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Hoshino, Katsumichi; Yamamoto, Takumi; Tamai, Hiroshi; Oasa, Kazumi; Kawashima, Hisato; Miura, Yukitoshi; Ogawa, Toshihide; Shoji, Teruaki*; Shibata, Takatoshi; Kikuchi, Kazuo; et al.
Fusion Science and Technology, 49(2), p.139 - 167, 2006/02
Times Cited Count：2 Percentile：17.18(Nuclear Science & Technology)The main results obtained by the various heating and current drive systems, external coil system and divertor bias system are reviewed from the viewpoint of the advanced active control of the tokamak plasma. Also, the features of each system are described. The contribution of the JFT-2M in these areas are summarized.
plasma operationTamai, Hiroshi; Akiba, Masato; Azechi, Hiroshi*; Fujita, Takaaki; Hamamatsu, Kiyotaka; Hashizume, Hidetoshi*; Hayashi, Nobuhiko; Horiike, Hiroshi*; Hosogane, Nobuyuki; Ichimura, Makoto*; et al.
Nuclear Fusion, 45(12), p.1676 - 1683, 2005/12
Times Cited Count：15 Percentile：45.53(Physics, Fluids & Plasmas)Design studies are shown on the National Centralized Tokamak facility. The machine design is carried out to investigate the capability for the flexibility in aspect ratio and shape controllability for the demonstration of the high-beta steady state operation with nation-wide collaboration, in parallel with ITER towards DEMO. Two designs are proposed and assessed with respect to the physics requirements such as confinement, stability, current drive, divertor, and energetic particle confinement. The operation range in the aspect ratio and the plasma shape is widely enhanced in consistent with the sufficient divertor pumping. Evaluations of the plasma performance towards the determination of machine design are presented.
Seki, Masami; Moriyama, Shinichi; Shinozaki, Shinichi; Hasegawa, Koichi; Hiranai, Shinichi; Yokokura, Kenji; Shimono, Mitsugu; Terakado, Masayuki; Fujii, Tsuneyuki
Fusion Engineering and Design, 74(1-4), p.273 - 277, 2005/11
Times Cited Count：3 Percentile：24.22(Nuclear Science & Technology)no abstracts in English
Ikeda, Yoshitaka; Oikawa, Toshihiro; Ide, Shunsuke
Purazuma, Kaku Yugo Gakkai-Shi, 81(10), p.773 - 778, 2005/10
In steady-state tokamak fusion reactors, an efficient external current drive and a large fraction of the bootstrap current are required for non-inductive operation at low circulating power. NBI is a powerful and reliable actuator for current drive and heating. A negative ion-based NBI (N-NBI) with a high beam energy more than 350 keV has been installed in the JT-60U tokamak in order to study the NBI current drive and heating in an ITER relevant regime. This paper presents recent progress of N-NBI experiments and its system in JT-60U towards steady-state operation for ITER and tokamak fusion reactors.
Suzuki, Takahiro; Isayama, Akihiko; Ide, Shunsuke; Fujita, Takaaki; Oikawa, Toshihiro; Sakata, Shinya; Sueoka, Michiharu; Hosoyama, Hiroki*; Seki, Masami; JT-60 Team
AIP Conference Proceedings 787, p.279 - 286, 2005/09
A real-time control system of safety factor (q) profile was developed in JT-60. 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 (rhoCD) by adjusting the parallel refractive index 
of lower-hybrid (LH) waves through the change of phase difference (dphi) of LH waves between multi-junction launcher modules. The method for q profile evaluation was newly developed, without time-consuming reconstruction of equilibrium, so that the method requires less computational time. The system evaluates q profile within every 10ms, which is much faster than current relaxation time, typically order of 1s. Safety factor profile by the real-time calculation agreed well with that by equilibrium reconstruction with MSE. From temporal evolution of q (or current) profile, we evaluate CD location in real-time, too. The control system controls rhoCD through (or directly dphi) in such a way to minimize difference between the real-time evaluated q profile and its reference profile. The real-time control system was applied to positive shear plasmas (
), having plasma current of 0.6MA, toroidal field of 2.3T, and electron density of 
. In order to keep good coupling of LH waves to the plasma, gap between the launcher and the plasma surface was controlled to about 0.1m. The reference q profile was set to q(0)=1.3. The real-time q profile approached to the reference after application of real-time control; the controlled q profile was sustained for 3s, which was limited by injected LH power. RF experiments in JT-60U, such as stabilization of neo-classical tearing modes, plasma startup experiments, etc., are also presented.
