal beta, β p -tot, over a wide range of the plasma current for type I ELMing H-mode plasmas, and this dependence becomes stronger with increasing triangularity. This dependence is not due to the profile stiffness, since the dependence is the same regardless of the existence of ITB. As the stored energy inside the ITB-foot radius (W ITB ) increases, the total thermal stored energy (W th ) increases and then the pedestal stored energy (W ped ) increases. On the other hand, as W ped increases, the ELM penetration expands more inwards and finally reaches the ITB-foot radius. At this situation, the ITB-foot radius cannot move outwards because of the erosion by ELMs. Then the fractions of W ITB /W th and W ped /W th become almost constant. It has also been found that the type I ELM expels/decreases the edge toroidal momentum larger than the edge ion thermal energy. The ELM penetration for the toroidal rotation tends to be deeper than that for the ion temperature and can exceed the ITB-foot radius. The ELM penetration is deeper for CO-rotating plasmas than CTR rotating plasmas. In both cases, the ELM penetration is deeper in the order of the toroidal rotation (V t ), the ion temperature (T i ) and then the electron temperature (T e ). The L-H transition also changes the V t profile more significantly than the T i profile. At the L-H transition, the pedestal V t shifts into the CTR-direction deeply and suddenly without a change in T i , and then the pedestal V t grows further together with a growth of the pedestal T i in a slower timescale. Such changes in V t by ELMs and L

Flow shear stabilization of hybrid electron-ion drift mode in tokamaks

International Nuclear Information System (INIS)

Bai, L.

1999-01-01

In this paper, a model of sheared flow stabilization on hybrid electron-ion drift mode is proposed. At first, in the presence of dissipative trapped electrons, there exists an intrinsic oscillation mode in tokamak plasmas, namely hybrid dissipative trapped electron-ion temperature gradient mode (hereafter, called as hybrid electron-ion drift mode). This conclusion is in agreement with the observations in the simulated tokamak experiment on the CLM. Then, it is found that the coupling between the sheared flows and dissipative trapped electrons is proposed as the stabilization mechanism of both toroidal sheared flow and poloidal sheared flow on the hybrid electron-ion drift mode, that is, similar to the stabilizing effect of poloidal sheared flow on edge plasmas in tokamaks, in the presence of both dissipative trapped electrons and toroidal sheared flow, large toroidal sheared flow is always a strong stabilizing effect on the hybrid electron-ion drift mode in internal transport barrier location, too. This result is consistent with the experimental observations in JT-60U. (author)

Flow shear stabilization of hybrid electron-ion drift mode in tokamaks

International Nuclear Information System (INIS)

Bai, L.

2001-01-01

In this paper, a model of sheared flow stabilization on hybrid electron-ion drift mode is proposed. At first, in the presence of dissipative trapped electrons, there exists an intrinsic oscillation mode in tokamak plasmas, namely hybrid dissipative trapped electron-ion temperature gradient mode (hereafter, called as hybrid electron-ion drift mode). This conclusion is in agreement with the observations in the simulated tokamak experiment on the CLM. Then, it is found that the coupling between the sheared flows and dissipative trapped electrons is proposed as the stabilization mechanism of both toroidal sheared flow and poloidal sheared flow on the hybrid electron-ion drift mode, that is, similar to the stabilizing effect of poloidal sheared flow on edge plasmas in tokamaks, in the presence of both dissipative trapped electrons and toroidal sheared flow, large toroidal sheared flow is always a strong stabilizing effect on the hybrid electron-ion drift mode in internal transport barrier location, too. This result is consistent with the experimental observations in JT-60U. (author)

Current drive and sustain experiments with the bootstrap current in JT-60

International Nuclear Information System (INIS)

Kikuchi, Mitsuru; Azumi, Masafumi; Tani, Keiji; Tsuji, Shunji; Kubo, Hirotaka

1989-11-01

The current drive and sustain experiments with the neoclassical bootstrap current are performed in the JT-60 tokamak. It is shown that up to 80% of total plasma current is driven by the bootstrap current in extremely high β p regime (β p = 3.2) and the current drive product I p (bootstrap) n-bar e R p up to 4.4 x 10 19 MAm -2 has been attained with the bootstrap current. The experimental resistive loop voltages are compared with the calculations using the neoclassical resistivity with and without the bootstrap current and the Spitzer resistivity for a wide range of the plasma current (I p = 0.5 -2 MA) and the poloidal beta (β p = 0.1 - 3.2). The calculated resistive loop voltage is consistent with the neoclassical prediction including the bootstrap current. Current sustain with the bootstrap current is tested by terminating the I p feedback control during the high power neutral beam heating. An enhancement of the L/R decay time than those expected from the plasma resistivity with measured T e and Zeff has been confirmed experimentally supporting the large non-inductive current in the plasma and is consistent with the neoclassical prediction. A new technique to calculate the bootstrap current in multi-collisionality regime for finite aspect ratio tokamak has bee developed. The neoclassical bootstrap current is calculated directly through the force balance equations between viscous and friction forces according to the Hirshman-Sigmar theory. The bootstrap current driven by the fast ion component is also included. Ballooning stability of the high β p plasma are analyzed using the current profiles including the bootstrap current. The plasma pressure is close to the ballooning limit in high β p discharges. (author)

Performance of the JT-60SA cryogenic system under pulsed heat loads during acceptance tests

Science.gov (United States)

Hoa, C.; Bonne, F.; Roussel, P.; Lamaison, V.; Girard, S.; Fejoz, P.; Goncalves, R.; Vallet, J. C.; Legrand, J.; Fabre, Y.; Pudys, V.; Wanner, M.; Cardella, A.; Di Pietro, E.; Kamiya, K.; Natsume, K.; Ohtsu, K.; Oishi, M.; Honda, A.; Kashiwa, Y.; Kizu, K.

2017-12-01

The JT-60SA cryogenic system a superconducting tokamak currently under assembly at Naka, Japan. After one year of commissioning, the acceptance tests were successfully completed in October 2016 in close collaboration with Air Liquide Advanced Technologies (ALaT), the French atomic and alternative energies commission (CEA), Fusion for Energy (F4E) and the Quantum Radiological Science and Technology (QST). The cryogenic system has several cryogenic users at various temperatures: the superconducting magnets at 4.4 K, the current leads at 50 K, the thermal shields at 80 K and the divertor cryo-pumps at 3.7 K. The cryogenic system has an equivalent refrigeration power of about 9.5 kW at 4.5 K, with peak loads caused by the nuclear heating, the eddy currents in the structures and the AC losses in the magnets during cyclic plasma operation. The main results of the acceptance tests will be reported, with emphasis on the management of the challenging pulsed load operation using a liquid helium volume of 7 m3 as a thermal damper.

Communication systems in JT-60 control

International Nuclear Information System (INIS)

Kimura, T.; Hosogane, N.; Kondo, I.; Kumahara, T.; Kurihara, K.; Yonekawa, I.; Yoshino, R.

1983-01-01

A new concept in communication is applied to the JT-60 control system which handles a large amount of data for the plant support and monitoring and for the discharge control including plasma feedback control. The communication systems are characterized by 1) adoption of an efficient protocol in the central highways which are composed of dual serial CAMAC ones, 2) standardization of the protocol and data format between the central controller and each subsystem one, 3) adoption of a polling method for plant monitoring and of block transfer for discharge conditions and results, and 4) use of novel modules for the fast data transfer in the real-time systems. A compact tool has also been developed for testing the data communication

Development and upgrade of new real time processor in JT-60 data processing system

International Nuclear Information System (INIS)

Sakata, Shinya; Koiwa, Motonao; Matsuda, Toshiaki; Aoyagi, Tetsuo

2000-07-01

At the beginning of JT-60 experiments, the real time processor (RTP) in the data processing system was mainly constructed by PANAFACOM U-1500. As the computer became superannuated, however, it gradually became difficult to maintain both hardware and software. A performance of a recent UNIX workstation has been remarkably progressed. The UNIX workstation has a large flexibility for user application programs, an easiness for maintenance of the hardware and an ability of expansion to peripheral devices. Therefore, the RTP system is newly reconstructed by using the UNIX workstation. This report describes the overview, the basic design and the recent upgrade on the RTP in the data processing system. (author)

Models for Predicting Boundary Conditions in L-Mode Tokamak Plasma

International Nuclear Information System (INIS)

Siriwitpreecha, A.; Onjun, T.; Suwanna, S.; Poolyarat, N.; Picha, R.

2009-07-01

Full text: The models for predicting temperature and density of ions and electrons at boundary conditions in L-mode tokamak plasma are developed using an empirical approach and optimized against the experimental data obtained from the latest public version of the International Pedestal Database (version 3.2). It is assumed that the temperature and density at boundary of L-mode plasma are functions of engineering parameters such as plasma current, toroidal magnetic field, total heating power, line averaged density, hydrogenic particle mass (A H ), major radius, minor radius, and elongation at the separatrix. Multiple regression analysis is carried out for these parameters with 86 data points in L-mode from Aug (61) and JT60U (25). The RMSE of temperature and density at boundary of L-mode plasma are found to be 24.41% and 18.81%, respectively. These boundary models are implemented in BALDUR code, which will be used to simulate the L-mode plasma in the tokamak

A new workstation based man/machine interface system for the JT-60 Upgrade

International Nuclear Information System (INIS)

Yonekawa, I.; Shimono, M.; Totsuka, T.; Yamagishi, K.

1992-01-01

Development of a new man/machine interface system was stimulated by the requirements of making the JT-60 operator interface more 'friendly' on the basis of the past five-year operational experience. Eleven Sun/3 workstations and their supervisory mini-computer HIDIC V90/45 are connected through the standard network; Ethernet. The network is also connected to the existing 'ZENKEI' mini-computer system through the shared memory on the HIDIC V90/45 mini-computer. Improved software, such as automatic setting of the discharge conditions, consistency check among the related parameters and easy operation for discharge result data display, offered the 'user-friendly' environments. This new man/machine interface system leads to the efficient operation of the JT-60. (author)

Review of JT-60U experimental results from January to October, 1992

International Nuclear Information System (INIS)

1993-03-01

Heating experiments have been carried out in JT-60U with plasma current up to 4 MA, toroidal field up to 4.2 T and neutral beam heating power of 30 MW. H-mode was observed at high toroidal field of 4.2 T. Confinement improvement factor of up to 1.6 over the L-mode scaling has been obtained during continuous ELMy phase at 4.2 T, 2.7 MA and an injection power of 25 MW. Confinement enhancement of 2.2 has been obtained. At plasma current of 3.5 MA, diamagnetic stored energy of 7.7 MJ and energy confinement time of 0.36 s have been obtained. In the discharges with high poloidal beta exhibit, high ion temperatures of 38 keV, electron temperatures of 12 keV, maximum diamagnetic stored energy of 6.1 MJ and neutron yield of 2.8 x 10 16 s -1 were obtained. Schemes to avoid locked modes and disruptions have been checked experimentally. Divertor measurement indicates that high q, high density operation is favorable for divertor heat handling. Quantitative analysis of divertor spectroscopy data indicates that the carbon production rate at the divertor plate is explained by deuterium, oxygen and carbon sputtering. Toroidal ripple loss power of NB-injected fast ions to the first wall was investigated by using an infrared TV camera. Absolute values of the ripple loss power and dependence on plasma parameters were consistent with values calculated by an orbit following Monte-Carlo code. In the LHCD area, the linear dispersion relation and accessibility condition for LH wave were experimentally validated. Furthermore, the power directly lost via high energy electron was shown to scale as the slowing down time. Current drive of 400 kA was realized with tangential NB. ICRF experiments using two new antennas started. Sawtooth stabilization at a high n-bar e /P tot value was achieved. (J.P.N.)

Development of a VME and CAMAC based data acquisition and transfer system for JT-60 control

International Nuclear Information System (INIS)

Totsuka, Toshiyuki

1993-08-01

Development of a VME and CAMAC based data acquisition and transfer system for JT-60 Control is reported. The present data acquisition and transfer system in JT-60 control is basically composed of CAMAC devices. Since the system equipped with 16-bit microcomputers was manufactured more than ten years ago, the performance and program development environment of the system are apparently worse than those of modern 32-bit microcomputers. To improve these disadvantages, a new data acquisition and transfer system using VME-based 32-bit microcomputers and CAMAC drivers is under design. Corresponding to this design, a CAMAC handler, which runs on the microcomputer, for the VME based CAMAC driver was newly developed. Moreover, the functions of the driver and data transfer performance of the VME and CAMAC complex system were tested. The test results shown that the VME based microcomputer and CAMAC serial driver can be applied for the fast and reliable acquisition and transfer system for JT-60 control. (author)

Predictions of toroidal rotation and torque sources arising in non-axisymmetric perturbed magnetic fields in tokamaks

Science.gov (United States)

Honda, M.; Satake, S.; Suzuki, Y.; Shinohara, K.; Yoshida, M.; Narita, E.; Nakata, M.; Aiba, N.; Shiraishi, J.; Hayashi, N.; Matsunaga, G.; Matsuyama, A.; Ide, S.

2017-11-01

Capabilities of the integrated framework consisting of TOPICS, OFMC, VMEC and FORTEC-3D, have been extended to calculate toroidal rotation in fully non-axisymmetric perturbed magnetic fields for demonstrating operation scenarios in actual tokamak geometry and conditions. The toroidally localized perturbed fields due to the test blanket modules and the tangential neutral beam ports in ITER augment the neoclassical toroidal viscosity (NTV) substantially, while do not significantly influence losses of beam ions and alpha particles in an ITER L-mode discharge. The NTV takes up a large portion of total torque in ITER and fairly decelerates toroidal rotation, but the change in toroidal rotation may have limited effectiveness against turbulent heat transport. The error field correction coils installed in JT-60SA can externally apply the perturbed fields, which may alter the NTV and the resultant toroidal rotation profiles. However, the non-resonant n=18 components of the magnetic fields arising from the toroidal field ripple mainly contribute to the NTV, regardless of the presence of the applied field by the coil current of 10 kA , where n is the toroidal mode number. The theoretical model of the intrinsic torque due to the fluctuation-induced residual stress is calibrated by the JT-60U data. For five JT-60U discharges, the sign of the calibration factor conformed to the gyrokinetic linear stability analysis and a range of the amplitude thereof was revealed. This semi-empirical approach opens up access to an attempt on predicting toroidal rotation in H-mode plasmas.

Tokamak experiments

International Nuclear Information System (INIS)

Robinson, D.C.

1987-01-01

With the advent of the new large tokamaks JET, JT-60 and TFTR important advances in magnetic confinement have been made. These include the exploitation of radio frequency and neutral beam heating on a much larger scale than previously, the demonstration of regimes of improved confinement and the demonstration of current drive at the Megamp level. A number of small and medium sized tokamaks have also come into operation recently such as WT-3 in Japan with an emphasis on radio frequency current drive and HL-1 a medium sized tokamak in China. Each of these new tokamaks is addressing specific problems which remain for the future development of the system. Of these particular problems: β, density and q limits remain important issues for the future development of the tokamak. β limits are being addressed on the DIII-D device in the USA. The anomalous confinement that the tokamak displays is being explored in detail on the TEXT device in the USA. Two other problems are impurity control and current drive. There is significant emphasis on divertor configurations at the present time with their enhanced confinement in the so called H mode. Due to improved discharge cleaning techniques and the ability to repetitively refuel using pellets, purer plasmas can be obtained even without divertors. Current drive remains a crucial issue for quasi of near steady state operation of the tokamak in the future and many current drive schemes are being investigated. (author) [pt

VME multiprocessor system for plasma control at the JT-60 Upgrade

International Nuclear Information System (INIS)

Kimura, T.; Kurihara, K.; Takahashi, M.; Kawamata, Y.; Akasaka, H.; Matsukawa, M.

1989-01-01

In this paper design and preliminary tests are reported of a VME multiprocessor system for the JT-60 Upgrade plasma control utilizing three MC88100 based RISC computers and VME buses. The design of the VME system was stimulated by faster and more accurate computation requirements for the plasma position and shape control

Global energy confinement in JT-60 neutral beam heated L-mode discharges

International Nuclear Information System (INIS)

Naito, O.; Hosogane, N.; Tsuji, S.; Ushigusa, K.; Yoshida, H.

