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Electron cyclotron current drive (ECCD) has been numerically simulated in magnetic islands caused by neo-classical tearing modes. The electron drift orbits are tracked with the Coulomb collisions and the quasi-linear diffusion by EC waves, which are simulated by the Monte-Carlo method. The EC resonance region is assumed to be located around the O-point and localized in the toroidal direction. The driven current is well confined in the helical flux tube which includes the EC resonance region. The driven current channel looks like a 'snake' in real space. As the results of the ECCD with 10 MW in a plasma with ne = 3 × 1019 m−3 and Te = 10 keV, the current drive efficiency is about 1.6 times higher than that of an axi-symmetric plasma with no magnetic island. The driven current profile tends to peak around the O-point with increasing EC wave power.

The mission of the National Centralized Tokamak (NCT) is to establish high-beta steady-state plasma operation for DEMO and to contribute to ITER. Plasma performance with high-beta conditions is discussed. It is found that near break-even class plasma ( Q DT eq ≥ 0.7 ) with high beta ( β N ≥ 3.5) is achievable with 25 MW of NB power at I P = 4.0 MA and an aspect ratio of 2.6. High beta with steady-state full current drive is possible with 13 MW of NB power at I P = 1.5–1.7 MA. In the engineering design activity of NCT, the strength of the support structure and strength of material for the conductor is confirmed for the superconducting coils, and the successful progress of R&D of boron-loaded resin for neutron shielding is discussed.

Design studies are shown on the National Centralized Tokamak facility, formerly called JT-60SC. The machine design is carried out to investigate the capability for flexibility in aspect ratio and shape controllability for the demonstration of the high-β 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 consistent with sufficient divertor pumping. Evaluations of the plasma performance towards the determination of the machine design are presented.

In the JT-60 tokamak at JAERI, an environment for remote participation is planning to be developed by concentrating ideas of experts of nuclear fusion research from research organizations and universities distributed throughout the country. We are constructing a hierarchical collaborative remote research system, which consists of remote experiment, remote analysis and remote diagnostic. In remote collaboration, it is important to share information, atmosphere and presence between the participants as well as to maintain the security of the system. To implement a collaborative environment, a video conferencing system, a video streaming system and remote analysis system called “VizAnalysis” have been developed for the communication. To assist the remote participation for the fusion research, the document management system based on the web browser called “VizSquare” has been developed. In JT-60, security and authentication methods utilizing a computer network and a new communication tool have been developed, and they are applicable to the remote participation of ITER.

The stabilization of the neoclassical tearing mode (NTM) by an electron cyclotron current drive (ECCD) has been numerically studied for the parameters of ITER in order to evaluate the necessary EC power for the stabilization of NTMs in ITER. The time evolution of an island width of NTM is calculated by the modified Rutherford equation. The modification of the background current density profile by the EC current located on the rational surface is calculated by the current diffusion equation and the variation of the tearing stability index Δ' is taken into account. When the EC power is higher than a threshold value, NTMs with any island width can be fully stabilized. The dependence of the threshold power on the parameters in the modified Rutherford equation is examined. The threshold power depends significantly on the parameters of the bootstrap current term in the modified Rutherford equation. The injected EC current decreases the value of Δ' through the background current modification, which results in the reduction of the threshold power. The effects of the peaked nature of the EC current density profile and the EC power modulation on the threshold power are investigated. When the width of the EC current density profile becomes half, the threshold power is reduced to half or less. The EC power modulation is found to be inessential for the threshold power reduction if the EC current density profile can be peaked. As regards the maximum value of the threshold power in the range of parameters, an EC power of about 25 MW is found to be sufficient for the simultaneous stabilization of both the m/n = 3/2 and 2/1 NTMs in ITER.

Recent progress in the development of electron cyclotron heating⁄current drive systems in regard to power and pulse duration have allowed the extension of two normalized parameters (toroidal electric field Eφ and wave power density p). In the extended regime, the conventional theory requires consideration of a distorted electron distribution function. The first validation of the ECCD theory in the extended regime having large Eφ and p is presented. Linear calculation has a tendency to overestimate the EC driven current, as normalized parameters for Eφ and p increase. While the EC driven current IEC obtained by a linearized Fokker-Planck calculation (1.1 MA) did not agree with the measured EC driven current (0.74 ± 0.06MA), non-linear calculation of the Fokker-Planck equation considering the effect of Eφ (0.76 MA) shows close agreement with the experimental result. Calculations show that the decreasing effect of Eφ on IEC was stronger than the increasing effect of p on IEC in the experiment.