Kishimoto, Hiroshi; Ishida, Shinichi; Kikuchi, Mitsuru; Ninomiya, Hiromasa
Nuclear Fusion, 45(8), p.986 - 1023, 2005/08
Times Cited Count：40 Percentile：28.81(Physics, Fluids & Plasmas)The Japanese large tokamak JT-60 has been focusing its research emphases to develop a high performance plasma, namely high confinement, high temperature and high density, and to sustain it non-inductively for a long time with possible minimization of external power input. The first demonstration of high bootstrap current discharges in a high-poloidal-beta mode (high-p) and the concept development of a steady-state tokamak reactor SSTR based on this experimental achievement initiated the so-called "advanced tokamak research". The first observation of internal transport barriers in the JT-60 high-p mode was followed by the world-wide explorations of reversed shear discharges associated with internal transport barriers. The advanced tokamak research is now the major trend of the current tokamak development. A new concept of compact ITER was developed and proposed in the context of this advanced tokamak approach pursued on JT-60.
Fujita, Takaaki; Suzuki, Takahiro; Oikawa, Toshihiro; Isayama, Akihiko; Hatae, Takaki; Naito, Osamu; Sakamoto, Yoshiteru; Hayashi, Nobuhiko; Hamamatsu, Kiyotaka; Ide, Shunsuke; et al.
Physical Review Letters, 95(7), p.075001_1 - 075001_4, 2005/08
Times Cited Count：14 Percentile：60.97(Physics, Multidisciplinary)We found that no current can be driven in a central region of high-temperature, magnetically-confined, axi-symmetric torus plasma once the central current density becomes nearly zero ("current hole"), in spite of high electric conductivity. The current clamp was observed against current drive by a toroidal electric field and a radio-frequency wave in experiments on the JT-60U tokamak. This is a new, stiff, self-organized structure of magnetic field in an axi-symmetric torus plasma.
Ikeda, Yoshitaka; Kubo, Shin*
Purazuma, Kaku Yugo Gakkai-Shi, 81(3), p.160 - 166, 2005/03
The local profile controls of electron temperature and current in fusion-oriented devices using the electron cyclotron (EC) wave is reviewed. Recent progress of the EC heating system that enabled those controls is briefly described. The specific feature of EC wave heating is local and high power density heating properties. Current drive and electron temperature profile control experiments using EC wave performed in order to improve and investigate plasma confinement properties are discussed.
Inoue, Takashi; Sakamoto, Keishi
Nihon Genshiryoku Gakkai-Shi, 47(2), p.120 - 127, 2005/02
no abstracts in English
Miura, Yukitoshi; Hoshino, Katsumichi; Kusama, Yoshinori
Purazuma, Kaku Yugo Gakkai-Shi, 80(8), p.653 - 661, 2004/08
A series of experimental program on the JAERI Fusion Torus-2M (JFT-2M) was completed in March, 2004. In the experimental operation for 21 years since the first plasma on April 27, 1983, many significant results leading the fusion energy research and plasma physics have been produced in researches on high confinement mode (H-mode), heating and current drive, advanced plasma control, compatibility of low activation ferritic steel with improved confinement mode, etc. Among these results, some important results are presented.