1990-01-01

The global energy confinement characteristics of neutral beam heated JT-60 discharges are presented. There is a difference in the dependence of the energy confinement time on the plasma current between limiter and divertor discharges. For limiter discharges, the energy confinement increases with plasma current up to 3.2 MA, whereas for divertor discharges this improvement saturates when the safety factor drops below 3, independent of the location of the X-point. The JT-60 L-mode results indicate that the deterioration in energy confinement for q < 3, which is also found in H-mode regimes of other devices, may be a universal characteristic of divertor discharges. Regarding the scaling with plasma size, it is shown that the global/incremental confinement time increases with plasma minor radius. For sufficiently large plasmas, however, the global/incremental confinement time is no longer a function of minor radius. (author). 13 refs, 14 figs

The design study of the JT-60SU device. No. 4. The vacuum vessel and cryostat of JT-60SU

International Nuclear Information System (INIS)

Neyatani, Yuzuru; Ushigusa, Kenkichi; Tobita, Kenji

1997-03-01

The vacuum vessel and the cryostat for the JT-60 Super Upgrade (JT-60SU) have been designed. Two types of the complex materials for the vacuum vessel were chosen on the basis of the avoidance of tritium occlusion and the low irradiation, i.e. (1) SUS316 covered by tungsten plate (30mm thickness) as a γ-ray shielding, (2) Ti-6Al-4V alloy covered by SUS430 plate (1mm thickness) as a tritium protector. Selecting the double skin type of vacuum vessel with toroidally continued structure gave the basic design of the vacuum vessel satisfying the design criteria of the vessel strength for the electromagnetic force, heat load and the property of radiation shielding. The characteristics of the SUS316 covered by tungsten plate type is that as the tungsten can shield the γ-ray, the dose rate inside the vacuum vessel during the maintenance can reduce effectively. The advantage of the Ti-6Al-4V alloy covered by SUS430 plate type vacuum vessel is the quick reduction of the radioactive isotope because of no production of the isotopes with long half-life periods. Channel type and vertical type of the divertor were designed. The sector type of toroidally separated structure was selected for the remote handling. The material of the armor plate was not determined because no material endure the high heat load on the divertor. The cryostat composing the dome and the tank was designed. The electromagnetic force by the eddy current, generated at the plasma start up phase and at the quench of CS super-conducting coil, were small compared to the force produced by the stress limit. (author)

The healthiness of JT-60 ICRF antenna and development of its temperature measurement device

International Nuclear Information System (INIS)

Hiranai, Shinichi; Yokokura, Kenji; Moriyama, Shinichi; Sato, Tomio; Ishii, Kazuhiro; Fujii, Tsuneyuki

1998-03-01

Ion Cyclotron Range of Frequency (ICRF) heating system in JT-60 employs two antennas to couple RF power in the range of 100 MHz to the plasma. The antennas are installed in the vacuum vessel of JT-60, facing to the high temperature plasma. Due to the severe heat load from the plasma, parts of the antenna surface are suffering from melt. It is important to investigate the mechanism of the heat load and the melting. 'Temperature measurement for ICRF antenna surface' employing an infrared thermographic camera has been developed, in order to investigate the heat load to the antenna and to maintain the antenna available. We have succeeded in minimizing the melting damage of the antenna surface using the temperature measurement device. (author)

Results of the H-mode experiments with JT-60 outer and lower divertors

International Nuclear Information System (INIS)

Nakamura, Hiroo; Tsuji, Shunji; Nagami, Masayuki

1989-08-01

In JT-60, hydrogen H-mode experiments with outer and lower divertors were performed. In the outer divertor, H-mode were obtained, similar to the ones observed in the other lower/upper divertors. Its threshold absorbed power and electron density were 16 MW and 1.8 x 10 19 m -3 . In the two combined heatings with NB+ICRF and NB+LHRF, H-mode discharges are also obtained. Moreover, in new configuration of lower divertor, H-mode phases without and with ELM are obtained. Typical results of the lower divertor are shown to compare the H-mode characteristics between the two configurations. Improvement of the energy confinement time in the two divertors was limited to 10 %. Analyses on ballooning/interchange instabilities were carried out with precise equlibria of JT-60. These results showed that the both modes were enough stable. (author)

Helium exhaust and forced flow effects with both-leg pumping in W-shaped divertor of JT-60U

International Nuclear Information System (INIS)

Sakasai, A.; Takenaga, H.; Higashijima, S.; Kubo, H.; Nakano, T.; Tamai, H.; Sakurai, S.; Akino, N.; Fujita, T.; Asakura, N.; Itami, K.; Shimizu, K.

2001-01-01

The W-shaped divertor of JT-60U was modified from inner-leg pumping to both-leg pumping. After the modification, the pumping rate was improved from 3% with inner-leg pumping to 5% with both-leg pumping in a divertor-closure configuration, which means both separatrixes close to the divertor slots. Efficient helium exhaust was realized in the divertor-closure configuration with both-leg pumping. A global particle confinement time of τ* He =0.4s and τ* He /τ E =3 was achieved in attached ELMy H-mode plasmas. The helium exhaust efficiency with both-leg pumping was extended by 45% as compared with inner-leg pumping. By using central helium fueling with He-beam injection, the helium removal from the core plasma inside the internal transport barrier (ITB) in reversed shear plasmas in the divertor-closure configuration was investigated for the first time. The helium density profiles inside the ITB were peaked as compared with those in ELMy H-mode plasmas. In the case of low recycling divertor, it was difficult to achieve good helium exhaust capability in reversed shear plasmas with ITB. However, the helium exhaust efficiency was improved with high recycling divertor. Carbon impurity reduction was observed by the forced flow with gas puff and effective divertor pumping. (author)

Design of a new P-NBI control system for 100-s injection in JT-60SA

International Nuclear Information System (INIS)

Okano, F.; Shinozaki, S.; Honda, A.; Ooshima, K.; Numazawa, S.; Ikeda, Y.

2008-01-01

Modification of JT-60U to a superconducting device (so-called JT-60SA) has been planned to contribute to ITER and DEMO. The positive-ion-based NBI system (P-NBI) is required to inject 24 MW for 100 s with 12 units. The P-NBI control system is to be fully remodeled with PLC (Programmable Logic Controller), which is featured by high market availability, system extensibility, cost-effectiveness, and independent development in programming. One of the critical issues to apply the PLC to the P-NBI control system is to control quickly the high voltage power supplies within 200 μs. For this purpose, the fastest PLC dealing with 4 refresh words at the processing time of 200 μs is to be employed. The second issue is to construct a data acquisition system for such a large number of data channels (∼2300 digital and ∼1300 analog data channels). The use of PLC linked with PC-based data measurement devices via Ethernet allows processing the large number of channels. The third issue is to make the man-machine interface simple. The marketed software giving an easy product of graphic menus is available for PLC programming. From these results, it is expected that commercial PLC could be applied to the large-scale control system of the P-NBI system for 100 s operations

Tokamaks - Third Edition

International Nuclear Information System (INIS)

Rogister, A L

2004-01-01

an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11-12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of

Dynamic simulations of the cryogenic system of a tokamak

International Nuclear Information System (INIS)

Cirillo, R.; Hoa, C.; Michel, F.; Rousset, B.; Poncet, J.M.

2015-01-01

In a tokamak plasma confinement is achieved through high magnetic fields generated by superconductive coils that need to be cooled down to 4.4 K with a forced flow of supercritical Helium. Tokamak's coil system works cyclically and so it is subject to pulsed heat loads which have to be handled by the refrigerator. This latter has to be sized on the average power value and not according to the peak to limit investment and operation costs and hence the heat load needs to be smoothed. CEA Grenoble is in charge of providing the cryogenic system for the Japanese tokamak JT60-SA, currently under construction in Naka (Japan). Hence, in order to model and study the smoothing strategies, an experimental set up: HELIOS (Helium Loop for high load smoothing) has been built. This is a scaled down model (1:20) of the helium distribution system whose main components are a saturated helium bath and a supercritical helium loop. This large installation can reproduce conditions of pressure, temperature and transport times, similar to those expected in the cooling circuits of the central solenoid superconducting magnets of JT-60SA. The peak loads representative of the tokamak operation have been reproduced and smoothed before they arrive in the refrigerator, by means of a saturated helium bath (thermal reservoir). A dynamic modelling of the cryogenic system is presented, with results on the pulsed load scenarios. All the simulations have been performed with EcosimPro software developed and the cryogenic library: CRYOLIB. This document is made up of an abstract and the slides of the presentation

Plasma diagnostics on large tokamaks

International Nuclear Information System (INIS)

Orlinskij, D.V.; Magyar, G.

1988-01-01

The main tasks of the large tokamaks which are under construction (T-15 and Tore Supra) and of those which have already been built (TFTR, JET, JT-60 and DIII-D) together with their design features which are relevant to plasma diagnostics are briefly discussed. The structural features and principal characteristics of the diagnostic systems being developed or already being used on these devices are also examined. The different diagnostic methods are described according to the physical quantities to be measured: electric and magnetic diagnostics, measurements of electron density, electron temperature, the ion components of the plasma, radiation loss measurements, spectroscopy of impurities, edge diagnostics and study of plasma stability. The main parameters of the various diagnostic systems used on the six large tokamaks are summarized in tables. (author). 351 refs, 44 figs, 22 tabs

A new shape reproduction method based on the Cauchy-condition surface for real-time tokamak reactor control

International Nuclear Information System (INIS)

Kurihara, K.

2000-01-01

A new shape reproduction method is investigated on the basis of an applied mathematical approach. An analytically exact solution of Maxwell's equations in a static current field yields an (boundary) integral equation. In application of this equation to tokamak plasma shape reproduction, it is made clear that a Cauchy condition (both Dirichlet and Neumann conditions) on a hypothetical surface is necessarily identified. To calculate the Cauchy condition using magnetic sensor signals, conversion to numerical formulation of this method is conducted. Then, reproduction errors by this method are evaluated through two numerical tests: The first test uses ideal signals produced from a full equilibrium code in the JT-60 geometry, and the second test uses actual sensor signals in JT-60 experiments. In addition, it is shown that positioning and shape of the Cauchy condition surface is insensitive to reproduction error. Finally, this method is clarified to have preferable features for real-time tokamak reactor control

Magnetic field measurements of JT-60SA CS model coil

Energy Technology Data Exchange (ETDEWEB)

Obana, Tetsuhiro, E-mail: obana.tetsuhiro@LHD.nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Takahata, Kazuya; Hamaguchi, Shinji; Chikaraishi, Hirotaka; Mito, Toshiyuki; Imagawa, Shinsaku [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Kizu, Kaname; Murakami, Haruyuki; Natsume, Kyohei; Yoshida, Kiyoshi [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2015-01-15

Highlights: • Magnetic fields of the JT-60SA CS model coil were measured. • While the coil current was held constant at 20 kA, magnetic fields varied slightly with several different long time constants. • We investigated coils consisting of CIC conductors and having long time constants. - Abstract: In a cold test of the JT-60SA CS model coil, which has a quad-pancake configuration consisting of a Nb{sub 3}Sn cable-in-conduit (CIC) conductor, magnetic fields were measured using Hall sensors. For a holding coil current of 20 kA, measured magnetic fields varied slightly with long time constants in the range 17–571 s, which was much longer than the time constant derived from a measurement using a short straight sample. To validate the measurements, the magnetic fields of the model coil were calculated using a computational model representing the positions of Nb{sub 3}Sn strands inside the CIC conductor. The calculated results were in good agreement with the measurements. Consequently, the validity of the magnetic field measurements was confirmed. Next, we investigated other coils consisting of CIC conductors and having long time constants. The only commonality among the coils was the use of CIC conductors. At present, there is no obvious way to prevent generation of such magnetic-field variations with long time constants.

Fast reciprocating probe system for local scrape-off layer measurements in front of the lower hybrid launcher on JT-60U

International Nuclear Information System (INIS)

Asakura, N.; Tsuji-Iio, S.; Ikeda, Y.; Neyatani, Y.; Seki, M.

1995-01-01

A fast reciprocating probe system with a long drive shaft was incorporated into a multi-junction lower hybrid (LH) wave launcher on JT-60U in order to investigate an improved coupling mechanism of the radio frequency wave to the core plasma. The system has been operated reliably over a horizontal scan of 25 cm in 1.5 s using a compact pneumatic cylinder drive and springs. A double probe measurement provided the scrape-off layer plasma profile between the last closed flux surface and the first wall with the spatial resolution of 1-2 mm measured with a laser displacement gauge. The profiles of the electron density n e and temperature T e were in good agreement with those obtained with a triple probe method. During the LH wave injection with good coupling to the core plasma, an increase in the local T e was observed in front of the LH launcher mouth. The local n e was (7-10)x10 16 m -3 , consistent values needed for the good coupling. copyright 1995 American Institute of Physics

The effect of plasma minor-radius expansion in the current build-up phase of a large tokamak

International Nuclear Information System (INIS)

Kobayashi, Tomofumi; Tazima, Teruhiko; Tani, Keiji; Tamura, Sanae

1977-03-01

A plasma simulation code has been developed to study the plasma current build-up process in JT-60. Plasma simulation is made with a model which represents well overall plasma behavior of the present-day tokamaks. The external electric circuit is taken into consideration in simulation calculation. An emphasis is placed on the simulation of minor-radius expansion of the plasma and behavior of neutral particles in the plasma during current build-up. A calculation with typical parameters of JT-60 shows a week skin distribution in the current density and the electron temperature, if the minor radius of the plasma expands with build-up of the plasma current. (auth.)

JT-60SA vacuum vessel manufacturing and assembly

Energy Technology Data Exchange (ETDEWEB)

Masaki, Kei, E-mail: masaki.kei@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan); Shibama, Yusuke K.; Sakurai, Shinji; Shibanuma, Kiyoshi; Sakasai, Akira [Japan Atomic Energy Agency, Naka, Ibaraki-ken 311-0193 (Japan)

2012-08-15

Highlights: Black-Right-Pointing-Pointer The design of the JT-60SA vacuum vessel body was completed with the demonstration of manufacturing procedure by the mock-up fabrication of the 20 Degree-Sign upper half of VV. Black-Right-Pointing-Pointer The actual VV manufacturing has started since November 2009. Black-Right-Pointing-Pointer The first product of the VV 40 Degree-Sign sector was completed in May 2011. Black-Right-Pointing-Pointer A basic VV assembly scenario and procedure were studied to complete the 360 Degree-Sign VV including positioning method and joint welding. - Abstract: The JT-60SA vacuum vessel (VV) has a D-shaped poloidal cross section and a toroidal configuration with 10 Degree-Sign segmented facets. A double wall structure is adopted to ensure high rigidity at operational load and high toroidal one-turn resistance. The material is 316L stainless steel with low cobalt content (<0.05%). The design temperatures of the VV at plasma operation and baking are 50 Degree-Sign C and 200 Degree-Sign C, respectively. In the double wall, boric-acid water is circulated at plasma operation to reduce the nuclear heating of the superconducting magnets. For baking, nitrogen gas is circulated in the double wall after draining of the boric-acid water. The manufacturing of the VV started in November 2009 after a fundamental welding R and D and a trial manufacturing of 20 Degree-Sign upper half mock-up. The manufacturing of the first VV 40 Degree-Sign sector was completed in May 2011. A basic concept and required jigs of the VV assembly were studied. This paper describes the design and manufacturing of the vacuum vessel. A plan of VV assembly in torus hall is also presented.

Relationship between particle and heat transport in JT-60U plasmas with internal transport barrier

International Nuclear Information System (INIS)

Takenaga, Hidenobu; Higashijima, S.; Oyama, N.

2003-01-01

The relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated in reversed shear (RS) and high β p ELMy H-mode plasmas in JT-60U. No helium and carbon accumulation inside the ITB is observed even with ion heat transport reduced to a neoclassical level. On the other hand, the heavy impurity argon is accumulated inside the ITB. The argon density profile estimated from the soft x-ray profile is more peaked, by a factor of 2-4 in the RS plasma and of 1.6 in the high β p mode plasma, than the electron density profile. The helium diffusivity (D He ) and the ion thermal diffusivity (χ i ) are at an anomalous level in the high β p mode plasma, where D He and χ i are higher by a factor of 5-10 than the neoclassical value. In the RS plasma, D He is reduced from the anomalous to the neoclassical level, together with χ i . The carbon and argon density profiles calculated using the transport coefficients reduced to the neoclassical level only in the ITB are more peaked than the measured profiles, even when χ i is reduced to the neoclassical level. Argon exhaust from the inside of the ITB is demonstrated by applying ECH in the high β p mode plasma, where both electron and argon density profiles become flatter. The reduction of the neoclassical inward velocity for argon due to the reduction of density gradient is consistent with the experimental observation. In the RS plasma, the density gradient is not decreased by ECH and argon is not exhausted. These results suggest the importance of density control to suppress heavy impurity accumulation. (author)

Relationship between particle and heat transport in JT-60U plasmas with internal transport barrier

International Nuclear Information System (INIS)

Takenaga, H.; Higashijima, S.; Oyama, N.

2003-01-01

The relationship between particle and heat transport in an internal transport barrier (ITB) has been systematically investigated in reversed shear (RS) and high β p ELMy H-mode plasmas in JT-60U. No helium and carbon accumulation inside the ITB is observed even with ion heat transport reduced to a neoclassical level. On the other hand, the heavy impurity argon is accumulated inside the ITB. The argon density profile estimated from the soft x-ray profile is more peaked, by a factor of 2-4 in the RS plasma and of 1.6 in the high β p mode plasma, than the electron density profile. The helium diffusivity (D He ) and the ion thermal diffusivity (χ i ) are at an anomalous level in the high β p mode plasma, where D He and χ i are higher by a factor of 5-10 than the neoclassical value. In the RS plasma, D He is reduced from the anomalous to the neoclassical level, together with χ i . The carbon and argon density profiles calculated using the transport coefficients reduced to the neoclassical level only in the ITB are more peaked than the measured profiles, even when χ i is reduced to the neoclassical level. Argon exhaust from the inside of the ITB is demonstrated by applying ECH in the high β p mode plasma, where both electron and argon density profiles become flatter. The reduction of the neoclassical inward velocity for argon due to the reduction of density gradient is consistent with the experimental observation. In the RS plasma, the density gradient is not decreased by ECH and argon is not exhausted. These results suggest the importance of density gradient control to suppress heavy impurity accumulation. (author)

Analyses of erosion and re-deposition layers on graphite tiles used in the W-shaped divertor region of JT-60U

International Nuclear Information System (INIS)

Gotoh, Y.; Yagyu, J.; Masaki, K.; Kizu, K.; Kaminaga, A.; Kodama, K.; Arai, T.; Tanabe, T.; Miya, N.