The electron cyclotron range of frequency (ECRF) system was designed and operated on the JT-60U to locally heat and control plasmas. The frequency of 110 GHz was adopted to inject the fundamental O-mode from the low field side with an oblique injection angle. The system is composed of four 1 MW-level gyrotrons, four transmission lines, and two antennae. The gyrotron is featured by a collector potential depression (CPD) and a gaussian beam output through a diamond window. The CPD enables JAERI to drive the gyrotron under the condition of the main DC voltage of 60 kV without a thyristor regulation. The gaussian mode from the gyrotron is effectively transformed to HE{sub 11} mode in the 31.75 mm diameter corrugated waveguide. About 75% of the output power of the gyrotrons can be injected into plasmas through the waveguides about 60 m in length. There are two antennae to control the deposition position of the EC wave during a plasma discharge. One is connected with three RF lines to steer the EC beams in the poloidal direction. The other is to control the EC beam in the toroidal and poloidal directions by two steerable mirrors.On the operation in 2000, the power ofmore » 1.5 to 1.6 MW for 3 s was successfully injected into plasmas using three gyrotrons. Local profile control was demonstrated by using the antennae. This capability was devoted to improve the plasma performance such as high T{sub e} production more than 15 keV and suppression of the MHD activities. In 2001, the fourth gyrotron, whose structure was improved for long pulse operation, has been installed for a total injection power of {approx}3 MW.« less

An electron cyclotron range of frequency (ECRF) program has been initiated to study the local heating and current drive in JT-60U. A frequency of 110 GHz was adopted to couple the fundamental O-mode from the low-field side with an oblique toroidal injection angle for the current drive. Experiments were performed at an injection power of ~1.5 MW by using three gyrotrons, each of which has generated the output power up to ~0.8 MW for 3 seconds. A strongly peakedTeprofile was observed and the central electron temperature increased up to ~15 keV when the O-mode was absorbed on the axis. The local electron heating clarified the significant difference in the heat pulse propagation between in the plasmas with internal transport barrier (ITB) and without. The driven current estimated by the Motional Stark Effect (MSE) diagnostic showed that the electron cyclotron (EC) waves drove the plasma current up to ~0.2 MA for an injected power of ~1.3 MW at the local electron temperature and density ofTe~6 keV,ne~0.7×1019 m-3. The measured driven current near the axis was consistent with the theoretical prediction using a Fokker-Planck code. In the case of co-electron cyclotron current drive (ECCD), the sawtooth activity in neutral beam (NB) heated plasma was completely suppressed for 1.5 s with the deposition at the inversion radius, while the sawtooth was enhanced for counter-ECCD at the same deposition condition.

Localized driven current profile by electron cyclotron (EC) waves is measured on the JT-60U tokamak, considering transient inductive current. The inductive current is derived from temporal evolution of internal magnetic field, which is evaluated by reconstructions of MHD equilibria with a spline function of total current and with motional Stark effect polarimetry. The analysis clearly shows that the EC-driven current is in a thin layer of about 10% of the minor radius of the plasma. The determined EC-driven current agrees with the one calculated by ray tracing and Fokker-Planck codes. The analysis shows the change in driven current profile when the deposition location of EC waves is changed. The current drive efficiency is about 5×1018 A W-1 m-2 at an electron temperature of 7 keV.

Quasi-steady-state high βp H mode discharges performed by suppressing neoclassical tearing modes (NTMs) are described. Two operational scenarios have been developed for long sustainment of the high βp H mode discharge: NTM suppression by profile optimization, and NTM stabilization by local electron cyclotron current drive (ECCD)/electron cyclotron heating (ECH) at the magnetic island. Through optimization of pressure and safety factor profiles, a high βp H mode plasma with H89PL = 2.8, HHy,2 = 1.4, βp ≈ 2.0 and βN ≈ 2.5 has been sustained for 1.3 s at small values of collisionality νe* and ion Larmor radius ρi* without destabilizing the NTMs. Characteristics of the NTMs destabilized in the region with central safety factor above unity are investigated. The relation between the beta value at the mode onset βNon and that at the mode disappearance βNoff can be described as βNoff/βNon = 0.05-0.4, which shows the existence of hysteresis. The value of βN/ρi* at the onset of an m/n = 3/2 NTM has a collisionality dependence, which is empirically given by βN/ρi* ∝ νe*0.36. However, the profile effects such as the relative shapes of pressure and safety factor profiles are equally important. The onset condition seems to be affected by the strength of the pressure gradient at the mode rational surface. Stabilization of the NTM by local ECCD/ECH at the magnetic island has been attempted. A 3/2 NTM has been completely stabilized by EC wave injection of 1.6 MW.

Numerical analysis of the island width of the neoclassical tearing mode, which is induced by the helical bootstrap current along the island, have been carried out to clarify effects of profiles on stability and to obtain conditions for stabilization by electron cyclotron current drive (ECCD). When the neoclassical tearing mode emerges, the saturated island grows up to some 10% of the minor radius at the beta value well below the ideal kink limit. The saturated island width of m/n mode can be small when the magnetic shear is large and the pressure gradient is low at the corresponding rational surface. Therefore, in the broad pressure and high li (internal inductance) plasma, the neoclassical tearing mode is more stable when qmin is slightly above one or two. In the broad pressure and low li plasma, the dependence of the island width on q0 is weak. A localized additional current at the O-point of the island by the ECCD significantly stabilizes the mode. The localization of the external current is effective, however, the stabilization also works well when the island width is smaller the radial width of the local current. The required current for the stabilization does not depend on the profiles very much.