Hayashi, Nobuhiko; Isayama, Akihiko; Nagasaki, Kazunobu*; Ozeki, Takahisa
Purazuma, Kaku Yugo Gakkai-Shi, 80(7), p.605 - 613, 2004/07
Neoclassical tearing mode stabilization by an electron cyclotron wave current drive (ECCD) has been studied by using the numerical model on the basis on the modified Rutherford eq. coupled with the 1.5D transport code and the EC code. Numerical model almost reproduces the JT-60U experimental results. Undetermined parameters in the modified Rutherford equation are estimated from JT-60U experiments. Sensitivity of stabilization to the EC current location is investigated. Low EC current and peaked EC current profile mitigates the sensitivity, while higher EC current and the peaked EC current profile moves the rational surface more largely through the background current modification by ECCD and intensify the sensitivity. High EC current and broad EC current profile also mitigates the sensitivity. EC current necessary for the full stabilization is studied. The necessary EC current much depends on the parameters of bootstrap current and ECCD terms in the modified Rutherford equation. Necessary ECCD power on ITER is evaluated on the basis of parameters estimated from JT-60U experiments.
Tamai, Hiroshi; Ishida, Shinichi; Kurita, Genichi; Shirai, Hiroshi; Tsuchiya, Katsuhiko; Sakurai, Shinji; Matsukawa, Makoto; Sakasai, Akira
Fusion Science and Technology, 45(4), p.521 - 528, 2004/06
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The 1.5D time-dependent transport analysis has been carried out to investigate steady state operation scenarios with a central current hole by off-axis current drive schemes consistent with a high bootstrap current fraction for a large superconducting tokamak JT-60SC. A steady state operation scenario with HH
=1.4 and 
=3.7 has been obtained at I
=1.5 MA, B
=2 T and q
=5 where non-inductive currents are developed during the discharge to form a current hole with beam driven currents by tangential off-axis beams in combination with bootstrap currents by additional on-axis perpendicular beams. The bootstrap fraction increases up to nearly 75% of the plasma current and the current hole region is enlarged up to about 30% of the minor radius at 35 s from the discharge initiation. The current hole is confirmed to be sustained afterward for a long duration of 60 s. The stability analysis shows that the beta limit with the conducting wall can be about =4.5, which is substantially above the no wall ideal MHD limit.
Hayashi, Nobuhiko; Ozeki, Takahisa; Hamamatsu, Kiyotaka; Takizuka, Tomonori
Nuclear Fusion, 44(4), p.477 - 487, 2004/04
Times Cited Count：35 Percentile：72.74(Physics, Fluids & Plasmas)NTM stabilization by ECCD has been numerically studied in order to evaluate the necessary EC power for the stabilization of NTM on ITER. The time evolution of an island width of NTM is calculated by the modified Rutherford equation. The modification of background current profile by the EC current is calculated by the current diffusion equation and the variation of the tearing stability index 
' is taken into account. When the EC power is higher than a threshold value, NTM with any island width can be fully stabilized. The dependence of the threshold power on parameters in the modified Rutherford equation is examined. The threshold power much depends on the parameters of the bootstrap current term in the modified Rutherford equation. The injected EC current decreases the value of ' through the background current modification, which results in the reduction of the threshold power. The effects of the peakedness of EC current profile and the EC power modulation on the threshold power are investigated. When the width of the EC current profile becomes half, the threshold power is reduced to half or less. The EC power modulation is inessential for the threshold power reduction if the EC current profile can be peaked. Considering the maximum value of the threshold power in the range of the parameters, the EC power of about 25 MW is found to be sufficient for the simultaneous stabilization of both the 
=3/2 and 2/1 mode NTM on ITER.
Fujita, Takaaki; JT-60 Team
Nuclear Fusion, 43(12), p.1527 - 1539, 2003/12
Times Cited Count：32 Percentile：68.11(Physics, Fluids & Plasmas)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 betap H-mode plasma with full non-inductive current drive has been obtained at 1.8 MA and the fusion triple product reached 3.1E20m
keVs. NTM suppression with EC was accomplished using a real-time feedback control system and improvement in betaN 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. A new type of AE mode has been proposed and found to explain the observed frequency chirp quite well. Ar exhaust with EC heating was obtained in a high betap mode plasma.
