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The critical beta, which is limited by an external kink mode with Alfven wave time scale, is shown to be decreased by ferromagnetic effect by about 8% for μ/μ0 ~ 2 where μ and μ0 denote the permeability of a ferromagnetic wall and vacuum, respectively, for a tokamak of aspect ratio 3. The existence of the stability window in the wall distance for resistive wall mode opened by the effects of both the toroidal plasma rotation and the plasma dissipation, which was not observed for the high aspect ratio tokamak, is found for the tokamak of aspect ratio 3. The effect of ferromagnetism on them is also investigated. The critical beta analyses of the National Centralized Tokamak (NCT) plasma using the VALEN code are started by stabilizing plate and vacuum vessel geometry with finite resistivity, and the results of the passive effect of the stabilizing plate are obtained. The calculations including the stabilizing effect of the vacuum vessel and also active feedback control are performed for a design of the NCT plasma.

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.

Advanced technologies for fabrication of superconducting coils have been developed for the National Centralized Tokamak which is based on modification of JT-60. One of the technologies developed is the application of the react-and-wind (R&W) method of fabrication of a Nb3Al D-shaped coil. The bending strain of 0.4% due to the R&W method did not affect the critical current characteristics. This finding indicates the possibilities that the manufacturing cost of large size coils can be reduced further by downsizing the heat treatment furnace, and large complicated shape coils can be manufactured by using the Nb3Al conductor. Another technology is an advanced winding technique for the reduction of the ac losses of Nb3Sn coils by loading bending strain on the conductor. It was found that 0.2% bending strain is enough to reduce the ac losses to one-fifth at the virgin state. The newly developed NbTi conductor attained both (i) low ac loss of 116 ms in coupling time constant and (ii) low cost owing to the stainless steel wrap of the sub-cables and Ni plated NbTi strands with 11 µm filaments.

Argon, krypton and xenon were puffed with and without simultaneous hydrogen gas puffing into Ohmically heated plasmas of the JT-60U tokamak with low plasma currents in order to study the capability of disruption mitigation. It was found that krypton gas puffing can provide a plasma termination with smaller amounts of runaway electrons in comparison to argon and xenon gas puffing.

Plasma control on high-βN steady-state operation for JT-60 superconducting modification is discussed. Accessibility to high-βN 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.

Modification of JT-60 as a full superconducting tokamak (JT-60SC) is planned. The objectives of the JT-60SC programme are to establish scientific and technological bases for steady-state operation of high performance plasmas and utilization of reduced-activation materials in an economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to the DEMO reactor have been developed for the superconducting magnet technology and plasma facing components of the JT-60SC design. To achieve a high current density in a superconducting strand, Nb3Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFCs) of JT-60SC. The R&D to demonstrate the applicability of the Nb3Al conductor to TFCs by a react-and-wind technique has been carried out using a full-size Nb3Al conductor. A full-size NbTi conductor with low ac loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the carbon fibre composite target was successfully demonstrated on an electron beam irradiation stand.

Fast plasma shutdown without runaway electron generation by gas puffing is investigated in the JT-60U tokamak. Argon-only injection enables a fast shutdown; however, it induces runaway electron generation. Hydrogen-only injection generates much less runaway electrons; however, the shutdown time is considerably longer. Mixed injection of hydrogen and argon achieves a fast plasma shutdown without runaway generation. Argon atoms contribute to radiate the energy of plasma leading to a fast plasma shutdown, whilst hydrogen atoms contribute to increase the electron density for avoiding runaway electron generation.

Effect of divertor pumping on the scrape-off layer (SOL) plasma and impurities has been studied in many divertor tokamaks. Recent experimental studies of particle exhaust and SOL flow in JT-60U are reviewed. Pumping flux at the inner and outer pumping slots of the W-shaped divertor was investigated under the attached and detached divertor conditions. The SOL flow pattern was determined, for the first time, at three locations, i.e. at the low-field-side SOL (the outer midplane and divertor X-point) and at the high-field-side SOL (above the inner baffle), using three reciprocating Mach probes. The measured profiles were compared with modelling results from two-dimensional plasma simulations, UEDGE-code, with the ion drift effects included.

JT-60 is planned to be modified to a fully superconducting coil tokamak (JT-60SC) to establish scientific and technological bases for an attractive DEMO reactor. A divertor geometry for JT-60SC is designed to achieve enough exhaust capability for high triangularity configuration, which is essential to obtain high plasma pressure and type II ELMs. For the improvement of detachment control capability, vertical targets are applied with baffle plates into which 3?cm SOL flux surface on the outer midplane can be led. Cryopanels are installed under both a private dome and an outer divertor to allow separated pressure control for inner and outer divertor legs, which is important for the detachment control. The preliminary results of the numerical simulation with SOLDOR/NEUT2D code show that the pumping speed of ?60?m3?s-1 at an outer pumping slot would be enough to control the detachment at the outer divertor without an X-point MARFE under the condition of the edge density of 3.2?1019?m-3 and the SOL power flow of qi = qe = 6?MW.

Termination of the runaway electron current generated during plasma disruptions is found in JT-60U during simulated vertical plasma displacement events where the safety factor at the plasma surface qs decreases. For all discharges with runaway electron generation, the runaway current disappears for qs≥2 with the appearance of spikes in the magnetic fluctuations. The growth rate of the spikes in the magnetic fluctuations decreases by an order of magnitude during the termination of runaway current. Corresponding to the loss of runaway electrons by magnetic fluctuations, heat flux pulses are measured at the inner divertor plates, which indicates interaction of the wall with the runaway electrons. The halo current during runaway termination is small and increases after runaway termination with a dominant toroidal mode of n = 1.