The controllability of current profile driven by the fundamental resonance of the O-mode wave is examined for the parameters of the Intermediate Aspect ratio Machine option of the International Thermonuclear Experimental Reactor (ITER). The driven current is numerically evaluated by using the full relativistic adjoint method which we have extended from the non-relativistic one. By changing both the toroidal and poloidal injection angles, the optimum direction to drive a current with maximum current density at the aimed position is obtained. The position of wave launching affects the optimum injection angles and the driven current profile. The dependence of driven current profile on beam divergence is also examined. When the beam is injected from the equatorial plane, the maximum current density is reduced and the width of current profile becomes wide in the outer region of r≳0.5a. The localized current profile is kept well within the region of r≳0.4a by injection from a location away from the equatorial plane.

The 110 GHz 1 MW electron cyclotron range of frequency (ECRF) system was designed and constructed on JT-60U to locally heat and control the plasmas. The gyrotron has a diamond window to transmit RF power with Gaussian mode, which is easily transformed to HE11 mode for the transmission line of the corrugated waveguide. The second diamond window is installed at the inlet of the antenna for a vacuum seal between the transmission line and the JT-60U tokamak. The total length of the transmission line from the gyrotron to the antenna is about 60 m including nine meter bends, The antenna has a focusing mirror and a flat steerable one to focus and to control the RF beam angle mainly in the poloidal direction. In the initial operation, the power of PEC∼0.75 MW for 2 s was successfully injected into plasma when the gyrotron generated the power up to 1 MW. The total transmission efficiency from the gyrotron to the plasma was about 75%. A controllability of local electron heating with the deposition width of =15 cm was well demonstrated by using the steerable mirror. A large downshift in the deposition position was observed at the high Te plasma. Strong central electron heating was obtained from 2.2 to 6.6 keV for PEC∼0.75 MW, 0.3 s at the optimized polarization. An effective electron heating was also obtained up to ∼10 keV during EC injection for ∼1.6 s in the high βp H-mode plasma produced by NBI.

An electron cyclotron current drive by the fundamental O-mode is numerically analysed in the off-axis region of a reactor-grade tokamak. By scanning the launching direction in both the toroidal and poloidal directions, the optimum direction of beam injection is obtained to drive a current with a maximum value of I/ρw at the desired magnetic surface, where I is the total driven current and ρw is the radial width of the driven current. When the current is driven at the position where a ray trajectory becomes tangent to the magnetic surface (referred to as the `tangential resonance'), Doppler broadening of the current profile is significantly reduced and I/ρw reaches a maximum value. In the case of tangential resonance, the driven current width is mainly determined by the effects of the beam divergence and finite spatial width of the beam at the launching position. The width of the driven current is kept in a range of 2-5% of the minor radius under the assumption of a beam divergence with a full angle of 2°. The dependence of the driven current on both the location of beam injection and the wave frequency is also examined.

A tearing mode with m = 3 and n = 2, destabilized in the steady state high-βp H-mode discharges with edge localized mode (ELM), was completely stabilized by local heating and current drive using the 110 GHz first harmonic O-mode electron cyclotron (EC) wave. Here, m and n are poloidal and toroidal mode numbers, respectively. The optimum EC wave injection angle was determined by identifying the mode location from an electron temperature perturbation profile and a safety factor profile. The optimum injection angle was also determined by scanning a steerable mirror during a discharge. In a typical discharge where the tearing mode is completely stabilized, the ratio of the electron cyclotron heating power to the total heating power is 0.17, and the ratio of the EC driven current to the total plasma current is 0.02. Stored energy and neutron emission rate were higher for the case with EC wave injection than that without EC wave injection, which suggests that the reduction of the stored energy and the neutron emission rate was recovered by the tearing mode stabilization.

Ripple loss of energetic alpha particles and neutral beam ions was calculated for reversed shear discharges in ITER-FDR(Final Design Report). The result indicates that, compared with normal operation with positive shear, reversed shear operation dramatically enhances the ripple loss. Ripple loss of alpha particles can reach 25 % with the maximum heat load of 3.7 MW/m2 on the outboard wall, while that of a 1 MeV negative ion source beam can be as high as 20 % as well. The heat load due to alpha particle loss is marginal in the light of a wall tolerance. The calculation suggests that the Toroidal Field (TF) ripple in a fusion reactor should be designed to be less than 0.6 % at the plasma surface so that ripple loss of alpha particles or beam ions can be acceptably low. Ferritic steel insert to the vacuum vessel, is a probable solution to reduce the ripple loss to an allowable level in the ITER-FDR design with the TF ripple reduction of by a factor of 0.4. With TF ripple reduction, ripple loss for alpha particles and beam ions is expected to decrease to 10 % and 4 %, respectively.