2003-01-01

Erosion and re-deposition profiles were studied on graphite tiles used in the W-shaped divertor of JT-60U in June 1997-October 1998 periods, operated with all-carbon walls with boronizations and inner-private flux pumping. Continuous re-deposition layers were found neither on the dome top nor on the outer wing, while re-deposition layers of around 20 μm thickness were found on the inner wing, in the region close to the dome top. On the outer divertor target, erosion was found to be dominant: maximum erosion depth of around 20 μm was measured, while on the inner target, re-deposition was dominant: columnar structure layers of maximum thickness at around 30 μm on the inner zone while laminar/columnar-layered structures of maximum thickness around 60 μm were found on the outer zone. Poloidal distributions of the erosion depth/re-deposition layer thickness were well correlated with the frequency histograms of strike point position, which were weighted with total power of neutral beam injection, on both the outer and inner targets. Through X-ray photoelectron spectroscopy, composition of the re-deposition layers at a mid zone on the inner target were 3-4 at.% B and <0.6 at.% O, Fe, Cr, and Ni with remaining C. Boron atoms are mostly bound to C atoms but some may precipitated as boron

Improvement of the real-time processor in JT-60 data processing system

International Nuclear Information System (INIS)

Sakata, S.; Kiyono, K.; Sato, M.; Kominato, T.; Sueoka, M.; Hosoyama, H.; Kawamata, Y.

2009-01-01

Real-time processor, RTP is a basic subsystem in the JT-60 data processing system and plays an important role in JT-60 feedback control for plasma experiment. During the experiment, RTP acquires various diagnostic signals, processes them into a form of physical values, and transfers them as sensor signals to the particle supply and heating control supervisor for feedback control via reflective memory synchronization with 1 ms clock signals. After the start of RTP operation in 1997, to meet the demand for advanced plasma experiment, RTP had been improved continuously such as by addition of diagnostic signals with faster digitizers, reducing time for data transfer utilizing reflective memory instead of CAMAC. However, it is becoming increasingly difficult to maintain, manage, and improve the outdated RTP with limited system CPU capability. Currently, a prototype RTP system is being developed for the next real-time processing system, which is composed of clustered system utilizing VxWorks computer. The processes on the existing RTP system will be decentralized to the VxWorks computer to solve the issues of the existing RTP system. The prototype RTP system will start to operate in August 2008.

Overview of JT-60U results toward high integrated performance in reactor-relevant regime

International Nuclear Information System (INIS)

Fujita, T.

2003-01-01

Recent JT-60U results toward high integrated performance are reported with emphasis on the projection to the reactor-relevant regime. N-NB and EC power increased up to 6.2 MW and 3 MW, respectively. A high β p H-mode plasma with full non-inductive current drive has been obtained at 1.8 MA and the fusion triple product reached 3.1x10 20 m -3 keVs. High beta with β N =2.7 was maintained for 7.4 s. NTM suppression with EC was accomplished using a real-time feedback control system and improvement in β N was obtained. A stable existence of current hole was observed. High DT-equivalent fusion gain of 0.8 was maintained for 0.55 s in a plasma with a current hole. The current profile control in high bootstrap current reversed shear plasmas was demonstrated using N-NB and LH. A new operation scenario has been established in which a plasma with high bootstrap current fraction and ITBs is produced without the use of OH coil. ECCD study was undertaken in a reactor-relevant high T e regime. A new type of AE mode has been proposed and found to explain the observed frequency chirp quite well. High confinement reversed shear plasmas with T e >T i were obtained. Ar exhaust with EC heating was obtained in a high β p mode plasma. Impurity accumulation related to strong ITBs in a reversed shear plasma and degradation of ITB by ECH in a weak positive shear plasma have been found. Dedicated measurement of ELM dynamics and SOL plasma flow advanced the physics understanding. N-NB heating in an Ar-seed plasma extended the density region to 95% of Greenwald density with HH y2 =0.9. The enhancement of pedestal pressure was obtained with an increase of β p in a high triangularity configuration. (author)

The plasma movie database system for JT-60

International Nuclear Information System (INIS)

Sueoka, Michiharu; Kawamata, Yoichi; Kurihara, Kenichi; Seki, Akiyuki

2007-01-01

The real-time plasma movie with the computer graphics (CG) of plasma shape is one of the most effective methods to know what discharge have been made in the experiment. For an easy use of the movie in the data analysis, we have developed the plasma movie database system (PMDS), which automatically records plasma movie according to the JT-60 discharge sequence, and transfers the movie files on request from the web site. The file is compressed to about 8 MB/shot small enough to be transferred within a few seconds through local area network (LAN). In this report, we describe the developed system from the technical point of view, and discuss a future plan on the basis of advancing video technology

ERATO-code analysis of vacuum magnetic field oscillations in JT-60 divertor configuration

International Nuclear Information System (INIS)

Ozeki, Takahisa; Tokuda, Shinji; Tsunematsu, Toshihide; Ishida, Shinichi; Neyatani, Yuzuru; Itami, Kiyoshi; Azumi, Masafumi

1989-07-01

Magnetic field oscillations caused by external kink instabilities are numerically studied by using the ideal MHD stability code ERATO-J. Dependence of a spatial distribution of their amplitude and phase on aspect-ratio, beta-poloidal, shaping of conducting shell and divertor/limiter configurations is examined in detail. In the low aspect ratio plasma, the amplitude of magnetic oscillations in the inner side of the torus is larger than that in the outer. On the contrary, as the poloidal beta increases, the amplitude in the outer side of the torus becomes larger than that in the inner. In the divertor configuration, the amplitude of oscillations reduces near the X-point and the phase is locally modulated. The coherent magnetic oscillations observed in JT-60 agree well with the theoretical results, where the vacuum vessel is assumed to be an ideal conducting shell. The non-uniformity of the poloidal distribution observed in JT-60 can be explained by the combined effects of the finite beta, the X-point and the shape of shell. (author)

Conceptual design of the steady state tokamak reactor (SSTR)

International Nuclear Information System (INIS)

Oikawa, A.; Kikuchi, M.; Seki, Y.; Nishio, S.; Ando, T.; Ohara, Y.; Takizuka, Tani, K.; Ozeki, T.; Koizumi, K.; Ikeda, B.; Suzuki, Y.; Ueda, N.; Kageyama, T.; Yamada, M.; Mizoguchi, T.; Iida, F.; Ozawa, Y.; Mori, S.; Yamazaki, S.; Kobayashi, T.; Adachi, H.J.; Shinya, K.; Ozaki, A.; Asahara, M.; Konishi, K.; Yokogawa, N.

1992-01-01

This paper reports that on the basis of a high bootstrap current fraction observation with JT-60, the concept of steady state tokamak reactor , the SSTR, was conceived and was evolved with the design activity of the SSTR at JAERI. Also results of ITER/FER design activities has enhanced the SSTR design. Moreover the remarkable progress of R and D for fusion reactor engineering, especially in the development of superconducting coils and negative ion based NBI at JAERI have promoted the SSTR conceptual design as a realistic power reactor. Although present fusion power reactor designs are currently considered to be too large and costly, results of the SSTR conceptual design suggest that an efficient and promising tokamak reactor will be feasible. The conceptual design of the SSTR provides a realistic reference for a demo tokamak reactor

A coil test facility for the cryogenic tests of the JT-60SA TF coils

International Nuclear Information System (INIS)

Chantant, M.; Genini, L.; Bayetti, P.; Millet, F.; Wanner, M.; Massaut, V.; Corte, A. Della; Ardelier-Desage, F.; Catherine-Dumont, V.; Dael, A.; Decool, P.; Donati, A.; Duchateau, J.L.; Garibaldi, P.; Girard, S.; Hatchressian, J.C.; Fejoz, P.; Jamotton, P.; Jourdheuil, L.; Juster, F.P.

2011-01-01

In the framework of the Broader Approach Activities, the EU will deliver to Japan the 18 superconducting coils, which constitute the JT-60SA Toroidal field magnet. These 18 coils, manufactured by France and Italy, will be cold tested before shipping to Japan. For this purpose, the European Joint Undertaking for ITER, the Development of Fusion Energy ('Fusion for Energy', F4E) and the European Voluntary Contributors are collaborating to design and set-up a coil test facility (CTF) and to perform the acceptance test of the 18 JT-60SA Toroidal Field (TF) coils. The test facility is designed to test one coil at a time at nominal current and cryogenic temperature. The test of the first coil of each manufacturer includes a quench triggered by increasing the temperature. The project is presently in the detailed design phase.

Joint resistance measurements of pancake and terminal joints for JT-60SA EF coils

Energy Technology Data Exchange (ETDEWEB)

Obana, Tetsuhiro, E-mail: obana.tetsuhiro@LHD.nifs.ac.jp [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Takahata, Kazuya; Hamaguchi, Shinji; Mito, Toshiyuki; Imagawa, Shinsaku [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan); Kizu, Kaname; Murakami, Haruyuki; Yoshida, Kiyoshi [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan)

2013-11-15

Highlights: • To evaluate the joint fabrication technology for the JT-60SA EF coils, joint resistance measurements were conducted with a joint sample. • The joint sample was composed of pancake and terminal joints. • The measurements demonstrated that both joints fulfilled the design requirement. • Considering the measurements, the characteristics of both joints were investigated using an analytical model that represents the joints. -- Abstract: To evaluate the joint fabrication technology for the JT-60SA EF coils, joint resistance measurements were conducted using a sample consisting of pancake and terminal joints. Both joints are shake-hands lap joints composed of cable-in-conduit conductors and a pure copper saddle-shaped spacer. The measurements demonstrated that both joints fulfilled the design requirement. Considering these measurements, the characteristics of both joints were investigated using analytical models that represent the joints. The analyses indicated that the characteristics of the conductors used in the joints affect the characteristics of the joints.

Fabrication and performance tests of a prototype in-situ coating machine for JT-60

International Nuclear Information System (INIS)

Obara, Kenjiro; Abe, Tetsuya; Murakami, Yoshio

1987-09-01

Prior to the design and construction of the JT-60's in-situ coating device, a prototype machine was fabricated and tested to confirm the applicability of proposed driving methods and mechanical elements to the device which would be operated in very severe conditions including high ambient temperature and high vacuum. The machine basically consists of an in-vessel manipulator, a fiberscope and an ohmically heated titanium evaporator. From the test results, we recommended to use the combination of Inconel 625 and a self-lubricating alloy for the solid-lubricated bearings and MoS 2 -coated Inconel 625 for the solid-lubricated gears. It was also found that TiC coating showed a effect for the prevention of welding between bolts and nuts. In order to optimize the operating parameters of the machine, many wall inspection tests and titanium evaporation tests were carried out in a large vacuum vessel by simulating the JT-60 conditions. (author)

Effect of resistivity profile on current decay time of initial phase of current quench in neon-gas-puff inducing disruptions of JT-60U

Energy Technology Data Exchange (ETDEWEB)

Kawakami, S.; Ohno, N. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Shibata, Y.; Isayama, A.; Kawano, Y. [Japan Atomic Energy Agency, Naka 311-0193 (Japan); Watanabe, K. Y. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); National Institute for Fusion Science, Toki 509-5292 (Japan); Takizuka, T. [Graduate School of Engineering, Osaka University, Suita 565-0871 (Japan); Okamoto, M. [Ishikawa National College of Technology, Ishikawa 929-0392 (Japan)

2013-11-15

According to an early work [Y. Shibata et al., Nucl. Fusion 50, 025015 (2010)] on the behavior of the plasma current decay in the JT-60U disruptive discharges caused by the radiative collapse with a massive neon-gas-puff, the increase of the internal inductance mainly determined the current decay time of plasma current during the initial phase of current quench. To investigate what determines the increase of the internal inductance, we focus attention on the relationship between the electron temperature (or the resistivity) profile and the time evolution of the current density profile and carry out numerical calculations. As a result, we find the reason of the increase of the internal inductance: The current density profile at the start of the current quench is broader than an expected current density profile in the steady state, which is determined by the temperature (or resistivity) profile. The current density profile evolves into peaked one and the internal inductance is increasing.

Review of JT-60U experimental results in 1998

International Nuclear Information System (INIS)

Adachi, H.; Akasaka, H.; Akino, N.

1999-09-01

Based on the high confinement regimes such as reversed shear mode, high-β p mode and H-mode, the JT-60U experiment in 1998 was devoted to expand the discharge regimes in terms of 1) achievement of high fusion gain, 2) concept optimization for long sustainment of the advanced modes for >>τ E and >τ p * with the current profile close to the steady-state solution, 3) high confinement by electron heating (T e >T i ), 4) high confinement at high electron density and/or at high divertor radiation and 5) active control of multiple parameters including both core and divertor plasmas. As for the reversed shear mode, high performance discharges satisfying Q DT eq (DT equivalent fusion gain ) >1 were obtained reproducibly and the record value of Q DT eq =1.25 was achieved in a reactor-relevant thermonuclear dominant regime due to the optimized discharge scenario using feedback control of the neutron production rate where β-values were kept in the MHD stable region during the I p ramping phase with a large radius of the internal transport barrier (ITB). The reduction of Z eff obtained after installation of W-shaped pumped divertor increased fusion reaction rate. Concerning long sustainment, the reversed shear ELMy H-mode with H 89PL -factor∼1.5-2 and β N =1.0-1.4 was kept for 5.5 s with NB heating. By off axis LH current drive, the reversed shear current profile with the ITB was kept constant for 4.7 s under full non- inductive current drive condition (LHCD=77%, bootstrap=23%) at T e -1.2T i . In the high-β p ELMy H-mode regime, benefits of the high triangularity shape were demonstrated. At a high triangularity δ X ∼0.46, β N =2.5-2.7 was sustained for 3.5 s even at the low value of q 95 =2.9-3.3. The product of β N xH-factor sustainable for >5τ E (>τ p *) increases with δ x and reaches ∼6 at δ X ∼0.46. In addition to extension of the discharge regimes, the key phenomena determining transport and stability around ITBs were studied intensively. For the

Review of JT-60U experimental results in 2000

International Nuclear Information System (INIS)

2002-11-01

The results of JT-60U experiments in 2000, from February to December, are reviewed. The performance under the full non-inductive current drive has been greatly advanced both in high β p H-mode plasmas and in reversed shear (RS) plasmas. In the high β p H-mode, with injection of the negative ion based neutral beam (NNB) of 360 keV and 4 MW into a high electron temperature plasma (T e (0) ∼ 13 keV), a high fusion triple product n D (0)T i (0)τ E = 2.0 x 10 20 keVm -3 s has been obtained at the plasma current of 1.5 MA, and the highest value of current drive efficiency of NNB (1.55 x 10 19 A/W/m 2 ) has been achieved. In RS, LHCD and NNB-CD were employed for current drive and high confinement (HH y2 ∼ 1.4) at high density (f GW ∼ 0.80) has been achieved. In the 110 GHz ECRF system, two more gyrotrons have been installed in addition to the one installed in 1999 and 1.5 MW was injected into the torus for 3 s. Complete stabilization of neoclassical tearing mode (NTM), realization of high confinement plasmas with T e ∼ T i , a high value (1 MA) of NNB-driven current in the high T e regime, and measurement of localized EC driven current were achieved with the upgraded EC system. The multiple pellet injection system has been newly installed. In high power NB heated plasmas, high-field-side pellet injection was more effective than low-field-side injection, and it extended the regime of high confinement high β p H-mode to a higher density. A new method, called CCS (Cauchy-condition surface method), for the control of the plasma position and shape in real-time became available and was found very useful especially for the control of plasma-wall clearance in LH and IC experiments. The active control of ITB strength by the switch of the injection direction of toroidal angular momentum was successfully demonstrated. In high triangularity H-mode plasmas, higher pressure and temperature at the edge pedestal were observed, which resulted in higher temperature and

Heat analysis of the magnetic limiter plate for JT-60

International Nuclear Information System (INIS)

Nakamura, Hiroo; Ninomiya, Hiromasa; Shimizu, Masatsugu; Ohta, Mitsuru

1977-03-01

Heat analysis has been made of the magnetic limiter plate for JT-60. Test materials of the magnetic limiter plate are molybdenum, graphite, pyrolytic graphite and silicon carbide. It is assumed in calculation of the heat analysis that 10MW is deposited on the 2 cm wide surface of the magnetic limiter plate in about 10 sec. The magnetic limiter plate of pyrolytic graphite is a stack of pyrolytic graphite sheets, heat input is in the deposition plane to take advantage of the large heat conductivity along this plane. Pyrolytic graphite is the best in terms of temperature rise. The temperature of molybdenum and graphite rise up to 1800 0 C and 620 0 C, respectively, in an deposition of 10 MWx10sec. Silicon carbide is not suitable for the magnetic limiter plate. Because the plasma of the JT-60 discharges every 10 min, the average heat flux decreases to 17 w/cm 2 during the each interval. When the magnetic limiter plate has the above heat inflow, a maximum of above 1000 0 C occurs at the edge far from the joint to the thick ring of the vacuum vessel. To reduce heat load of the magnetic limiter plate, an alternating current (2 -- 5Hz) is superposed on the magnetic limiter coil current. The intersection of separatrix line and magnetic limiter plate then moves cyclically more than 10 cm. Concerning temperature distribution of the multi-groove magnetic limiter plate, its dimensions are determined by the limitation in vapor pressure to prevent the impurity inflow. (auth.)