JT-60 Team
JAERI-Review 2003-029, 197 Pages, 2003/11
JAERI-Review-2003-029.pdf:13.06MB
no abstracts in English
Fujii, Tsuneyuki; Kasugai, Atsushi; JT-60 Team
Proceedings of 20th IEEE/NPSS Symposium on Fusion Engineering (SOFE 2003), p.222 - 227, 2003/10
The key factors to realize highly integrated performance plasma performances are control of profiles of current, pressure, rotation and so on. Therefore, several types of heating and current drive systems, ECH, LH, ICH, negative and positive ion based NBI systems, have been introduced into JT-60U. The ECH system with output power 4MW at 110 GHz has been developed using four 1 MW gyrotrons. The gyrotron has achieved 1MW-5sec output of the designed value by suppressing the parasitic oscillation with SiC RF absorber built-in. This gyrotron may get higher output power because it can adjust the anode voltage and extend the range of oscillation parameters. The NBI system has been attained 5.8 MW at 400 kV of beam energy with the negative ion based one and 28 MW with the positive ion based one. In a development work of the negative ion based NBI system, a detailed study on beamlet steering for multi-beam focusing was done, including of the space charge effects among beamlets. Then, 2.6 MW for 10s has been achieved.
Inoue, Takashi; Hanada, Masaya; Iga, Takashi*; Imai, Tsuyoshi; Kashiwagi, Mieko; Kawai, Mikito; Morishita, Takatoshi; Taniguchi, Masaki; Umeda, Naotaka; Watanabe, Kazuhiro; et al.
Fusion Engineering and Design, 66-68, p.597 - 602, 2003/09
Times Cited Count：21 Percentile：78.49(Nuclear Science & Technology)The neutral beam (NB) injection has been one of the most promising methods for plasma heating and current drive in tokamak fusion devices. JAERI has developed high energy electrostatic accelerators for the NB systems in JT-60U and ITER. Recent progress on this R&D are as follows: 1) In the JT-60U NB system, some of the beams has been deflected due to distorted electric field in the accelerator, resulting in an excess heat load on the NB port. By correcting the electric field, a continuous injection of H
beam was succeeded for 10 s with the NB power of 2.6 MW at 355 keV. 2) To increase the beam energy, a metal structure called stress ring was designed. The ring reduces electric field concentration at the triple junction point (interface between metal and dielectric insulator inside vacuum). Initial test of the accelerators with the stress rings has shown higher voltage hold off performance in both accelerators for JT-60U and ITER R&D than that without rings.
Kaneko, Osamu*; Yamamoto, Takumi; Akiba, Masato; Hanada, Masaya; Ikeda, Katsunori*; Inoue, Takashi; Nagaoka, Kenichi*; Oka, Yoshihide*; Osakabe, Masaki*; Takeiri, Yasuhiko*; et al.
Fusion Science and Technology, 44(2), p.503 - 507, 2003/09
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)High energy negative-ion-based neutral beam injection (N-NBI) is expected as an efficient and reliable tool of heating and current driving for reactor plasmas such as ITER. A world wide activity on developing technology of negative ion production and beam formation started in 1980's and the great progress has been achieved up to now. In particular, Japan has two large projects that planned adopting N-NBI for real plasma experiments; the JT-60U tokamak and the LHD heliotron, which further motivated the R&D activity. These R&D programs were carried out at JAERI and NIFS separately in Japan, and both were successfully done. The first beam injection experiment was made on the JT-60U in 1996, followed by the LHD in 1998. They were the first experiments on heating plasma by high energy beam in tokamaks and in stellerators, and the obtained results were very promising.
Ishii, Yasutomo; Azumi, Masafumi; Kishimoto, Yasuaki
Physics of Plasmas, 10(9), p.3512 - 3520, 2003/09
Times Cited Count：12 Percentile：37.11(Physics, Fluids & Plasmas)no abstracts in English
JT-60 Team
JAERI-Review 2002-022, 149 Pages, 2002/11
JAERI-Review-2002-022.pdf:10.21MB
no abstracts in English