The locality and spatial controllability of ECCD are numerically analyzed by changing the direction and location of EC beam injection. The optimum angles of beam injection are obtained for the most localized current drive and the maximum drive efficiency. The profile of the driven current with the maximum efficiency is very broad, i.e., not suitable for localized current drive. When the location of driven current is controlled by the beam injection from the equatorial plane, the locality is extremely destroyed due to the refraction of the ray according to the location of cyclotron resonance layer. The locality is well kept by the injection from an upper or lower side of the equatorial plane.

Characteristics of energetic ion tail formation during the second-harmonic ion cyclotron range of frequency (2CH ICRF) heating and confinement of tail ions were investigated in negative magnetic shear plasmas on JT-60U. Charge-exchange neutral particle measurements show significantly lower tail temperature in negative-shear plasmas than those in positive-shear plasmas. The tail temperature in negative-shear plasmas increases with plasma current and shows improvement of energetic ion confinement. Charge-exchange neutral fluxes are also lower in negative-shear plasmas than in positive-shear plasmas. Meanwhile, tail ion stored energy is comparable between negative- and positive-shear plasmas. In order to understand these experimental results, simple modelled calculations were carried out with an orbit-following Monte Carlo (OFMC) code. The model calculations suggest that the ICRF power absorption is dominant near the 2CH resonance in the negative-shear plasma and in the low-field side away from the 2CH resonance in the positive-shear plasma. A larger scale length of pressure profile in the negative-shear plasma compensates for a lower density of tail ions than that in the positive-shear plasma. Accordingly, negative-shear plasmas can contain comparable stored energy of energetic tail ions to positive-shear plasmas.

Helium ash removal across the last closed flux surfaces in a deuterium-tritium fuelled tokamak plasma is studied using an orbit-following Monte Carlo simulation. Ripple transport of helium ash ions is induced by ion cyclotron resonance heating (ICRH) of short-lived ions, which are produced by charge exchange of with incoming neutral species. It is found that the required ICRH power in an ITER grade tokamak plasma is 10-15 MW and the helium transport to the first wall is localized within an area given by 1.3 m in the toroidal direction and 0.7 m in the poloidal direction.

Fusion performance of reversed shear discharges with an {ital L}-mode edge has been significantly improved in a thermonuclear dominant regime with up to 2.8 MA of plasma current in the JT-60U tokamak. The core plasma energy is efficiently confined due to the existence of persistent internal transport barriers formed for both ions and electrons at a large minor radius of r/a{approximately}0.7 near the boundary of the reversed shear region. In an assumed deuterium-tritium fuel, the peak fusion amplification factor defined for transient conditions involving the dW/dt term would be in excess of unity. {copyright} {ital 1997} {ital The American Physical Society}

A neutral particle analyser of MeV energy range has been used to measure atomic hydrogen fluxes in ICRF and combined ICRF+NBI-heated JT-60U plasmas. Successful application of a pulse height analysis system to separate the particle signal from a noise produced by DD neutrons and -rays is demonstrated. The energy dependence of the neutron sensitivity and energy resolution of the NPA detectors are comparable with calibration data. Charge exchange on and ions is considered as the dominant process for neutralization of fast ICRF-driven protons. A simplified model based on analysis of the steady-state ion balance equation system is applied to estimate and target densities. Model calculations are used to interpret the observed difference for the relatively low (E 0.6 MeV) energy parts of the atomic hydrogen spectrum. It is confirmed that the model calculations can well explain the dependence of the atomic fluxes on the relative toroidal angle between the injected beam and the analyser, and the energy dependence of the flux ratio between `active' (ICRF + NBI) and `passive' (ICRF alone) phases. The role of the charge-exchange target effect in the plasma density scan is also discussed.

The properties of protons under acceleration 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 ( and ) 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 occurrence of cavity resonance in the and heating discharges reduces the stored energy of the accelerated ions and the incremental energy confinement time.

Ripple enhanced banana drift loss of fast ions is studied on the basis of heat load measurements at the first wall near the outer equatorial plane in JT-60 U. Hot spots in the ion cyclotron range of frequency (ICRF) and neutral beam injection (NBI) heating are observed at the outboard wall by using an infrared camera. In ICRF heating, the total heat load to the wall is around 5% of the injected ICRF power when the plasma-wall clearance gap is less than 0.1 m. The heat flux on the wall increases with the ICRF power and decreases strongly with the plasma density, which is consistent with Fokker-Planck calculations. In NBI heating, hot spots are observed at almost the same positions, and the heat flux and position agree with the banana drift loss as predicted by an orbit following Monte Carlo code

Experiments have been carried out in JT-60U to verify the modelling of fast ion ripple transport. The ripple induced loss was estimated from the neutron decay following neutral beam pulse injection and the loss related heat load on the first wall. Comparison of the lost fraction and the hot spot positions between measurements and orbit following Monte Carlo calculations exhibited good agreement, indicating that the ripple transport governing fast ion losses is explained within the framework of existing theory. Neutral beam heating experiments in JT-60U also indicate that H modes free of ELMs are still obtainable for ripple amplitudes of up to 2.2%

The density perturbation induced by electron cyclotron resonance heating was examined on the JFT-2M tokamak using two different heating methods, a pulse injection and a modulated injection. The measured temporal behaviours of the density perturbations were compared with numerical simulations, in which the ECRH induced flux was modelled as local outward flux. It was found that this model can explain the measured density perturbation satisfactorily

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.