Improvement of voltage holding capability in the 500 keV negative ion source for JT-60SA.

Science.gov (United States)

Tanaka, Y; Hanada, M; Kojima, A; Akino, N; Shimizu, T; Ohshima, K; Inoue, T; Watanabe, K; Taniguchi, M; Kashiwagi, M; Umeda, N; Tobari, H; Grisham, L R

2010-02-01

Voltage holding capability of JT-60 negative ion source that has a large electrostatic negative ion accelerator with 45 cm x 1.1 m acceleration grids was experimentally examined and improved to realize 500 keV, 22 A, and 100 s D- ion beams for JT-60 Super Advanced. The gap lengths in the acceleration stages were extended to reduce electric fields in a gap between the large grids and at the corner of the support flanges from the original 4-5 to 3-4 kV/mm. As a result, the voltage holding capability without beam acceleration has been successfully improved from 400 to 500 kV. The pulse duration to hold 500 kV reached 40 s of the power supply limitation.

Review of JT-60U experimental results in 2001 and 2002

International Nuclear Information System (INIS)

2003-11-01

The results from the JT-60U experiments in 2001 and 2002 are reviewed. A high poloidal beta (β p ) H-mode plasma with full non-inductive current drive was obtained at 1.8 MA, and the fusion triple product reached 3.1x10 20 m -3 keV s. High normalized beta (β N ) around 2.7 was maintained for 7.4 s. Neoclassical tearing mode suppression by electron cyclotron current drive (ECCD) was accomplished using a real-time feedback control system, and improvement in β N was obtained. In a reversed shear plasma, a high DT-equivalent fusion gain (>0.8) was maintained for 0.55 s with a current hole. High confinement reversed shear plasmas where the electron temperature was higher than the ion temperature were obtained. A new operation scenario was established in which a plasma with high bootstrap current fraction and internal transport barriers (ITBs) was produced without using the OH coil. For physics studies of ITB formation, dedicated experiment was performed, in which the heating power was deliberately scanned. Accumulation of heavy impurity related to strong ITBs in reversed shear plasmas, and degradation of ITBs by electron cyclotron heating (ECH) in weak positive shear plasmas was found. In high β p mode plasmas, Ar exhaust by ECH was obtained. Correlation between the particle and thermal diffusivities was observed in ITB plasmas with both negative and weak shear. As for the current hole, the mechanism was investigated in detail by accurate current profile measurement and ECCD. A new type of Alfven eigenmode (Reversed shear Alfven eigenmode) was proposed, and the observed frequency chirp was explained quite well. As for ECCD, the current drive efficiency was evaluated in a reactor-relevant high electron-temperature regime. In Ar-seeded H-mode plasmas with the outer strike point located on the divertor dome-top, high-power negative-ion-based neutral beam injection (N-NBI) extended the electron density regime to 0.95 Greenwald density keeping confinement enhancement

Examples of data processing systems. Data processing system for JT-60

International Nuclear Information System (INIS)

Aoyagi, Tetsuo

1996-01-01

JT-60 data processing system is a large computer complex system including a lot of micro-computers, several mini-computers, and a main-frame computer. As general introduction of the original system configuration has been published previously, some improvements are described here. Transient mass data storage system, network database server, a data acquisition system using engineering workstations, and a graphic terminal emulator for X-Window are presented. These new features are realized by utilizing recent progress in computer and network technology and carefully designed user interface specification of the original system. (author)

Magnetic Fluctuations during plasma current rise of divertor discharge in JT-60

International Nuclear Information System (INIS)

Ushigusa, Kenkichi; Kikuchi, Mitsuru; Hosogane, Nobuyuki; Tsuji, Syunji; Hayashi, Kazuo.

1986-03-01

During a current rise phase in the JT-60 divertor discharge, a series of magnetic fluctuations which do not rotate poloidally (phase-locking) is observed. They cause a cooling of plasma periphery and an enhancement of H α emission in the divertor chamber. A significant increase in β P + 1 i /2 with minor disruptions during the phase-locked magnetic fluctuation suggests a relaxation of the current profile in the current rise phase of the divertor discharge. (author)

The physics of tokamak start-up

International Nuclear Information System (INIS)

Mueller, D.

2013-01-01

Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current

Development of the piezoelectric gas injection valve for JT-60

International Nuclear Information System (INIS)

Kawasaki, Kazuo; Hiratuka, Hajime

1986-01-01

Piezoelectric gas injection valve (PEV) for JT-60 have been developed which was a piezo-electric element. The raliability of the PEV under the actual condition of high magnetic fields and high temperatures are veryfied, and it became clear that the PEV had enough throughput range and sufficient repetability for long life throughput characteristics. Remarkables of the developed PEV are summarized as follows, (1) The maximum throughput rate, responce time and helium leakage rate satisfy the desiged specifications. (2) Throughput equation for PEV is clarified by comparison with experiment. (3) Reliabilities of PEV under the actual condition during coil power test become clear. (author)

Development of residual thermal stress-relieving structure of CFC monoblock target for JT-60SA divertor

Energy Technology Data Exchange (ETDEWEB)

Tsuru, Daigo, E-mail: tsuru.daigo@jaea.go.jp; Sakurai, Shinji; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Suzuki, Satoshi

2015-10-15

Highlights: • We carried out numerical simulations on residual thermal stress of targets for the JT-60SA divertor. • We developed three measures to reduce residual thermal stress. • We proposed two structures of CFC monoblock target for the JT-60SA divertor. • We confirmed the effectiveness of the structure by infrared thermography inspection and high heat flux test. - Abstract: Carbon fibre-reinforced carbon composite (CFC) monoblock target for JT-60SA divertor is under development towards the mass-production. CFC monoblocks, WCu interlayers and a CuCrZr cooling tube at the centre of the monoblocks were bonded by vacuum brazing in a high temperature, to a target. If residual thermal stress due to difference of thermal expansions between CFC and CuCrZr exceeds the maximum allowable stress of the CFC after the bonding, cracks are generated in the CFC monoblock and heat removal capacity of the target degrades. In this paper, new structures of the targets were proposed, to reduce residual thermal stress and to mitigate the degradation of heat removal capacity of the targets. Some measures, including slitting of the CFC monoblock aside of the cooling tube, replacement of the interlayer material and shifting the position of the cooling tube, were implemented. The effectiveness of the measures was evaluated by numerical simulations. Target mock-ups with the proposed structures were manufactured. Infrared thermography inspection and high heat flux test were carried out on the mock-ups in order to evaluate the heat removal capacity.

Effective bending strain estimated from I c test results of a D-shaped Nb3Al CICC coil fabricated with a react-and-wind process for the National Centralized Tokamak

International Nuclear Information System (INIS)

Ando, T.; Kizu, K.; Miura, Y.M.; Tsuchiya, K.; Matsukawa, M.; Tamai, H.; Ishida, S.; Koizumi, N.; Okuno, K.

2005-01-01

Japan National Centralized Tokamak (NCT) is a superconducting tokamak proposed as a modification to JT-60U. As part of the R and D for the National Centralized Tokamak, a two-turn, approximately 2 m tall, D-shaped Nb 3 Al coil was wound and tested using a full-size cable-in-conduit conductor (CICC). The Nb 3 Al cable-in-conductor was bent following the heat treatment reaction with a maximum bending strain of 0.4% to simulate the react-and-wind fabrication. The comparison of the coil performance to the measured strand data shows that the effective axial strain of the conductor strands is essentially zero despite the 0.4% bending strain of the conductor. This suggests that the strands in the cable slipped relatively to each other during bending of the conduit, thus reducing the effective strain transmitted to the strands. This result is very encouraging for the low-cost fabrication of high-current-density fusion coils using the react-and-wind method

Review of JT-60U experimental results in 1998

Energy Technology Data Exchange (ETDEWEB)

Adachi, H.; Akasaka, H.; Akino, N. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1999-09-01

Based on the high confinement regimes such as reversed shear mode, high-{beta}{sub p} mode and H-mode, the JT-60U experiment in 1998 was devoted to expand the discharge regimes in terms of 1) achievement of high fusion gain, 2) concept optimization for long sustainment of the advanced modes for >>{tau}{sub E} and >{tau}{sub p}* with the current profile close to the steady-state solution, 3) high confinement by electron heating (T{sub e}>T{sub i}), 4) high confinement at high electron density and/or at high divertor radiation and 5) active control of multiple parameters including both core and divertor plasmas. As for the reversed shear mode, high performance discharges satisfying Q{sub DT}{sup eq} (DT equivalent fusion gain ) >1 were obtained reproducibly and the record value of Q{sub DT}{sup eq}=1.25 was achieved in a reactor-relevant thermonuclear dominant regime due to the optimized discharge scenario using feedback control of the neutron production rate where {beta}-values were kept in the MHD stable region during the I{sub p} ramping phase with a large radius of the internal transport barrier (ITB). The reduction of Z{sub eff} obtained after installation of W-shaped pumped divertor increased fusion reaction rate. Concerning long sustainment, the reversed shear ELMy H-mode with H{sup 89PL}-factor{approx}1.5-2 and {beta}{sub N}=1.0-1.4 was kept for 5.5 s with NB heating. By off axis LH current drive, the reversed shear current profile with the ITB was kept constant for 4.7 s under full non- inductive current drive condition (LHCD=77%, bootstrap=23%) at T{sub e}-1.2T{sub i}. In the high-{beta}{sub p} ELMy H-mode regime, benefits of the high triangularity shape were demonstrated. At a high triangularity {delta}{sub X}{approx}0.46, {beta}{sub N}=2.5-2.7 was sustained for 3.5 s even at the low value of q{sub 95}=2.9-3.3. The product of {beta}{sub N}xH-factor sustainable for >5{tau}{sub E} (>{tau}{sub p}*) increases with {delta}{sub x} and reaches {approx}6 at {delta

Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design

International Nuclear Information System (INIS)

Qiu Qing-Lai; Xiao Bing-Jia; Guo Yong; Liu Lei; Wang Yue-Hang

2017-01-01

Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other. (paper)

Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

International Nuclear Information System (INIS)

Kawashima, H.; Sakurai, S.; Shimizu, K.; Takizuka, T.; Tamai, H.; Matsukawa, M.; Fujita, T.; Miura, Y.

2006-01-01

The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2 . Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design

HT-7U superconducting tokamak: Physics design, engineering progress and schedule

International Nuclear Information System (INIS)

Wan Yuanxi

2002-01-01

The superconducting tokamak research program begun in China in ASIPP since 1994. The program is included in existent superconducting tokamak HT-7 and the next new superconducting tokamak HT-7U which is one of national key research projects in China. With the elongation cross-section, divertor and higher plasma parameter the main objectives of HT-7U are widely investigation both of the physics and technology for steady state advanced tokamak as well as the investigation of power and particle handle under steady-state operation condition. The physics and engineering design have been completed and significant progresses on R and D and fabrication have been achieved. HT-7U will begin assembly at 2003 and possible to get first plasma around 2004. (author)

Small ELM regimes with good confinement on JET and comparison to those on ASDEX Upgrade, Alcator C-mod, and JT-60U

International Nuclear Information System (INIS)

Stober, J.; Lomas, P.; Saibene, G.

2005-01-01

Since it is uncertain if ITER operation is compatible with type-I ELMs, the study of alternative H-mode pedestals is an urgent issue. This paper reports on experiments on JET aiming to find scenarios with small ELMs and good confinement, such as the type-II ELMs in ASDEX Upgrade, the enhanced D-alpha H-mode in Alcator C-mod or the grassy ELMs in JT-60U. The study includes shape variations, especially the closeness to a double-null configuration, variations of q 95 , density and beta poloidal. H-mode pedestals without type-I ELMs have been observed only at the lowest currents (≤ 1.2 MA), showing similarities to the observations in the devices mentioned above. These are discussed in detail on the basis of edge fluctuation analysis. For higher currents, only the mixed type-I/II scenario is observed. Although the increased inter-ELM transport reduces the type-I ELM frequency, a single type-I ELM is not significantly reduced in size. Obviously, these results do question the accessibility of such small ELM scenarios on ITER, except perhaps the high beta-poloidal scenario at higher q 95 , which could not be tested at higher currents at JET due to limitations in heating power. (author)

Development of a VME multi-processor system for plasma control at the JT-60 Upgrade

International Nuclear Information System (INIS)

Takahashi, M.; Kurihara, K.; Kawamata, Y.; Akasaka, H.; Kimura, T.

1992-01-01

Design and initial operation results are reported of a VME multi-processor system [1] for plasma control at a large fusion device named 'the JT-60 Upgrade' utilizing three 32-bit MC88100 based RISC computers and VME components. Development of the system was stimulated by faster and more accurate computation requirements for the plasma position and current control. The RISC computers operate at 25 MHz along with two cashe memories named MC88200. We newly developed VME bus modules of up/down counter, analog-to-digital converter and clock pulse generator for measuring magnetic field and coil current and for synchronizing the processing in the three RISCs and direct digital controllers (DDCs) of magnet power supplies. We also evaluated that the speed of the data transfer between the VME bus system and the DDCs through CAMAC highways satisfies the above requirements. In the initial operation of the JT-60 upgrade, it has been proved that the VME multi-processor system well controls the plasma position and current with a sampling period of 250 μsec and a delay of 500 μsec. (author)

Monte Carlo approach to define the refrigerator capacities for JT-60SA

International Nuclear Information System (INIS)

Wanner, Manfred; Barabaschi, Pietro; Lamaison, Valerie; Michel, Frederic; Reynaud, Pascal; Roussel, Pascal

2011-01-01

The JT-60SA cryogenic system shall provide refrigeration to keep the superconducting magnets and their structures at 4.4 K, cryo-pumps at 3.7 K, thermal shields at 80-100 K, and deliver a flow of 50 K helium to the current leads. A Monte Carlo method is proposed to determine the capacity contingencies for the refrigeration system. Attributing individual contingencies and distribution probability functions to the design variables allows the different load contributions to be statistically averaged. The total refrigeration contingency is derived for each temperature level from the 95% confidence level of the integrated distribution function.

Review of JT-60 experimental results from June to October, 1987

International Nuclear Information System (INIS)

1988-03-01

This is a prompt report on JT-60 experimental results from June to October, 1987. Experiments in hydrogen plasmas have been undertaken with up to 3.2 MA of plasma current in limiter discharges, 2.7 MA in divertor discharges, and 30 MW of total NB (H 0 → H + ) and RF injection power. In limiter discharges with ∼3 MA plasma current, the maximum central electron density of 1.3 x 10 20 m -3 and the energy confinement time of 0.15 - 0.18 sec were obtained with heating powers of 13 - 20 MW. The total plasma stored energy of 3.1 MJ, and n e (0)τ E of 1.4 - 1.8 x 10 19 m -3 sec were achieved. The best set of parameters achieved is n e (0)τ E = 1.8 x 10 19 m -3 sec and T i (0) = 3.7 KeV at plasma current of 3.2 MA. By applying deutrium discharge condition, n e (0)τ E and T i (0) enter the JT-60 target area determined by Atomic Energy Commission. Many short periods (50 - 100 ms) of H-mode phase are found in outside X-point divertor discharges with NB or NB + RF (LH or IC) heating power of more than 16 MW, although at present the energy confinement improvement is limited to within 10 %. In combined LH and NB heating of low-n-bar e discharge, the plasma stored energy increases with the same rate as NB heating only, with preferential absorption of LH wave to the high energy beam ions. In combined NB and on-axis ICRF heating of low-n-bar e discharge, a high incremental energy confinement time is obtained, with strong high energy beam ion acceleration and electron heating in the central region of the plasma. (author)

Tokamaks: from A D Sakharov to the present (the 60-year history of tokamaks)

International Nuclear Information System (INIS)

Azizov, E A

2012-01-01

The paper is prepared on the basis of the report presented at the session of the Physical Sciences Division of the Russian Academy of Sciences (RAS) at the Lebedev Physical Institute, RAS on 25 May 2011, devoted to the 90-year jubilee of Academician Andrei D Sakharov - the initiator of controlled nuclear fusion research in the USSR. The 60-year history of plasma research work in toroidal devices with a longitudinal magnetic field suggested by Andrei D Sakharov and Igor E Tamm in 1950 for the confinement of fusion plasma and known at present as tokamaks is described in brief. The recent (2006) agreement among Russia, the EU, the USA, Japan, China, the Republic of Korea, and India on the joint construction of the international thermonuclear experimental reactor (ITER) in France based on the tokamak concept is discussed. Prospects for using the tokamak as a thermonuclear (14 MeV) neutron source are examined. (conferences and symposia)

Review of JT-60 experimental results from January to March, 1987

International Nuclear Information System (INIS)

1987-08-01

This is a prompt report on JT-60 experimental results from January to March, 1987. Maximum LHRF injection power of 6.3 MW was achieved by using 3 launchers. The correlation between flattening of plasma current profile and energy confinement improvement was demonstrated by controlling N 11 spectrum of current drive and heating launchers. The maximum electron temperature of 6 kW was achieved with 2.4 MW LH injection in electron heating mode. Non-inductive driven current of 2 MA with 2.5 sec duration was demonstrated with 3 MW LHRF injection. Substantial acceleration of NB injected fast ions up to 160 kW was observed during ICRF + NB (60 keV) injection. In low density high power NB injected limiter discharges, high ion temperature of ∼ 11 keV and electron temperature of 5 keV were obtained with Z eff = 5. Various attempts were made to produce H-mode in divertor and limiter discharges. (author)

Status of JT-60 data processing system

International Nuclear Information System (INIS)

Matsuda, T.; Tsugita, T.; Oshima, T.; Sakata, S.; Sato, M.; Koiwa, M.; Aoyagi, T.

2000-01-01

The JT-60 data processing system is a large computer complex and gradually modernized by utilizing progressing computer and network technology. There are two major changes in our system. A main computer of FACOM M-780 has been replaced with compatible GS8300 using state-of-art CMOS technology, which results in lower power and space usage with nearly the same performance. Now it can handle ∼500 MB of data per discharge. A gigabit ethernet switch with FDDI ports has been introduced to cope with the increase of handling data. The switch will connect a tera-byte (TB) data server at the bandwidth of a gigabit per second with the main computer and many data acquisition workstations. Other developments in our system are the realization of three workstation-based plans, the TB data server, the VME-based fast data acquisition system and a CICU. The TB data server is basically a UNIX workstation with ∼100 GB RAID disks and ∼900 GB MO auto-exchangers. The VME-based fast data acquisition system has been developed to enlarge the present TMDS. The CICU, which has a function of interfacing the main computer with the CAMAC system, has been replaced with the workstation-based system after the fine tuning

Evaluation of mechanical strength of the joints in JT-60 toroidal field coil conductors

International Nuclear Information System (INIS)

Nishio, Satoshi; Ohkubo, Monoru; Sasajima, Hiroshi

1980-04-01

Toroidal field (TF) coils of JT-60 produce a toroidal field of 45 kG at a plasma axis, they have an inner bore of 3.90 m and a weight of about 80 metric tons per coil. Eighteen TF coils are located around a torus axis at regular intervals. TF coil conductors are mostly jointed by high frequency induction brazing, the rest jointed by welding. In deciding the details of the jointing procedures, the conductor size and the requested mechanical strength are mainly taken into consideration. Described are non-destructive inspection methods for the brazed joints, strength evaluation, and the inspection criteria. Ultrasonic testing method is found to be the most effective in evaluation of mechanical properties of the brazed joints especially in terms of fatigue strength. In section 1, specifications of the TF coils are given. In section 2, the ultrasonic inspection method and the detectability of this apparatus are described in detail, the defects of known size are compared with the indication values and display figures. The apparatus developed for JT-60 is operated automatically also recording the inspectionresults. In section 3, mechanical strength of the brazed joints with initial defects is discussed on the basis of Fracture Mechanics theory and results of the fatigue crack growth test. The inspection criteria in accordance with the descriptions of section 2 and 3 are given in section 4. (author)

Operation experiences of the JT-60 first walls during high-power additional heating experiments

International Nuclear Information System (INIS)

Takatsu, H.; Ando, T.; Yamamoto, M.; Arai, T.; Kodama, K.; Suzuki, M.; Shimizu, M.

1989-01-01

JT-60 started its operation in May 1985 with TiC-coated molybdenum or Inconel 625 first walls. They provided very clean surfaces as well as superior plasma characteristics during Joule heating discharges. Though 20 μm-thick TiC coatings showed good adhesion characteristics, melting of the TiC coating and also the molybdenum or Inconel 625 substrate was observed at some specific spots, and an influx of heavy metals to the main plasma was inevitable during discharges. Initial results of the additional heating experiments showed degrading effects of locally melted TiC-coated molybdenum or Inconel 625 on plasma operation. Therefore, about a half of the TiC-coated first walls were removed and new graphite first walls were installed during the venting period from April to May 1987. The start-up of the discharge conditioning after installation of a significant number of graphite tiles was very rapid. Flexibility in plasma operation was increased, and JT-60 extended the operation region beyond its original specifications. The graphite first walls of the main chamber performed admirably and maintained their integrity under the conditions of plasma current and additional heating power up to 3.2 MA and 30 MW, respectively. On the other hand, the number of damaged divertor plates was much larger than that expected. The reason of unexpected failure is now under examination. (orig.)

Development of a pellet cutting and loading device for the JT-60 repetitive pellet injector

International Nuclear Information System (INIS)

Hiratsuka, Hajime; Ichige, Hisashi; Kizu, Kaname; Iwahashi, Takaaki; Honda, Masao

2001-03-01

In JT-60, a pellet injector that repetitively injects deuterium pellets is under development to supply fuel to high temperature plasmas and sustain high-density plasmas. The pellet injector generates cubic pellets and accelerates them with a straight-arm rotor by centrifugal force. In this acceleration method, it is important to supply pellets reliably and stably, to prevent pellet orbits from disordering and to stabilize the launching direction. To achieve higher performance of the injector, a pellet cutting and loading device that cuts a deuterium ice rod into cubic pellets and loads them to the pellet injector successively and stably has been developed. The pellet cutting and loading device can cut a deuterium ice rod produced at low temperature of -8 Pam 3 /s, cutting time of <3 ms, cutting frequency of 1-20 Hz and cutter stroke of 2.5 mm were confirmed in the device test. In the operation test after assembling this device to the centrifugal pellet injector, the operational performance of pellet injection frequency of ∼10 Hz, pellet speed of ∼690 m/s and pellet injection duration time of ∼3.5 s was achieved. Thus, the development of the pellet cutting and loading device contributed to the upgrade of the JT-60 pellet injector. (author)

Axisymmetric MHD simulation of ITB crash and following disruption dynamics of Tokamak plasmas with high bootstrap current

International Nuclear Information System (INIS)

Takei, Nahoko; Tsutsui, Hiroaki; Tsuji-Iio, Shunji; Shimada, Ryuichi; Nakamura, Yukiharu; Kawano, Yasunori; Ozeki, Takahisa; Tobita, Kenji; Sugihara, Masayoshi

2004-01-01

Axisymmetric MHD simulation using the Tokamak Simulation Code demonstrated detailed disruption dynamics triggered by a crash of internal transport barrier in high bootstrap current, high β, reversed shear plasmas. Self-consistent time-evolutions of ohmic current bootstrap current and induced loop voltage profiles inside the disrupting plasma were shown from a view point of disruption characterization and mitigation. In contrast with positive shear plasmas, a particular feature of high bootstrap current reversed shear plasma disruption was computed to be a significant change of plasma current profile, which is normally caused due to resistive diffusion of the electric field induced by the crash of internal transport barrier in a region wider than the internal transport barrier. Discussion based on the simulation results was made on the fastest record of the plasma current quench observed in JT-60U reversed shear plasma disruptions. (author)

Improvement on control system of the JT-60 radio frequency heating system

Energy Technology Data Exchange (ETDEWEB)

Shinozaki, Shin-ichi; Moriyama, Shinichi; Hiranai, Shinichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sato, Fumiaki [Nippon Advanced Technology Co., Ltd., Tokai, Ibaraki (Japan)

2003-03-01

On the JT-60 radio frequency (RF) heating system, the decrease in the activity ratio was a problem because of the deterioration of the control system. To improve the reliability, we replaced CAMAC system for a power injection control system, which was a main cause of the trouble, with the microprocessor system. And, a function of computer supported programming function of RF power injection form was introduced, which contributed to reduce a load of operators. Furthermore, personal computers with network communication were introduced to improve a maintenance ability of the control system. As a result, the activity ratio of the RF heating system was improved significantly. (author)

Estimation of the lifetime of resin insulators against baking temperature for JT-60SA in-vessel coils

Energy Technology Data Exchange (ETDEWEB)

Sukegawa, Atsuhiko M., E-mail: morioka.atsuhiko@jaea.go.jp; Murakami, Haruyuki; Matsunaga, Go; Sakurai, Shinji; Takechi, Manabu; Yoshida, Kiyoshi; Ikeda, Yoshitaka

2015-10-15

Highlights: • The lifetime of resin insulators at about 200 °C was estimated. • We make use of the Arrhenius plot by the Weibull analysis for the estimation. • A suitable temperatures for the in-vessel coils were discussed. - Abstract: In the present study, the thermal endurance of epoxy-based, bismaleimides, and cyanate ester resins for the current design of the in-vessel coils was measured by performing acceleration tests to assess their insulation properties using the thermal endurance defined by the International Electrotechnical Commission (IEC-60216 Part1–Part 6) for a minimum of 5,000 h in the 180–240 °C temperature range. It was found that none of the resin insulators could tolerate the baking conditions of 40,000 h at ∼200 °C in the JT-60SA vacuum vessel. Therefore, the design of the in-vessel coils, including the error field correction coils (EFCC), was changed from the type without water cooling to with water cooling on JT-60SA.

Neutronics design of the next tokamak. (Swimming pool type)

International Nuclear Information System (INIS)

Seki, Y.; Iida, H.; Kitamura, K.; Minato, A.; Sako, K.; Mori, S.; Nishida, H.

1983-01-01

A swimming pool type tokamak reactor (SPTR) has been proposed in the Japan Atomic Energy Research Institute as a candidate for the next generation tokamak reactor after the JT-60. The concept of the SPTR evolved from an incentive to relieve the difficulties of repair and maintenance procedures of a tokamak reactor. After about two years of the reactor design studies, several advantages of the SPTR over the conventional tokamak reactors such as the ease of penetration shielding, reduction in solid radwaste have been shown. On the other hand, some drawbacks and uncertainties of the SPTR have also been pointed out but so far no serious defect negating the concept has been found. This paper describes the neutronics aspect of the SPTR based mostly on the result of one dimensional calculations. At first, the radiation shielding capability of water is compared with those of other candidate materials used in the blanket and shield of fusion reactors. Based on the result of the comparison and other requirements such as tritium breeding, thermal mechanical design, repair and maintenance procedures, the material arrangements of the blanket and shield are determined. The result of the blanket neutronics calculations, the radiation shielding calculations for the superconducting magnets, shutdown dose calculations are given together with major penetration shielding considerations. (author)

The T1u x 8 hg Jahn-Teller system - an improved model for the C60-molecule

International Nuclear Information System (INIS)

Rough, S.M.; Dunn, J.L.; Bates, C.A.

1997-01-01

The ground state of C 60 - gives rise to a T 1u x 8 h g Jahn-Teller (JT) system. A proof is presented showing that the presence of eight active h g modes rather than one makes little difference to the mathematical complexity of this problem compared to the simpler single-mode variant. After showing that the T 1u x 8 h g Jahn-Teller system has the same electronic eigenstates as the T 1u x h g Jahn-Teller system, the inversion splitting and first-order reduction factors are derived. (orig.)

Experimental transport analysis code system in JT-60

International Nuclear Information System (INIS)

Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

1988-03-01

Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

Two dimensional neutral transport analysis in tokamak plasma

International Nuclear Information System (INIS)

Shimizu, Katsuhiro; Azumi, Masafumi

1987-02-01

Neutral particle influences the particle and energy balance, and play an important role on sputtering impurity and the charge exchange loss of neutral beam injection. In order to study neutral particle behaviour including the effects of asymmetric source and divertor configuration, the two dimensional neutral transport code has been developed using the Monte-Carlo techniques. This code includes the calculation of the H α radiation intensity based on the collisional-radiation model. The particle confinement time of the joule heated plasma in JT-60 tokamak is evaluated by comparing the calculated H α radiation intensity with the experimental data. The effect of the equilibrium on the neutral density profile in high-β plasma is also investigated. (author)

Integrated plasma control for high performance tokamaks

International Nuclear Information System (INIS)

Humphreys, D.A.; Deranian, R.D.; Ferron, J.R.; Johnson, R.D.; LaHaye, R.J.; Leuer, J.A.; Penaflor, B.G.; Walker, M.L.; Welander, A.S.; Jayakumar, R.J.; Makowski, M.A.; Khayrutdinov, R.R.

2005-01-01

Sustaining high performance in a tokamak requires controlling many equilibrium shape and profile characteristics simultaneously with high accuracy and reliability, while suppressing a variety of MHD instabilities. Integrated plasma control, the process of designing high-performance tokamak controllers based on validated system response models and confirming their performance in detailed simulations, provides a systematic method for achieving and ensuring good control performance. For present-day devices, this approach can greatly reduce the need for machine time traditionally dedicated to control optimization, and can allow determination of high-reliability controllers prior to ever producing the target equilibrium experimentally. A full set of tools needed for this approach has recently been completed and applied to present-day devices including DIII-D, NSTX and MAST. This approach has proven essential in the design of several next-generation devices including KSTAR, EAST, JT-60SC, and ITER. We describe the method, results of design and simulation tool development, and recent research producing novel approaches to equilibrium and MHD control in DIII-D. (author)

Detailed Analysis of the Transient Voltage in a JT-60SA PF Coil Circuit

International Nuclear Information System (INIS)

Yamauchi, K.; Shimada, K.; Terakado, T.; Matsukawa, M.; Coletti, R.; Lampasi, A.; Gaio, E.; Coletti, A.; Novello, L.

2013-01-01

A superconducting coil system is actually complicated by the distributed parameters, e.g. the distributed mutual inductance among turns and the distributed capacitance between adjacent conductors. In this paper, such a complicated system was modeled with a reasonably simplified circuit network with lumped parameters. Then, a detailed circuit analysis was conducted to evaluate the possible voltage transient in the coil circuit. As a result, an appropriate (minimum) snubber capacitance for the Switching Network Unit, which is a fast high voltage generation circuit in JT-60SA, was obtained. (fusion engineering)

Computer system for the beam line data processing at JT-60 prototype neutral beam injector

International Nuclear Information System (INIS)

Horiike, Hiroshi; Kawai, Mikito; Ohara, Yoshihiro

1987-08-01

The present report describes the hard and soft wares of the data acquisition computer system for the prototype neutral injector unit for JT-60. In order to operate the unit, more than hundreds of signals of the beam line components have to be measured. These are mainly differential thermometers for the coolant waters and thermocouples for the beam dump components but not include those for the cryo system. Since the unit operates in a series of pulses, the measurement should be conducted very quickly in order to ensure the simultaneity of large number of the measured data. The present system actualize fast data acquisition using a small computer of 128 kB and measuring instruments connected through the bus. The system is connected to the JAERI computer center since the data capacity is fairly large to completely process them by the small computer. Therefore the measured data can be transferred to the computer center to calculate there, and the results can be received. After the system was completed the computer quickly print out the power flow data, which needed much work to calculate with hands. This system was very useful. It enhanced the experiments at the unit and reduced the labor. It enables us to early demonstrate the rated operation of the unit and to accurately estimate such operation data of the JT-60 NBI as the injection power. (author)

Edge Localized Modes: resent experimental findings and related issues

International Nuclear Information System (INIS)

Kamiya, K.