Recent results on the minority ion second harmonic heating on JT-60U are presented. Maximum coupled power reached 6.4 MW. 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 3 MW. The sawtooth stabilization by minority ion second harmonic heating was realized over a wide parameter range, i.e. Ip = 0.9 MA−4 MA, q95 = 2.3−8.6, ne = (1.3−5) × 1019m−3 and PIC ⩾ 2.2 MW. A figure of merit VP〈ne〉Ptot 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 2.33 s. 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 (q95 ⩽ 2.9).

A review of the results from JT-60 ICRF heating experiments is presented. The JT-60 ICRF experiments (120 and 131 MHz) are characterized by higher-harmonic heating with phase control of the compact 2*2 loop antenna array. Second-harmonic heating is most intensively investigated with antenna phase control for various cases where the resonant species (hydrogen) are either a majority component or a minority component, and with or without hydrogen HBI heating. It is found that antenna phase control plays a key role in optimizing antenna power injection characteristics as well as heating efficiency. The out-of-phase mode (( pi ,0) mode) can avoid RF sheath and parametric decay instabilities which cause harmful effects on antenna power injection capability with the in-phase mode ((0,0 ) mode) is some operating conditions. The incremental energy confinement time ( tau Einc) of the ( pi ,0) mode is about 50% better than that of the (0,0) mode. Minority hydrogen second-harmonic heating with ( pi ,0) phasing in helium discharges shows the best results over a wide range of electron densities (ne=2.5-6.5*1019 m-3) and plasma currents (Ip=1.5-2.4 MA (qa=4.2-2.6)). The best confinement enhancement factor ( tau E/ tau E(L-mode)) is 1.3, which is obtained with ohmic target plasmas. Higher-harmonic beam ion acceleration is observed up to fourth harmonics in a combined ICRF (( pi ,0) mode) and NBI heating scenario. Strong central-electron heating associated with beam ion acceleration is observed in combined third-harmonic ICRF and NBI heating.

Current drive via nonresonant interaction between radio‐frequency (rf) waves and plasma is studied. The averaged force of rf waves acting on each species of a plasma can be divided into a resonant force and a nonresonant one. A part of the nonresonant force cannot be expressed by a gradient of a scalar potential and remains after integrating along the direction of the force. This force mainly acts as an internal force among plasma species and the net momentum input from the wave to the plasma is usually small. This process is not included in the conventional current drive schemes but is associated with the rf wave helicity injection. Quantitative analysis using a one‐dimensional kinetic wave code is applied to waves in the ion cyclotron range of frequencies and low‐frequency Alfven waves in a large tokamak. The driven current is estimated taking account of the effect of the toroidally trapped particles. The spatial profile of the forces acting on electrons and ions as well as the driven current are obtained. The parameter dependence of the current drive efficiency on the wave number, the plasma density, the temperature, and the toroidal magnetic field is compared with the estimate based on a local analysis.

The relation between RF-driven current and wave helicity is analytically and numerically studied for a tokamak plasma. The helicity conversion ratio from the wave to the plasma is discussed and numerically examined for the waves in the range of ion cyclotron frequency in JT-60 plasma. The wave propagation equation which includes kinetic effects is solved as a boundary-value problem with one-dimensional inhomogeneities. Both the RF helicity-driven current and the RF momentum driven current are calculated. It is shown that the efficiency of the current drive by the wave helicity is small compared with the conventional scheme for the fast wave excitation of the present-day plasma parameters.

Results are presented of the second harmonic heating experiments in the ion cyclotron range of frequencies (ICRF) on JT-60. Heating and confinement characteristics are investigated by using the hydrogen second harmonic (H-majority 2ωcH scheme and the hydrogen minority second harmonic (H-minority 2ωcH scheme (nH/nHe ≈0.1) in two toroidal phasing modes of antenna currents, in phase ((0,0)) and out of phase ((π, 0)), at PIC ≤ 3 MW, ne =(1.3−6.6) × 1019 m−3 and Ip = 1–2.4 MA. Efficient plasma heating has been demonstrated in both heating schemes. Intense non-Maxwellian ion tails are observed at low density. Sawtooth oscillations with an enhanced period in the electron temperature are found over a wide range of electron densities. The incremental energy confinement time, τEinc, for (π, 0) phasing is larger than for (0,0) phasing, irrespective of the heating scheme. In particular, H-minority 2ωcH heating with (π, 0) phasing shows excellent plasma heating with τEinc ≈ 100–120 ms, which is larger than that of neutral beam injection (NBI) heating by a factor of about two. Different dependences of τEinc on Ip are obtained in the two heating schemes. The (0,0) phasing data of τEinc in H-majority 2ωcH heating increase with Ip up to 1.9 MA, while the (π, 0) phasing data in H-minority 2ωcH heating are almost constant with Ip up to 2.4 MA. The global energy confinement shows an L-mode behaviour. The energy confinement time for H-minority 2ωcH heating at high density agrees well with the Shimomura-Odajima scaling.