2007-01-01

Edge Localized Mode (ELM) measurements in the tokamaks, including JT-60U, DIII-D, ASDEX-U and JET, are reviewed. An ELMy H-mode operation having Type-I ELMs is nominated as the reference inductive operational scenario for ITER (Q DT =10), which is normally observed for the best performing H-mode in many tokamaks,. However, the ELMs produce pulsed heat and particle fluxes that can lead to a rapid erosion of the divertor plate. It is estimated that the peak heat flux to the divertor would reduce the lifetime of the divertor to several hundred shots in ITER (e.g. an acceptable divertor lifetime could be realized only by an upper limit of ELM energy loss normalized by pedestal stored energy, ΔDW ELM /W ped ∼ 5-6%). Approaches to control the Type-I ELMs, such as '' Ergodization '' on DIII-D, '' Pace making by a shallow pellet injection '' on ASDEX-U, '' Vertical motion '' on TCV, have been successfully demonstrated in many tokamaks. On the other hand, finding alternative scenarios to Type-I ELMy H-mode operation are also a key area of research for current tokamaks. Specifically, '' Quiescent H-mode (QH-mode) '' on DIII-D, ASDEX-U and JT-60U, and '' Grassy ELMs '' on JT-60U demonstrated a high confinement (being comparable to that of Type-I ELMy H-mode plasmas at similar parameters) in the absence of large, ELM induced, transient heat/particle fluxes to the divertor targets. ELM dynamics measurements in the SOL at the midplane show large, rapid variations of the SOL parameters. Recent data from a fast resolved measurements, such as scanning probe, radial interferometer chord, BES and tangentially viewing fast-gated camera at the midplane, suggest a filamentary structure of the perturbation with fast radial propagation in later phases and parallel propagation of the ELM pulse at around the sound speed of pedestal ions. The results are qualitatively consistent with nonlinear ballooning theory, although a more quantitative physics understanding, including detailed

Variations of current profiles in tokamaks. Formation mechanism and confinement property of current-hole configuration

International Nuclear Information System (INIS)

Takizuka, Tomonori

2003-01-01

The formation mechanism of the current hole in tokamak plasmas is reviewed. Experimental results of JT-60U are shown. Increase of the off-central noninductive current is a key factor for the current-hole formation. The internal Transport Barrier (ITB), which generates large bootstrap current, plays an important role. The central current density in the hole stays nearly 0. The idea of a new equilibrium for a tokamak plasma with a current hole is introduced. This equilibrium configuration called Axisymmetric Tri-Magnetic-Islands (ATMI) equilibrium', has three islands along the R direction (a central-negative-current island and side-positive-current islands). The equilibrium is stable with the elongation coils when the current in the ATMI region is limited to a small amount. The confinement properties of a current-hole configuration with box-type ITB is described. A scaling of the core poloidal beta inside the ITB, β p,core , is given as ε f β p,core approx. = 1, which suggests the equilibrium limit (ε f : inverse aspect ratio at the ITB foot). Though the core stored energy is little dependent on the heating power, the estimated heat diffusivity in the ITB region moderately correlates with a neoclassical diffusivity. (author)

Radiation losses and global power balance of JT-60 plasmas

International Nuclear Information System (INIS)

Nishitani, T.; Itami, K.; Nagashima, K.; Tsuji, S.; Hosogane, N.; Yoshida, H.; Ando, T.; Kubo, H.; Takeuchi, H.

1990-01-01

The radiation losses and the global power balance for Ohmic and neutral beam heated plasmas have been investigated in different JT-60 configurations. Discharges with a TiC coated molybdenum wall and with a graphite wall, with limiter, outer and lower X-point configurations have been studied by bolometric measurements, thermocouples and an infrared TV camera. In neutral beam heated outer X-point discharges with a TiC coated molybdenum first wall, the radiation loss of the main plasma was very low (10% of the absorbed power). The radiation loss due to oxygen was dominant in this case. On the contrary, in discharges with TiC coated molybdenum limiters the radiation loss was very high (>60% of the absorbed power). In the discharges with a graphite wall the radiated power from the main plasma was 20-25% for both limiter and lower X-point configurations. In lower X-point discharges the main contributor to the radiation loss was oxygen, whereas in limiter discharges the loss due to carbon was equal to the loss due to oxygen. The radiation loss from the lower X-point divertor increased with increasing electron density of the main plasma. (author). 33 refs, 14 figs, 1 tab

Electrical conductivity in tokamaks and extended neoclassical theory

International Nuclear Information System (INIS)

Segre, S.E.; Zanza, V.

1992-01-01

The electrical conductivity measurements reported from various tokamaks (D-III, PLT, TEXT, ASDEX, JT-60, TEXTOR, JET, TFTR) and compared with the usual neoclassical theory are here also compared with the extended neoclassical theory where the electron-electron collision rate is anomalous while the electron-ion collision rate remains Coulombian. It is found that, out of the 14 experiments considered, three are consistent with both the neoclassical and the extended neoclassical theories, four are consistent only with the extended neoclassical theory, and four are consistent with the neoclassical theory and also, within the experimental errors, not inconsistent with the extended neoclassical theory; the remaining three experiments appear to be incompatible with both theories. It is concluded that the extended neoclassical theory is in better agreement with conductivity experiments than the conventional neoclassical theory and, indeed, the extended theory is a serious candidate for explaining tokamak behaviour, since it accommodates naturally an anomalous electron thermal transport, which the conventional neoclassical theory is unable to do. (author). 31 refs, 1 fig

Annual report of the Fusion Research Center for the period of April 1, 1982 to March 31, 1983

International Nuclear Information System (INIS)

1983-11-01

Research and development activities of the Fusion Research Center (Department of Thermonuclear Fusion Research and Department of Large Tokamak Development) from April 1982 to March 1983 are described. The JFT-2 tokamak was shutdown after 10 years operation. Operation test of a new device JFT-2M was near completion. In the joint JAERI-USDOE experiment on Doublet-III a record value of beta, 4.6 %, was achieved. Major efforts in theory and computation was on high beta tokamak stability, second stability regions being found for low m internal modes. The JT-60 program progressed as scheduled, installation of the tokamak machine being initiated in February 1983. A 100 kV test was completed of prototype unit for JT-60 NBI. In the development of a high power klystron for JT-60 LH heating, a test fabricated tube generated 1 MW, 10 s RF pulses. Development of TiC coatings for JT-60 first wall was successfully concluded. In the superconducting magnet technology, the Japanese coil for IEA Large Coil Task was installed in a test facility at ORNL after successful performance test at Naka site. A 10 T experiment of a Nb 3 Sn coil with 60 cm inner bore was made. Construction of the Tritium Process Laboratory was started in February 1983. Design studies of the Fusion Experimental Reactor and INTOR were continued. (author)

Mechanical strength evaluation of the welded bellows for the ports of the JT-60 vacuum vessel

International Nuclear Information System (INIS)

Takatso, H.; Shimizu, M.; Yamamoto, M.

1983-01-01

Mechanical strength of the welded bellows for the ports of the JT-60 vacuum vessel was evaluated, laying the emphasis on the fatigue strength under the torsional electromagnetic force. The welded bellows were designed to be loaded with the forced deflection due to the relative displacement between the vacuum vessel and the external fixed point, the atmospheric pressure and the forced torsional angle due to the electromagnetic force. Stresses caused by the former two were estimated following the formulae proposed by the Kellogg Company. On the other hand, two formulae were established to estimate the stress caused by the last, after examining experimentally the behavior of the welded bellows under the torsional load; one is the shearing stress evaluation formula and the other is the axial bending stress evaluation formula. It was found that the welded bellows can easily buckle under the torsional load and the former formula corresponds to the case of non-buckling and the latter to the case of buckling. The present mechanical strength evaluation method was applied to the three kinds of the welded bellows to be used in the ports of the JT-60 vacuum vessel (neutral beam injection ports, vacuum pumping ports and the adjustable limiter ports) and it was confirmed that they have sufficient strength in the range of the design load conditions

Feedback control of plasma configuration in JT-60

International Nuclear Information System (INIS)

Ninomiya, Hiromasa; Kikuchi, Mitsuru; Yoshino, Ryuji; Hosogane, Nobuyuki; Kimura, Toyoaki; Kurihara, Kenichi; Takahashi, Minoru; Hayashi, Kazuo.

1986-08-01

Plasma current, plasma position (center of the outermost magnetic surface), decay index n index and width of the divertor throat are feedback controlled by using 5 kinds of poloidal field coils in JT-60. 5 control commands are calculated in a feedback control computer in each 1 msec. These feedback control functions are checked in ohmically heated plasma. The control characteristics of the plasma are well understood by the simplified control analysis and are consistent with the precise matrix transfer function analysis in the frequency domain and the simulation analysis which include the effects of eddy currents, delay time elements and mutual interactions between controllers. The usefulness of these analyses is experimentally confirmed. Each controlled variable is well feedback controlled to the command and the experimentally realized equilibrium configuration is checked by the well calibrated magnetic probes. Fast boundary identification code is used for the identification of the equilibrium and results are consistent with the precalculated plasma equilibria. By using this feedback control system of the plasma configuration and the equilibrium identification method, we have obtained the stable limiter and divertor configuration. The maximum parameters obtained during OH(I) experimental period are plasma current I p = 1.8 MA, the effective safety factor q eff e = 5.7 x 10 19 m -3 (Murakami parameter of 4.5) and the pulse length of 5 ∼ 10 sec. (author)

Fabrication and tests of EF conductors for JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Kizu, Kaname, E-mail: kizu.kaname@jaea.go.jp [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Kashiwa, Yoshitoshi; Murakami, Haruyuki [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Obana, Tetsuhiro; Takahata, Kazuya [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Tsuchiya, Katsuhiko; Yoshida, Kiyoshi [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Hamaguchi, Shinji [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Matsui, Kunihiro [Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Nakamura, Kazuya; Takao, Tomoaki [Sophia University, Tokyo 102-8554 (Japan); Yanagi, Nagato; Imagawa, Shinsaku; Mito, Toshiyuki [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan)

2011-10-15

The conductors for plasma equilibrium field (EF) coils of JT-60SA are NbTi cable-in-conduit (CIC) conductor with stainless steel 316L jacket. The production of superconductors for actual EF coils started from February 2010. Nine superconductors with 444 m in length were produced up to July 2010. More than 300 welding of jackets were performed. Six nonconformities were found by inspections as go gauge, visual inspection and X-ray test. In order to shorten the manufacturing time schedule, helium leak test was conducted at once after connecting the long length jacket not just after the welding. The maximum force to pull the cable into jacket was about 7.6 kN on average. The mass flow rates of 9 conductors showed almost same values indicating that there are no blockages in the conductors. The measured current sharing temperature agreed with the expectation values from strand performance indicating that no degradation was caused by production process. The coupling time constants of conductors ranged from 80 to 90 ms which are much smaller than the design value of 200 ms.

Development of the plasma movie database system in JT-60

International Nuclear Information System (INIS)

Sueoka, Michiharu; Kawamata, Yoichi; Kurihara, Kenichi; Seki, Akiyuki

2008-03-01

A plasma movie is generally expected as one of the most efficient methods to know what plasma discharge has been conducted in the experiment. The JT-60 plasma movie is composed of video camera picture looking at a plasma, computer graphics (CG) picture, and magnetic probe signal as a sound channel. In order to use this movie efficiently, we have developed a new system having the following functions: (a) To store a plasma movie in the movie database system automatically combined with the plasma shape CG and the sound according to a discharge sequence. (b) To make a plasma movie is available (downloadable) for experiment data analyses at the Web-site. Especially, this system aimed at minimizing the development cost, and it tried to develop the real-time plasma shape visualization system (RVS) without any operating system (OS) customized for real-time use. As a result, this system succeeded in working under Windows XP. This report deals with the technical details of the plasma movie database system and the real-time plasma shape visualization system. (author)

Proceedings of the workshop of three large tokamak cooperation on energy confinement scaling under intensive auxiliary heating, May 18 ∼ 20, 1992, Naka

International Nuclear Information System (INIS)

1992-09-01

The workshop of three large tokamak cooperation W22 on 'Energy confinement scaling under intensive auxiliary heating' was held 18-20 May, 1992, at Naka Fusion Research Establishment. This proceedings compiles 14 synopses of contributions (5 from JET, 4 from JT-60, 3 from TFTR, and 1 each from DIII-D JFT-2M) and the summary of the workshop. Topic sections are ; (i) L-mode confinement and scaling, (ii) Confinement at high β P regimes, Supershots, High poloidal beta enhanced confinement mode etc., (iii) Confinement at various H-mode regimes and scaling (including the VH-mode), (iv) Characteristic time scales for present tokamak regimes, and (v) Theoretical comparison with experimental data. (author)

Experience with high heat flux components in large tokamaks

International Nuclear Information System (INIS)

Chappuis, P.; Dietz, K.J.; Ulrickson, M.

1991-01-01

The large present day tokamaks. i.e.JET, TFTR, JT-60, DIII-D and Tore Supra are machines capable of sustaining plasma currents of several million amperes. Pulse durations range from a few seconds up to a minute. These large machines have been in operation for several years and there exists wide experience with materials for plasma facing components. Bare and coated metals, bare and coated graphites and beryllium were used for walls, limiters and divertors. High heat flux components are mainly radiation cooled, but stationary cooling for long pulse duration is also employed. This paper summarizes the experience gained in the large machines with respect to material selection, component design, problem areas, and plasma performance. 2 tabs., 26 figs., 50 refs

Structural design of vacuum bulkheads in piping penetration for the cryostat base of JT-60SA

International Nuclear Information System (INIS)

Nakamura, Shigetoshi; Shibama, Yusuke K.; Masaki, Kei

2016-11-01

This study examined the structure of the boundary box that is capable of installing the cryostat base of JT-60SA in a narrow space. Since other devices stand close in the neighborhood, it was designed to fit within a limited space to avoid interference. Spatial limitation and generated stress caused by each load were used as design conditions. From the calculation results of the generated stress with respect to each load, the maximum stress is generated by the displacement of the pipeline associated with the displacement of the vacuum container at the time of earthquake and 200degC baking, so bellows were designed to absorb the displacement of the piping. It was confirmed through 3-D finite element analysis that this generated stress is less than the allowable stress and there is no problem in structural integrity. This paper explained the composition of major equipment of JT-60SA and the structure of cryostat base. In the structural analysis of the boundary box, consideration was given to the pressure difference during vacuum closure or abnormal events, temperature distribution, pipe displacement associated with the deformation of vacuum vessel, and seismic load. As a result of finite element analysis, it was confirmed that the displacement amount and temperature distribution during plasma operation and baking were within the allowable range. In addition, the maximum stress during cryostat helium leak was also within the allowable range. (A.O.)

Joule loss on a Faraday shield of JT-60 ICRF test antenna

International Nuclear Information System (INIS)

Fujii, Tsuneyuki; Saigusa, Mikio; Ikeda, Yoshitaka; Kimura, Haruyuki; Hirashima, Teruhisa; Uehara, Munenori.

1988-01-01

Joule loss on a Faraday shield of JT-60 ICRF test antenna with a conductive casing is investigated at the frequency range of 120 MHz. The magnetic field radiated from the antenna is measured by three-dimensionally scanning an rf probe both inside and outside the antenna casing. The magnetic field perpendicular to the Faraday shield, B x , is found to be the largest component near the Faraday shield. It consequently gives the major part of the joule loss on the Faraday shield. The temperature distribution of the Faraday shield due to joule loss is measured directly with a thermocamera. It is confirmed that the area of the high temperature rise is consistent with the peak positions of the B x field. Faraday shield resistance which is estimated from power measurements agrees with the theoretical value. (author)

Effect of magnetic and density fluctuations on the propagation of lower hybrid waves in tokamaks

Science.gov (United States)

Vahala, George; Vahala, Linda; Bonoli, Paul T.

1992-12-01

Lower hybrid waves have been used extensively for plasma heating, current drive, and ramp-up as well as sawteeth stabilization. The wave kinetic equation for lower hybrid wave propagation is extended to include the effects of both magnetic and density fluctuations. This integral equation is then solved by Monte Carlo procedures for a toroidal plasma. It is shown that even for magnetic/density fluctuation levels on the order of 10-4, there are significant magnetic fluctuation effects on the wave power deposition into the plasma. This effect is quite pronounced if the magnetic fluctuation spectrum is peaked within the plasma. For Alcator-C-Mod [I. H. Hutchinson and the Alcator Group, Proceedings of the IEEE 13th Symposium on Fusion Engineering (IEEE, New York, 1990), Cat. No. 89CH 2820-9, p. 13] parameters, it seems possible to be able to infer information on internal magnetic fluctuations from hard x-ray data—especially since the effects of fluctuations on electron power density can explain the hard x-ray data from the JT-60 tokamak [H. Kishimoto and JT-60 Team, in Plasma Physics and Controlled Fusion (International Atomic Energy Agency, Vienna, 1989), Vol. I, p. 67].

Analyses of the impact of connections’ layout on the coil transient voltage at the Quench Protection Circuit intervention in JT-60SA

International Nuclear Information System (INIS)

Maistrello, Alberto; Gaio, Elena; Novello, Luca; Matsukawa, Makoto; Yamauchi, Kunihito

2015-01-01

The transient overvoltages associated to the interruption of high direct currents with high current derivative, at the base of the operation of a Quench Protection System for Superconducting Coils, have been studied, with particular reference to the JT-60SA project, which adopt edge technology solutions for current interruption: a Hybrid mechanical-static Circuit Breaker as main circuit breaker in series with a PyroBreaker as backup protection. The paper reports in particular on the analyses of the intervention of the backup circuit breaker in the final circuital conditions, considering the actual power connections that will be implemented on Site. The key elements which influence the peak value of the voltage and the relation existing among the different stray impedances of the circuit are identified, thus giving general guidelines for the design of the layout of the power connections. The specific case of JT-60SA is considered, but general criteria can be derived.

General Tokamak Circuit Simulation Program-GTCSP

International Nuclear Information System (INIS)

Matsukawa, Makoto; Miura, Yushi; Aoyagi, Tetsuo.

1997-05-01

General Tokamak Circuit Simulation Program (GTCSP) was originally developed for the design work of JT-60 Power Supply System in JAERI. Therefore the prepared models (components) to be analyzed are generator, thyristor converter and coils. This is one of the unique points of GTCSP in comparison with other conventional electric circuit analysis program, because they make a circuit from the small devices such as resister, coil, condenser, transistor and so on. However, GTCSP is also clearly conventional because it is possible to construct an electric circuit freely with the prepared components. Moreover, a similar function could be realized by addition a new component to GTCSP. This report is assumed to be used as an User Manual of the GTCSP, not only to present the development and the analytical functions. Then some useful examples are described, and how to get graphic outputs are also mentioned. (author)

Stationary Flowing Liquid Lithium (SFLiLi) systems for tokamaks

Science.gov (United States)

Zakharov, Leonid; Gentile, Charles; Roquemore, Lane

2013-10-01

The present approach to magnetic fusion which relies on high recycling plasma-wall interaction has exhausted itself at the level of TFTR, JET, JT-60 devices with no realistic path to the burning plasma. Instead, magnetic fusion needs a return to its original idea of insulation of the plasma from the wall, which was the dominant approach in the 1970s and upon implementations has a clear path to the DEMO device with PDT ~= 100 MW and Qelectric > 1 . The SFLiLi systems of this talk is the technology tool for implementation of the guiding idea of magnetic fusion. It utilizes the unique properties of flowing LiLi to pump plasma particles and, thus, insulate plasma from the walls. The necessary flow rate, ~= 1 g3/s, is very small, thus, making the use of lithium practical and consistent with safety requirements. The talk describes how chemical activity of LiLi, which is the major technology challenge of using LiLi in tokamaks, is addressed by SFLiLi systems at the level of already performed (HT-7) experiment, and in ongoing implementations for a prototype of SFLiLi for tokamak divertors and the mid-plane limiter for EAST tokamak (to be tested in the next experimental campaign). This work is supported by US DoE contract No. DE-AC02-09-CH11466.

Annual report of the Fusion Research Center for the period of April 1, 1983 to March 31, 1984

International Nuclear Information System (INIS)

1985-03-01

Research and development activities of the Fusion Research Center (Department of Thermonuclear Fusion Research and Department of Large Tokamak Development) from April 1983 to March 1984 are described. Installation and commissioning of the new tokamak JFT-2M had been completed. The 2nd ICRF heating experiment and LH current drive experiment were started. In the field of plasma theory, the scaling law of the critical beta in a tokamak was obtained and the ICRF heating was analyzed in detail. The first phase of the cooperation of Doublet III will be finished in Sept. 1984. The JT-60 program progressed as scheduled. Installation of the tokamak machine, initiated in Feb. 1983, will be finished in Sept. 1984. The tests of power supply and control system on site and the fabrication of the neutral beam injectors in factory proceeded successfully. Performance tests of prototype injector unit for JT-60 NBI progressed as scheduled. A new advanced source plasma generator was developed to provide a high proton ratio exceeding 90%. Klystrons for JT-60 LH heating achieved the output power of 1 MW for 10 sec. Performance tests of titanium evaporators for JT-60 were completed. The Japanese coil for IEA Large Coil Task was installed in a test facility at ORNL and the partial cool-down was carried out. Construction of the Tritium Process Laboratory was completed. Design studies of the Fusion Experimental Reactor (FER) and INTOR proceeded. (author)

Tokamak plasma power balance calculation code (TPC code) outline and operation manual

International Nuclear Information System (INIS)

Fujieda, Hirobumi; Murakami, Yoshiki; Sugihara, Masayoshi.

1992-11-01

This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)

Role of low order rational q-values in the ITB events in JT-60U reverse shear plasmas

International Nuclear Information System (INIS)

Neudatchin, S.V.; Takizuka, T.; Hayashi, N.; Isayama, A.; Shirai, H.; Fujita, T.; Kamada, Y.; Koide, Y.; Suzuki, T.

2004-01-01

Non-local confinement bifurcations inside and around internal transport barriers (ITBs) with a ms timescale (ITB events) have previously been found in JT-60U reverse shear (RS) and high-β p plasmas. ITB events are observed as the simultaneous rise and decay of T e in two zones. They are created by an abrupt non-local reduction (or increase) of heat flux inside 30-40% of the minor radius. Under sufficient neutral beam power P nbi (above ∼8 MW for the 1.2-1.5 MA/3.8 T pulses described below), ITB events were previously detected at various q min values. However, the role of q min equal to 3.5, 3, 2.5, 2 is not obvious for ITB formation. In this paper, we focus on new features of ITB evolution near low-order-rational values of q min . The formation of a stronger ITB and its further splitting into two radially separated ITBs is described. These ITBs are located in both positive and negative shear zones of a plasma with L-mode edge. The similarity of space-time evolution of T e and T i at sufficient power is highlighted (even when the variation is significant and complicated in space and time). Within error-bars, ITB splitting occurs as q min passes through 2.5. The similarity of space-time evolution of T e and T i suggests a similarity in the qualitative behaviour of electron and ion heat diffusivities in time and space. The temporal formation of an ITB in the zone with small positive shear, while q min passes through 3 (after periodical improvements and degradations via ITB events with 8 ms period) in H-mode, with P nbi = 8 MW, is described. At lower powers, ITB events are observed only near rational values of q min . In weak RS shots with P nbi = 4 MW, transport is reduced via ITB events during 0.08 s at q min = 3.5, and repetitive short-term phases of reduced transport are observed as q min passes through 3. The behaviour of T i looks different. The difference in T e and T i evolution, which was detected regularly under low power, probably indicates a decoupling

Engineering study of the neutral beam and rf heating systems for DIII-D, MFTF-B, JET, JT-60 and TFTR

International Nuclear Information System (INIS)

Lindquist, W.B.; Staten, S.H.

1987-01-01

An engineering study was performed on the rf and neutral beam heating systems implemented for DIII-D, MFTF-B, JET, JT-60 and TFTR. Areas covered include: methodology used to implement the systems, technology, cost, schedule, performance, problems encountered and lessons learned. Systems are compared and contrasted in the areas studied. Summary statements were made on common problems and lessons learned. 3 refs., 6 tabs

Pulsed-laser ablation of co-deposits on JT-60 graphite tile

International Nuclear Information System (INIS)

Sakawa, Youichi; Watanabe, Daisuke; Shibahara, Takahiro; Sugiyama, Kazuyoshi; Tanabe, Tetsuo

2007-01-01

Pulsed laser ablation of the co-deposits on a JT-60 open-divertor tile using the fourth harmonic of a 20 ps-Nd: YAG laser has been investigated. With increasing the laser intensity, three regions, non-ablation region (NAR), weak-ablation region (WAR), and strong-ablation region (SAR) were distinguished. Transition from NAR to WAR and WAR to SAR occurred at the threshold laser intensity for laser ablation and that for strong ionization of carbon atoms, respectively. The ablation accompanied desorption of H 2 and C 2 H 2 , with minor contribution of other hydrocarbons, while production of H 2 O was small. In NAR and WAR the number of the hydrogen desorbed by the laser irradiation was less than that of hydrogen retained in the ablated volume, while in SAR it was much larger, owing to thermal desorption of hydrogen gas from the region surrounding the ablated volume. For the ablative removal of hydrogen isotopes, SAR is more desirable because of higher removal efficiency and less production of hydrocarbons

Pulsed-laser ablation of co-deposits on JT-60 graphite tile

Energy Technology Data Exchange (ETDEWEB)

Sakawa, Youichi [Institute of Laser Engineering, Osaka University, Yamadaoka, Suita, Osaka 565-0871 (Japan)]. E-mail: sakawa-y@ile.osaka-u.ac.jp; Watanabe, Daisuke [Graduate School of Engineering, Nagoya University, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Shibahara, Takahiro [Graduate School of Engineering, Nagoya University, Chikusa-ku, Nagoya, Aichi 464-8603 (Japan); Sugiyama, Kazuyoshi [Interdisciplinary School of Engineering Science, Kyushu University, Fukuoka, Fukuoka 812-8581 (Japan); Tanabe, Tetsuo [Interdisciplinary School of Engineering Science, Kyushu University, Fukuoka, Fukuoka 812-8581 (Japan)

2007-08-01

Pulsed laser ablation of the co-deposits on a JT-60 open-divertor tile using the fourth harmonic of a 20 ps-Nd: YAG laser has been investigated. With increasing the laser intensity, three regions, non-ablation region (NAR), weak-ablation region (WAR), and strong-ablation region (SAR) were distinguished. Transition from NAR to WAR and WAR to SAR occurred at the threshold laser intensity for laser ablation and that for strong ionization of carbon atoms, respectively. The ablation accompanied desorption of H{sub 2} and C{sub 2}H{sub 2}, with minor contribution of other hydrocarbons, while production of H{sub 2}O was small. In NAR and WAR the number of the hydrogen desorbed by the laser irradiation was less than that of hydrogen retained in the ablated volume, while in SAR it was much larger, owing to thermal desorption of hydrogen gas from the region surrounding the ablated volume. For the ablative removal of hydrogen isotopes, SAR is more desirable because of higher removal efficiency and less production of hydrocarbons.

Detailed design studies at CEA for JT-60SA TF coils

International Nuclear Information System (INIS)

Decool, P.; Marechal, J.L.; Portafaix, C.; Lacroix, B.; Gros, G.; Verger, J.M.

2011-01-01

Following a first conceptual design activity in which the general design of the JT-60SA TF system was defined and frozen in agreement with all the participants in the project (CEA, ENEA, F4E), a second phase had to be launched to deal with the detailed design. In this paper, we present the work performed at CEA on the TF coil design during this second phase. Part of this work, concerns the determination of conductor hydraulic performances during operation as well as in factory. The thermohydraulic of the conductor was also assessed to confirm the need of helium inlets and a specific design was developed and qualified to be compatible with the available hydraulic performance of the cryoplant. The mechanical behavior is still to be assessed and qualified. Last but not least, the inner electrical joints of the coil have been modified with respect to the original twin-box design developed by CEA for the ITER coils in order to simplify the fabrication process. A dedicated qualification program for their manufacture is ongoing.

Application of PLC to dynamic control system for liquid He cryogenic pumping facility on JT-60U NBI system

International Nuclear Information System (INIS)

Honda, A.; Okano, F.; Ooshima, K.; Akino, N.; Kikuchi, K.; Tanai, Y.; Takenouchi, T.; Numazawa, S.; Ikeda, Y.

2008-01-01

The control system of the cryogenic facility in the JT-60 NBI system has been replaced by employing the PLC (Programmable Logic Controller) and SCADA (Supervisory Control And Data Acquisition) system. The original control system was constructed about 20 years ago by specifying the DCS (Distributed Control System) computer to deal with ∼400 feedback loops. Recently, troubles on this control system have increased due to its age-induced deterioration. To maintain the high reliability of the cryogenic facility, a new control system has been planned with the PLC and SCADA systems. Their attractive features include high market availability and cost-effectiveness, however, the use of PLC for such a large facility with ∼400 feedback loops has not been established because of insufficient processing capability of the early PLC. Meanwhile, the recent progress in the PLC enables to use the FBD (function block diagram) programming language for 500 function blocks. By optimizing the function blocks and connecting them in the FBD language, the feedback loops have been successfully replaced from DCS to PLC without a software developer. Moreover, an oscillation of the liquid He level, which often occurs during the cooldown mode of the cryopumps, can be automatically stabilized by easily adding a new process program in the PLC. At present, the new control system has worked well

Development of hard-seal gate valve and fast shutter for JT-60 neutral beam injectors

International Nuclear Information System (INIS)

Kuribayashi, S.; Minami, M.; Matsuoka, T.; Takeshita, K.; Morita, H.; Kuriyama, M.; Matsuda, S.; Shirakata, H.

1983-01-01

A 600 mm hard-seal valve and a fast shutter for the JT-60 Neutral Beam Injector were developed. The 600 mm hard-seal gate valve was fabricated and tested for 500 cycles at various temperatures of up to 250 0 C. In consequence, requirements of the endurance and vacuum tightness were satisfied. Major components of the fast shutter, i.e., swing action bellows and a high-speed pneumatic cylinder, were tested for 30,000 cycles, and their reliability was confirmed. Then the fast shutter was fabricated and tested. The test result indicated that the fast shutter fully satisfied the requirements of the molecular gas flow conductance and opening/closing speed. (author)

Origin of the various beta dependences of ELMy H-mode confinement properties

International Nuclear Information System (INIS)

Takizuka, T; Urano, H; Takenaga, H; Oyama, N

2006-01-01

Dependence of the energy confinement in ELMy H-mode tokamak on the beta has been investigated for a long time, but a common conclusion has not been obtained so far. Recent non-dimensional transport experiments in JT-60U demonstrated clearly the beta degradation. A database for JT-60U ELMy H-mode confinement was assembled. Analysis of this database is carried out, and the strong beta degradation consistent with the above experiments is confirmed. Two subsets of ASDEX Upgrade and JET data in the ITPA H-mode confinement database are analysed to find the origin of the various beta dependences. The shaping of the plasma cross section, as well as the fuelling condition, affects the confinement performance. The beta dependence is not identical for different devices and conditions. The shaping effect, as well as the fuelling effect, is a possible candidate for causing the variation of beta dependence

JT-60SA TF magnet industrial manufacturing preparation and qualifications

International Nuclear Information System (INIS)

Decool, P.; Cloez, H.; Gros, G.; Marechal, J.L.; Torre, A.; Verger, J.M.; Nusbaum, M.; Billotte, G.; Crepel, B.; Bourquard, A.; Davis, S.; Phillips, G.

2014-01-01

The general design of the JT-60SA toroidal field system was defined in agreement with all the participants in the project (CEA, ENEA, F4E), the detailed design was issued by the Voluntary Contributors. For the French part including the procurement of 9 of the 18 TF winding packs and their integration in the casings, an industrial contract was signed mid-2011 with Alstom (France). After agreement on manufacturing drawings and QA documentation, the manufacturing process was defined giving the guidelines for the workshop organization and the definition of the required tooling. The critical manufacturing points were identified in the process and, regarding technical requirements, have led to the definition of a set of qualification mockups. They are related to helium inlets, conductor winding and insulation, local conductor bending, electrical joint and terminal areas for the winding pack (WP), as well as winding embedding, case welding, and impregnations for WP integration in the casing. The fabrication processes have been improved and shall be qualified thanks to the manufacture and testing of 12 corresponding mockups. The successful achievement of several key mock-ups gives confidence in the feasibility of the manufacture, and their completion will give the green light to the industrial coils manufacture. (authors)

Development of remote pipe welding tool for divertor cassettes in JT-60SA

Energy Technology Data Exchange (ETDEWEB)

Hayashi, Takao, E-mail: hayashi.takao@jaea.go.jp [Fusion Research and Development Directorate, Japan Atomic Energy Agency, Naka (Japan); Sakurai, Shinji; Sakasai, Akira; Shibanuma, Kiyoshi [Fusion Research and Development Directorate, Japan Atomic Energy Agency, Naka (Japan); Kono, Wataru; Ohnawa, Toshio; Matsukage, Takeshi [Toshiba Corporation, Yokohama, Kanagawa (Japan)

2015-12-15

Highlights: • Remote pipe welding tool accessing from inside of the pipe has been newly developed. • Cooling pipe with a jut on the edge expands the acceptable welding gap up to 0.5 mm. • Positioning accuracy of the laser beam is realized to be less than ±0.1 mm. • We have achieved robust welding for an angular misalignment of 0.5°. - Abstract: Remote pipe welding tool accessing from inside of the pipe has been newly developed for JT-60SA. Remote handling (RH) system is necessary for the maintenance and repair of the divertor cassette in JT-60SA. Because the space around the cooling pipe connected with the divertor cassette is very limited, the cooling pipe is to be remotely cut and welded from inside for the maintenance. A laser welding method was employed to perform circumferential welding by rotating the focusing mirror inside the pipe. However, the grooves of connection pipes are not precisely aligned for welding. The most critical issue is therefore to develop a reliable welding tool for pipe connection without a defect such as undercut weld due to a gap caused by angular and axial misalignments of grooves. In addition, an angular misalignment between two pipes due to inclination of pipe has to be taken into account for the positioning of the laser beam during welding. In this paper, the followings are proposed to solve the above issues: (1) Cooling pipe connected with the divertor is machined to have a jut on the edge so as to expand the acceptable welding gap up to 0.5 mm by filling the gap with welded jut. (2) Positioning accuracy of the laser beam for reliable welding is realized to be less than ±0.1 mm along the circumferential target for welding by a position control mechanism provided in the tool even in the case of up to angular misalignment of 0.5° between connection pipes. Based on the above proposals, we have achieved robust welding for a large gap up to 0.5 mm as well as the maximum angular misalignment of 0.5° between connection pipes

Library system for a one dimensional tokamak transport code: (LIBJT60), 1

International Nuclear Information System (INIS)

Hirayama, Toshio

1982-12-01

A library system is developed to control and manage huge programs in terms of FORTRAN source. It is applied to widely used one dimensional tokamak transport codes (LIBJT60), which have been developed in the Division of Large Tokamak Development. The structure of data and program in the transport code turn out to be flexible enough to respond to various demands and this gigantic code frame work can be decomposed into groups of a compact code with a specific function. Some editing support tools for programming and debugging are also developed to save programming work. By applying this library system, users can obtain a code whose functions can be efficiently developed. (author)

Recent RF Experiments and Application of RF Waves to Real-Time Control of Safety Factor Profile in JT-60U

International Nuclear Information System (INIS)

Suzuki, T.; Isayama, A.; Ide, S.; Fujita, T.; Oikawa, T.; Sakata, S.; Sueoka, M.; Hosoyama, H.; Seki, M.