Beam acceleration by heating in the second-harmonic ion cyclotron range of frequency (ICRF) in combination with heating by hydrogen neutral beam injection (NBI) was investigated in the JT-60 tokamak. The energy spectra of the accelerated fast ions were measured by a charge exchange neutral energy analyser whose line of sight was intersected by specific beam lines in the plasma core in order to obtain the required information. The dependences of the tail ion temperature on various parameters (electron density, NBI power and the toroidal phasing of the antenna) were examined. Optimized conditions for beam acceleration were evaluated. For combined ICRF heating and NBI heating, an incremental energy confinement time of 210 ms was achieved, which was three times higher than that obtained with NBI heating alone or with ICRF heating alone. This improvement of the energy confinement during combined NBI and ICRF heating can be explained by a build-up of the fast ions accelerated by the ICRF wave. The scaling of the incremental energy confinement time during combined ICRF+NBI heating was obtained.

Since 1996, the remote participation in JT-60 experiments has been successfully conducted in collaboration with JAERI, Los Alamos National Laboratory and Princeton Plasma Physics Laboratory. The remote analysis system is a computer complex consisting of an analysis server and a data server for access to JT-60 data. Collaborators of the remote participation can analyze JT-60 data by using this system through the Internet. For the remote diagnostic system, an X Window control program has been customized to communicate via the overseas line using a UNIX workstation at the remote site and one in the JT-60 control room. It has a feature of real-time remote control of diagnostic equipment and remote access to CAMAC data. Both systems are used together with the ISDN-based video-conferencing system for real-time communication between the remote laboratory and the JT-60 control room.

The design of the modification of JT-60U, JT-60SA has been optimized from the viewpoint of plasma performance, and operation regimes have been evaluated with the latest design. Upper and lower divertors with different geometries will be prepared for flexibility of the plasma shape, which will enable both low aspect ratio (A ~ 2.65) and ITER shape (A = 3.1) configurations. The beam lines of negative-ion neutral beam injection will be shifted downwards by ~0.6 m for the off-axis current drive (CD), in order to obtain a weak/reversed shear plasma, as well as having the capability of heating the central region. The feedback control coils along the openings in the stabilizing plate are found effective in suppressing the resistive wall mode and sustaining high βN close to the ideal wall limit. Sustainment of plasma current of 3–3.5 MA for 100 s will be possible in ELMy H-mode plasmas with moderate heating power, βN, and density within an available flux swing. It is also expected that higher βN, high-density ELMy H-mode plasmas will be maintained for 100 s with higher heating power. The expected regime of full CD operation has been extended with upgraded heating and CD power. Full CD operation for 100 s with reactor-relevant high values of normalized beta and bootstrap current fraction (Ip = 2.4 MA, βN = 4.3, fBS = 0.69, , HH98y2 = 1.3) is expected in a highly-shaped low-aspect-ratio configuration (A = 2.65).

Recent results of control of current profile and instability using radiofrequency wave in JT‐60U and prediction analysis in JT‐60SA are descried. In JT‐60U, control of current profile in high‐beta regime was demonstrated by using a real‐time system, where the motional Stark effect diagnostic and lower hybrid wave were used as a detector and actuator, respectively. The minimum value of the safety factor was raised from 1.3 to 1.7 so as to follow the commanded value. Complete stabilization of a neoclassical tearing mode (NTM) with the poloidal mode number m = 2 and the toroidal mode number n = 1 was demonstrated using electron cyclotron (EC) current drive. By scanning the location of EC current drive in detail, strong stabilization effect was found for misalignment less than about half of the full island width. In addition, destabilization of the 2/1 NTM was observed for misalignment comparable to the full island width. Simulation of NTM stabilization in JT‐60SA was performed by using the TOPICS code combined with the modified Rutherford equation. The TOPICS simulation showed that complete stabilization can be achieved more effectively by optimizing the EC wave injection angle and modulating the EC wave.Recent results of control of current profile and instability using radiofrequency wave in JT‐60U and prediction analysis in JT‐60SA are descried. In JT‐60U, control of current profile in high‐beta regime was demonstrated by using a real‐time system, where the motional Stark effect diagnostic and lower hybrid wave were used as a detector and actuator, respectively. The minimum value of the safety factor was raised from 1.3 to 1.7 so as to follow the commanded value. Complete stabilization of a neoclassical tearing mode (NTM) with the poloidal mode number m = 2 and the toroidal mode number n = 1 was demonstrated using electron cyclotron (EC) current drive. By scanning the location of EC current drive in detail, strong stabilization effect was found for misalignment less than about half of the full island width. In addition, destabilization of the 2/1 NTM was observed for misalignment comparable to the full island width. Simulation of NTM stabilization in JT‐60SA was performed by using the TOPICS code combin...

Integrated tokamak modelling that enables the simulation of an entire discharge period is indispensable for designing advanced tokamak plasmas. For this purpose, we extend the integrated code TOPICS to make it more suitable for transient analyses in the fast-ion part. The fast-ion Fokker–Planck solver is integrated into TOPICS at the same level as the bulk transport solver so that the time evolutions of the fast ion and the bulk plasma are consistent with each other as well as with the equilibrium magnetic field. The fast-ion solver simultaneously handles neutral beam-injected ions and alpha particles. Parallelisation of the fast-ion solver in addition to its computational lightness owing to a dimensional reduction in the phase space enables transient analyses for long periods in the order of tens of seconds. The fast-ion Fokker–Planck calculation is compared and confirmed to be in good agreement with an orbit following a Monte Carlo calculation. The integrated code is applied to ramp-up simulations for JT-60SA and ITER to confirm its capability and effectiveness in transient analyses. In the integrated simulations, the coupled evolution of the fast ions, plasma profiles, and equilibrium magnetic fields are presented. In addition, the electric acceleration effect on fast ions is shown and discussed.

It is found that no current is driven in a central region of a tokamak plasma once the central current density becomes nearly zero ("current hole"), in spite of high electric conductivity, at the 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 a magnetic field in an axisymmetric toroidal plasma.

ECRF experiments in JT-60U have been started since 1999. ECRF system in JT-60U is characterized by a) collector potential depression (CPD) gyrotrons, b) two diamond windows at the gyrotron output and the vacuum vessel, c) a steerable mirror which enables to change a poloidal injection angle, d) two grooved polarizing mirrors which enable to change elliptical polarization of the incident wave, e) sixty-meter evacuated transmission line with corrugated waveguides. Duration and output power of the gyrotrons have been steadily increased, and ECRF of ∼1.5 MW was successfully injected to the torus for 3 seconds with 3 gyrotrons. By using the ECRF system, the following experimental results are obtained: a) An m/n=3/2 neoclassical tearing mode was completely stabilized by the fundamental ECH/ECCD. The optimum injection angle for the mode suppression was identified by scanning the injection angle during a discharge. Real-time detection of a neoclassical tearing mode and ECRF injection were also performed; b) EC-dr...

We calculate the heating power of the neutral beam injection (NBI) in a 〈β〉 = 4.8% high-beta discharge achieved in the Large Helical Device (LHD). We investigate the difference of the heating efficiency and the heating power profile with and without the re-entering fast ion effects. When the re-entering fast ion effects are taken into account, the heating efficiency of the co-injection of the NBI (co-NBI case) is improved and it is about 1.8 times larger than that without the re-entering effects. In contrast, the heating efficiency with the re-entering effects in the counter-injection of the NBI (ctr-NBI case) scarcely differs from that without the re-entering ones. We also study the re-entering fast ion effects on the transport properties in the LHD high-beta discharges. It is found that the tendency of the thermal conductivities on the beta value is not so much sensitive with and without the re-entering effects. In addition, we investigate the difference in the re-entering fast ion effects caused by the field strength and the magnetic configuration. In the co-NBI case, the heating efficiency with the re-entering effect was improved with a decrease in the field strength. In contrast, in the ctr-NBI case, it barely differs by changing the field strength.

Sawtooth relaxation is suppressed during ELM‐free period by second harmonic minority ICRF heating in 4MA H‐mode experiments with high power NBI heating. TAE modes in high plasma current regime (up to 4MA) are investigated with ICRF heating, where high‐n TAE modes (n more than 10) appear. TAE amplitude increases with increasing toroidal mode number. The ion energy spectrum is measured up to 1 MeV during higher harmonic beam ion acceleration (from 2nd to 5th harmonics). The tail ion temperature increases with increasing harmonic number up to 4th harmonics. The lower hybrid (LH) wave deposition profile (therefore the driven current profile) can be controlled by changing launched wave spectra (phase velocity, spectral width, compound spectra) or changing geometry (launching position, safety factor). It is also demonstrated to drive noninductive current in the opposite direction to the ohmic current by LH waves for enhancement of the current profile controllability. Antenna coupling with a large separatrix‐antenna gap is demonstrated for both fast waves and LH waves and some related issues are discussed.

Development of an electron cyclotron (EC) wave system was conducted in an effort to achieve the capability required in JT-60SA. Pulse duration at 1 MW output was extended to 31 s. Transmission line components with a diameter of 60.3 mm were installed in 2011 to reduce temperature rise during a gyrotron oscillation. Development of a dual-frequency gyrotron was started to enable heating and current drive in the core region of the JT-60SA plasma for the toroidal field of 2.3 T. The EC wave frequency was chosen to be 138 GHz to meet the requirements of physics experiments and gyrotron design. Fabrication of the gyrotron began in 2011.