2005-01-01

Two topics of applications of RF waves to current profile control in JT-60U are presented; application of lower-hybrid (LH) waves to safety factor profile control and electron cyclotron (EC) waves to neo-classical tearing mode (NTM) control. A real-time control system of safety factor (q) profile was developed. This system, for the first time, enables 1) real time evaluation of q profile using local magnetic pitch angle measurement by motional Stark effect (MSE) diagnostic and 2) control of current drive (CD) location (ρCD) by controlling the parallel refractive index N parallel of LH waves through control of phase difference (Δφ) of LH waves between multi-junction launcher modules. The method for real-time q profile evaluation was newly developed, without time-consuming reconstruction of equilibrium, so that the method requires less computational time. Safety factor profile by the real-time calculation agrees well with that by equilibrium reconstruction with MSE. The control system controls ρCD through Δφ in such a way to decrease the largest residual between the real-time evaluated q profile q(r) and its reference profile qref(r). The real-time control system was applied to a positive shear plasma (q(0)∼1). The reference q profile was set to monotonic positive shear profile having qref(0)=1.3. The real-time q profile approached to the qref(r) during application of real-time control, and was sustained for 3s, which was limited by the duration of the injected LH power. Temporal evolution of current profile was consistent with relaxation of inductive electric field induced by theoretical LH driven current. An m/n=3/2 NTM that appeared at βN∼3 was completely stabilized by ECCD applied to a fully-developed NTM. Precise ECCD at NTM island was essential for the stabilization. ECCD that was applied to resonant rational surface (q=3/2) before an NTM onset suppressed appearance of NTM. In order to keep NTM intensity below a level, ECCD before the mode onset was

Review of JT-60 experimental results from January to October, 1989

International Nuclear Information System (INIS)

1990-03-01

Emphases in recent JT-60 experiments are placed on 1) improvements in plasma confinement with profile control and 2) steady state operation study. Both limiter and lower X-point divertor configurations were employed. The operating gas was hydrogen and, in some cases, helium. Improvements in confinement were demonstrated with pellet injection, LH current drive, high-Ti mode operation or ICRF. Current profile controllability with energetic electrons has been improved by the new LH launcher. The H-mode was achieved in limiter discharges with LH current drive for the first time. Nearly steady-state ELM-free H-mode with durations up to 3.3. sec was established without significant impurity accumulation. High-Ti and high-βp discharges were obtained in high field and low plasma current. Major experimental issues for the steady-state operation research were non-inductive current drive and He-ash exhaust. The neoclassical bootstrap current was confirmed in the wide range of βp. 30 keV helium NB was injected into NB heated lower X-point discharges, which produced centrally peaked birth profile of α-particles. A simple extension of the present result is promising for the helium exhaust in future device. (J.P.N.)

Increase of the positive ion source power in JT-60 NBI

International Nuclear Information System (INIS)

Kawai, Mikito; Akino, Noboru; Ebisawa, Noboru

1998-09-01

Neutral Beam Injection (NBI) heating experiment in JT-60 started in 1986, and the rated injection power of 20MW at 75keV with hydrogen was achieved after several month operation. In 1991, the ion sources and power supply had been upgraded for a higher beam energy up to 120keV with deuterium, following which the ion source operation re-started aiming for an injection power of 40MW at 110keV. In the operation, the beam acceleration voltage was tried to increase by modifying the ion source structure against the break-down which occurred frequently in the ion source. The beam acceleration was, however, unstable in a beam energy range of more than 105keV because of voltage-holding deterioration in the accelerator. Therefore we changed the strategy to increase the injection power: i.e. we tried to increase the beam current with keeping the beam energy. The structure of the source has been modified to be operated in a high current regime. As a result, the deuterium neutral beam injection of 40MW at 91-96keV was achieved in July 1996. (author)

Using plasma waves to create in tokamaks the necessary quasi-stationary conditions for controlled fusion

International Nuclear Information System (INIS)

Moreau, D.

1993-04-01

It is studied, on the one hand, how using hybrid waves with frequency near from lower hybrid frequency in fusion plasma. Works about coupling waves in plasma (chap.I), their propagation and response of the plasma to the absorption of the waves (chap.II). This method is the most effective until today. Because of limits, it has been investigated, on the other hand, fast magnetosonic wave to control current density in the centre of the discharge in a reactor or a very hot plasma. Theoretical study (chap.III) and experimental results (chap.IV) are presented. Experiments are in progress or planned in following tokamaks: D3-D (USA), JET (Europe), TORE SUPRA (France), JT-60 (Japan). figs. refs. tabs

A Conceptual Design Study for the Error Field Correction Coil Power Supply in JT-60SA

International Nuclear Information System (INIS)

Matsukawa, M.; Shimada, K.; Yamauchi, K.; Gaio, E.; Ferro, A.; Novello, L.

2013-01-01

This paper describes a conceptual design study for the circuit configuration of the Error Field Correction Coil (EFCC) power supply (PS) to maximize the expected performance with reasonable cost in JT-60SA. The EFCC consists of eighteen sector coils installed inside the vacuum vessel, six in the toroidal direction and three in the poloidal direction, each one rated for 30 kA-turn. As a result, star point connection is proposed for each group of six EFCC coils installed cyclically in the toroidal direction for decoupling with poloidal field coils. In addition, a six phase inverter which is capable of controlling each phase current was chosen as PS topology to ensure higher flexibility of operation with reasonable cost.

Improvement of confinement characteristics of tokamak plasma by controlling plasma-wall interactions

International Nuclear Information System (INIS)

Sengoku, Seio

1985-08-01

Relation between plasma-wall interactions and confinement characteristics of a tokamak plasma with respect to both impurity and fuel particle controls is discussed. Following results are obtained from impurity control studies: (1) Ion sputtering is the dominant mechanism of impurity release in a steady state tokamak discharge. (2) By applying carbon coating on entire first wall of DIVA tokamak, dominant radiative region is concentrated more in boundary plasma resulting a hot peripheral plasma with cold boundary plasma. (3) A physical model of divertor functions about impurity control is empilically obtained. By a computer simulation based on above model with respect to divertor functions for JT-60 tokamak, it is found that the allowable electron temperature of the divertor plasma is not restricted by a condition that the impurity release due to ion sputtering does not increase continuously. (4) Dense and cold divertor plasma accompanied with strong remote radiative cooling was diagnosed along the magnetic field line in the simple poloidal divertor of DOUBLET III tokamak. Strong particle recycling region is found to be localized near the divertor plate. by and from particle control studies: (1) The INTOR scaling on energy confinement time is applicable to high density region when a core plasma is fueled directly by solid deuterium pellet injection in DOUBLET III tokamak. (2) As remarkably demonstrated by direct fueling with pellet injection, energy confinement characteristics can be improved at high density range by decreasing particle deposition at peripheral plasma in order to reduce plasma-wall interaction. (3) If the particle deposition at boundary layer is necessarily reduced, the electron temperature at the boundary or divertor region increases due to decrease of the particle recycling and the electron density there. (J.P.N.)

Neutron measurements as fusion plasma diagnostics

International Nuclear Information System (INIS)

Nishitani, Takeo; Hoek, M.

1993-01-01

Neutron measurements play important roles as the diagnostics of many aspects of the plasma in large tokamak devices such as JT-60U and JET. In the d-d discharges of JT-60U, the most important application of the neutron measurement is the investigation of the fusion performance using fission chambers. The ion velocity distribution function, and the triton slowing down are investigated by the neutron spectrometer and the 14 MeV neutron detector, respectively. TANSY is a combined proton-recoil and neutron time-of flight spectrometer for 14 MeV neutrons to be used during the d-t phase at JET. The detection principle is based on the measurements of the flight time of a scattered initial neutron and the energy of a corresponding recoil proton. The scattering medium is a polyethylene foil. The resolution and efficiency, using a thin foil (0.95 mg/cm 2 ), is 155 keV and 1.4x10 -5 cm 2 , respectively. (author)

Energy confinement and transport of H-mode plasmas in tokamak

International Nuclear Information System (INIS)

Urano, Hajime

2005-02-01

A characteristic feature of the high-confinement (H-mode) regime is the formation of a transport barrier near the plasma edge, where steepening of the density and temperature gradients is observed. The H-mode is expected to be a standard operation mode in a next-step fusion experimental reactor, called ITER-the International Thermonuclear Experimental Reactor. However, energy confinement in the H-mode has been observed to degrade with increasing density. This is a critical constraint for the operation domain in the ITER. Investigation of the main cause of confinement degradation is an urgent issue in the ITER Physics Research and Development Activity. A key element for solving this problem is investigation of the energy confinement and transport properties of H-mode plasmas. However, the influence of the plasma boundary characterized by the transport barrier in H-modes on the energy transport of the plasma core has not been examined sufficiently in tokamak research. The aim of this study is therefore to investigate the energy confinement properties of H-modes in a variety of density, plasma shape, seed impurity concentration, and conductive heat flux in the plasma core using the experimental results obtained in the JT-60U tokamak of Japan Atomic Energy Research Institute. Comparison of the H-mode confinement properties with those of other tokamaks using an international multi-machine database for extrapolation to the next step device was also one of the main subjects in this study. Density dependence of the energy confinement properties has been examined systematically by separating the thermal stored energy into the H-mode pedestal component determined by MHD stability called the Edge Localized Modes (ELMs) and the core component governed by gyro-Bohm-like transport. It has been found that the pedestal pressure imposed by the destabilization of ELM activities led to a reduction in the pedestal temperature with increasing density. The core temperature for each

Power system for tokamak fusion experiments. Motor generator with flywheel effect

International Nuclear Information System (INIS)

Miyachi, Kengo

1997-01-01

JT-60 requires an enormous electric power pulse about 1,300 MVA periodically for its plasma initiation, containment and heating. JT-60 could not receive all electric power from a commercial line for plasma experiment except about 160 MVA because the 275 kV commercial line has some limitations. Therefore JT-60 needs huge electric power sources. The power supply system of JT-60 has 3 motor generators (MG). The total capacity of MG is 1,115 MVA that consists of a toroidal MG (TMG), poloidal MG (PMG) and Heating power supply MG (HMG), and each MG has a huge flywheel effect. For example, TMG has a 4.02 GJ energy yield that consists of 6 disk flywheel. The total weight of flywheel of TMG is 650 ton. This report describes the structure, operating system, and maintenance history of 3 types of MG. (author)

Annual report of the Naka Fusion Research Establishment for the period of April 1, 1990 to March 31, 1991

International Nuclear Information System (INIS)

1991-10-01

R and D activities of the Naka Fusion Research Establishment, JAERI, are reported for the period from April 1, 1990 to March 31, 1991. Since the shutdown of JT-60 in November 1989, the reconstruction work of the JT-60 device was continued until the end of March 1991. In the JT-60 Upgrade, the poloidal field coils and vacuum vessel were renewed and the plasma current was planned to increase up to 6 MA with lower single null divertor. The divertor plates were designed to be toroidally continuous and to use high-heat-conduction C/C composite materials. Another objective of JT-60U is to facilitate tokamak experiments with deuterium as the working gas. In the JFT-2M program, a system for divertor bias experiments was brought into operation and initial experiments were started to study its effects on plasma discharges. Effects of ergodic magnetic limiter on H-modes were examined and stationary H-modes were obtained under the control of ergodic magnetic limiter currents. The DIII-D program was highlighted by the attainment of 11% beta with a double null divertor plasma. As for the fusion engineering research, development activities of the ceramic turbo-viscous pump and the surface insulation techniques for the tokamak in-vessel components are remarked in the vacuum technology area. In the high heat flux experiments with the JAERI Electron Beam Irradiation Stand (JEBIS), carbon-based materials and refractory metals were tested to evaluate surface erosion at plasma disruptions. The ITER Conceptual Design Activities, which began in April 1988 under the auspices of the IAEA, were successfully completed in December 1990. A lot of contributions to the program were made by JAERI people to support the design and R and D activities and to prepare a plan for the forthcoming Engineering Design Activities. (J.P.N.)

Annual report of the Fusion Research Center for the period of April 1, 1984 to March 31, 1985

International Nuclear Information System (INIS)

1986-01-01

Research and development activities of the Fusion Research Center (Department of Large Tokamak Development and Department of Thermonuclear Fusion Research) from April 1984 to March 1985 are described. The JT-60 program progressed as scheduled. Commissioning of the JT-60 tokamak was completed by the end of the period under review. In parallel with installation and test of the tokamak machine, installation of basic diagnostic instruments and examination of the procedure for experiment had been made to meet the first phase Joule heating experiment. (The first plasma discharge was recorded on April 8, 1985). Construction of auxiliary heating systems had continued. A medium-sized tokamak, JFT-2M, had been operated for high-power ICRF heating and ECH assisted LH current drive experiments. Installation of a power supply for plasma shaping in JFT-2M was completed. In the field of plasma theory, detailed analysis had been made on nonlinear kink/tearing modes in a plasma with free boundary and also on ICRF heating. Development of a high-voltage, high-current He ion source for JT-60 plasma diagnostics had proceeded successfully, and tests of JT-60 LH and ICRF luanchers as well. Surface erosion of a new ceramics, SiC with BeO addition by proton bombardment was studied. In IEA's Large Coil Task, three coil test was made at ORNL. A 11 T experiment of TMC-1, a large-bore Nb 3 Sn coil was completed. Commissioning tests of tritium handling facilities had proceeded in the Tritium Process Laboratory. Design studies of the Fusion Experimental Reactor (FER) and INTOR had been advanced. (author)

Mathematical modeling plasma transport in tokamaks

Energy Technology Data Exchange (ETDEWEB)

Quiang, Ji [Univ. of Illinois, Urbana-Champaign, IL (United States)

1997-01-01

In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10²⁰/m³ with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%.

Mathematical modeling plasma transport in tokamaks

International Nuclear Information System (INIS)

Quiang, Ji

1995-01-01

In this work, the author applied a systematic calibration, validation and application procedure based on the methodology of mathematical modeling to international thermonuclear experimental reactor (ITER) ignition studies. The multi-mode plasma transport model used here includes a linear combination of drift wave branch and ballooning branch instabilities with two a priori uncertain constants to account for anomalous plasma transport in tokamaks. A Bayesian parameter estimation method is used including experimental calibration error/model offsets and error bar rescaling factors to determine the two uncertain constants in the transport model with quantitative confidence level estimates for the calibrated parameters, which gives two saturation levels of instabilities. This method is first tested using a gyroBohm multi-mode transport model with a pair of DIII-D discharge experimental data, and then applied to calibrating a nominal multi-mode transport model against a broad database using twelve discharges from seven different tokamaks. The calibrated transport model is then validated on five discharges from JT-60 with no adjustable constants. The results are in a good agreement with experimental data. Finally, the resulting class of multi-mode tokamak plasma transport models is applied to the transport analysis of the ignition probability in a next generation machine, ITER. A reference simulation of basic ITER engineering design activity (EDA) parameters shows that a self-sustained thermonuclear burn with 1.5 GW output power can be achieved provided that impurity control makes radiative losses sufficiently small at an average plasma density of 1.2 X 10 20 /m 3 with 50 MW auxiliary heating. The ignition probability of ITER for the EDA parameters, can be formally as high as 99.9% in the present context. The same probability for concept design activity (CDA) parameters of ITER, which has smaller size and lower current, is only 62.6%

Review of preliminary additional heating experiments in JT-60 (Aug. - Nov., 1986)

International Nuclear Information System (INIS)

1987-03-01

This is a prompt report on preliminary additional heating experiments in JT-60 from August to November in 1986. Neutral beam heating power was raised up to 20 MW in about a month. Plasma stored energy is about 2 MJ and energy confinement time is 0.1 ∼ 0.12 sec with the maximum heating power. The energy confinement time shows L-mode like deterioration with power, while it has little dependence on electron density. The maximum ion temperature of ∼ 7 keV and electron temperature of 4.5 keV were obtained at relatively low electron density (n-bar e = 2 - 3 x 10 19 m -3 ). Lower hybrid wave could efficiently drive plasma current up to 1.7 MA with 1.2 MW LH power. The current drive efficiency is 1 ∼ 1.7 in ohmically heated plasmas and 2 ∼ 2.8 in NB heated plasmas. Futhermore the energy confinement was improved when neutral beam was injected into entirely current driven discharges of 1 MA by LH in contrast to inductively driven target plasmas. Similar improvement in energy confinement was observed during combined heating with NB and ion cyclotron wave. (author)

Development of the computer system for the JT-60 negative-ion based NBI

International Nuclear Information System (INIS)