Extension of an operation range for the sawtooth stabilization by a second harmonic minority ion ICRF heating is being pursued in JT‐60U. The sawtooth stabilization is obtained at relatively high density (ne∼3.8×1019 m−3) and low‐q (qeff=4 at Ip=2.4 MA) for OH target plasmas with only 2.8 MW of ICRF power. The sawtooth stabilization is thus obtained at high 〈ne〉/Ptot value of 0.8×1019 m−3 mW−1, which is notably larger than the value achieved on JET, where fundamental resonance minority ion heating is employed. Another peculiar effect of the second harmonic heating is that stabilization occurs only after the inversion radius expands sufficiently, i.e., rinv/a≥0.26. The sawtooth stabilization by ICRF heating becomes easier with NBI‐heated target plasmas (longest stable period up to 1.7 sec), while it tends to be difficult with the LHCD target plasmas.

A plasma current of 2 MA is driven by lower hybrid current drive (LHCD) and the maximum CD efficiency of ηCD=2.5×1019 m−2 A/W is achieved, which agrees with the previously reported scaling: ηCD∼12

Development and integration of models for the whole tokamak plasma have progressed on the basis of experimental analyses and first principle simulations. Integrated models of core, edge-pedestal and scrape-offlayer (SOL)-divertor clarified complex and autonomous features of reactor relevant plasmas. The integrated core plasma model including an anomalous transport of alpha particles by Alfven eigenmodes is developed in the core transport code TOPICS-IB and indicates the degradation of fusion performance. The integrated rotation model is developed in the advanced transport code TASK/TX and clarifies the mechanism of alpha particle-driven toroidal flow. A transport model of high-Z impurities is developed and predicts large inward pinch in a plasma rotating in the direction counter to the plasma current. TOPICS-IB is extended to include the edge-pedestal model by integrating with the stability code, simple SOL-divertor and pellet models, and clarifies the mechanism of pellet triggered ELM. The integrated SOL-divertor code SONIC is further integrated with TOPICS-IB and enables to study and design operation scenarios compatible with both the high confinement in the core and the low heat load on divertor plates.

A new Monte Carlo code based on particle tracing using real coordinates has been developed to properly treat the re-entering particles that repeatedly pass in and out of the last closed flux surface (LCFS). The particle loss due to the charge-exchange reaction has also been taken into account in this code. We apply this new code to the analysis of high-energy particles produced by tangential neutral beams (NBs) of the large helical device (LHD). It is confirmed that reasonable solutions of distribution functions are obtained for particles produced by the tangential-NBs. It is also confirmed that the effect of the particle orbit and the charge-exchange loss on the distribution function is properly included. The shapes of the distribution functions of particles, produced by the tangential-NBs in two temperature cases (1 keV and 0.1 keV), are the same. It is found that the re-entering particles play an important role in the analyses of the distribution function of particles produced by the NBs.

(Received 4 November 2009 /Accepted 28 December 2009)We present a study of ions produced by the perpendicular neutral beam in the high-beta plasma of LargeHelical Device (LHD) that uses the Monte-Carlo code based on orbit following in real coordinates with Coulombcollisions. For no neutrals outside the plasma, the results indicate that there are remarkable differences in thedistribution functions that depend on whether the loss boundary is set on the vacuum vessel wall or on the lastclosed flux surface (LCFS). Also, for finite neutral density

Aiming at optimization of current profile in high-β plasmas for higher confinement and stability, a real-time control system of the minimum of the safety factor (q min) using the off-axis current drive has been developed. The off-axis current drive can raise the safety factor in the centre and help to avoid instability that limits the performance of the plasma. The system controls the injection power of lower-hybrid waves, and hence its off-axis driven current in order to control q min. The real-time control of q min is demonstrated in a high-β plasma, where q min follows the temporally changing reference q min,ref from 1.3 to 1.7. Applying the control to another high-β discharge (βN = 1.7, βp = 1.5) with m/n = 2/1 neo-classical tearing mode (NTM), q min was raised above 2 and the NTM was suppressed. The stored energy increased by 16% with the NTM suppressed, since the resonant rational surface was eliminated. For the future use for current profile control, current density profile for off-axis neutral beam current drive (NBCD) is for the first time measured, using the motional Stark effect diagnostic. Spatially localized NBCD profile was clearly observed at the normalized minor radius ρ of about 0.6–0.8. The location was also confirmed by multi-chordal neutron emission profile measurement. The total amount of the measured beam driven current was consistent with the theoretical calculation using the ACCOME code. The CD location in the calculation was inward shifted than the measurement.

An overview is given of the National Centralized Tokamak (NCT) programme as a research programme for advanced tokamak research to succeed JT-60U. The mission of NCT is to establish high beta steady-state operation for DEMO and to contribute to ITER. The machine flexibility is pursued in aspect ratio and shape controllability for the demonstration of the high-β steady-state, feedback control of resistive wall modes, wide current and pressure profile control capability and also very long pulse steady-state operation. Existing JT-60 infrastructure such as the heating and current drive system, power supplies and cooling systems will be best utilized for this modification.