First results of the two-dimensionally reconstructed distribution of divertor radiation in JT-60U are presented. Hardware improvements of in-vessel divertor bolometer cameras to withstand severe electrical and thermal loads and the development of tomography software have made detailed and visual studies of divertor radiation possible. Line-integrated bolometer signals are successfully mapped onto the JT-60U geometry, indicating characteristic profiles for radiative divertor operation. A unique measurement of the radiating layer width at the target plate independently confirms the results of tomographic analysis. Radiation which is almost uniformly distributed along the separatrix flux line from the inboard to the outboard divertor is identified in the moderate density H-mode. Temporal evolution of radiative collapse due to heavy argon radiation is explored. Difficulties in the reconstruction associated with the local hot spot and neutrals are also discussed.

In order to obtain improved confinement plasmas with high radiation at high density, Ar gas was injected into ELMy H mode plasmas in JT-60U. A confinement improvement of HH98(y,2) ≈ 1 was obtained with a high radiation loss power fraction (~80%) at an electron density of ~0.65nGW. The HH factor was about 50% higher than that in plasmas without Ar injection.

A pair of probe pins whose lengths differ by 5 mm are inserted into the separatrix of H-mode plasmas. When the long pin is biased positively against the short one, the H - L transition takes place and the current through the resistor making a connection between these pins increases for 10 and the potential of the pin increases for 100 . The energy dissipated by the resistor during the H - L transition is comparable to the energy of the radial electric field in H-mode plasma. When the long pin is biased negatively, the H - L transition also occurs; however, it takes 1 ms to change the potential of the pin and the plasma returns to H-mode after the extraction of the probe.

Measurements obtained by the novel technique of an assembly of planar Langmuir probes rotating in the scrape-off layer of JFT-2M are described. The probes scan the radial direction and rotate 360 degrees so measuring the direction of the magnetic field and the density of current in the magnetic surfaces. Experiments carried out with and without divertor biasing lead to the following main results: (1) The direction of the ion flow has been measured. (b) It has been found that j*B not=0 with the resulting force always directed towards the magnetic axis. (c) By applying a negative divertor bias a narrow layer appears where the current flows in a direction opposite to the main plasma current

Fluctuations of JFT-2M scrape-off plasma are measured by triple probes in detail. Electrostatic fluctuation can account for the cross field particle flux. When the radial electric field is made by the divertor biasing, reductions of the fluctuations and fluctuation induced flux are observed at the shear layer.

This paper describes an update of the H mode confinement database that has been assembled for the ITER project. Data were collected from six machines of different sizes and shapes: ASDEX, DIII-D, JET, JFT-2M, PBX-M and PDX. The updated database contains better estimates of fast ion energy content and thermal energy confinement times, discharges with RF heating, data using boronization, beryllium and pellets, more systematic parameter scans, and other features. The list of variables in the database has been expanded, and the selection criteria for the standard dataset have been modified. We also present simple scalings of the total and thermal energy confinement time to the new dataset

The dependence of the H-mode energy confinement time τE on the plasma elongation and the divertor configuration is studied. A regression analysis of the H-mode database is presented in which the dependence on the closed/open divertor configuration, estimated from the ASDEX experiments, is included. The dependence of τE on the elongation factor k is found to be positive (approximately k0.6) for ELM-free H-mode discharges, in accordance with experimental results for ELM-free limiter H-mode plasmas in JFT-2M. In addition, the influence of the X-point configuration (single null or double null) on the confinement time is briefly addressed

Describes the content of an H-mode confinement database that has been assembled for the ITER project. Data were collected from six machines of different sizes and shapes: ASDEX, DIII-D, JET, JFT-2M, PBX-M and PDX. A detailed description of the criteria used in the selection of the data and the definition of each of the variables is given. The authors also present an analysis of the conditions of the database, the scalings (power law and offset linear) of the data with both dimensional and dimensionless variables, and predictions of the expected confinement time for ITER

The characteristics of the edge localized model (ELM), also referred to as edge relaxation phenomenon, were investigated in H-mode plasmas of JFT-2M. It was found that the ELM is mainly a density fluctuation phenomenon at the edge and that the electron temperature at the edge, except near the separatrix, is not very strongly perturbed. Some experimental conditions controlling the ELM intensity are plasma density, plasma ion species, heating power and plasma current ramping. ELMs found in low density deuterium discharges are suppressed by raising the density. ELMs are more strongly pronounced in hydrogen plasmas than in deuterium plasmas. ELMs seen in hydrogen plasmas or in near marginal H-mode conditions are suppressed by increasing the heating power. ELMs are found to be suppressed by plasma current ramp-down whereas they are enhanced by current ramp-up. The MHD aspects of the ELM were investigated. Although no consistent MHD features of the ELMs were found, they sometimes trigger bursts of the m = 2 mode. The enhanced broadband magnetic fluctuation suggests a temporary return to the L-mode during the ELM event.

The coupling characteristics of the fast wave excited by a phased four-loop antenna array are described. Fast waves in the lower hybrid frequency range have the potential of generating plasma current in hot and dense plasmas. Fast waves are excited at the plasma edge by an oscillating magnetic field parallel to the toroidal direction. The parallel wavenumber of the excited fast waves is determined by the relative phase of the RF current on each antenna. The loading resistance of the antenna increases with density, but at densities above 2 × 1019 m−3 it starts to decrease with density, because of the steepening of the density gradient. The loading resistance is strongly dependent on the antenna phasing. The maximum loading due to surface wave excitation is obtained at Δ = 0 and the minimum at Δ = π. Substantial absorption of the excited fast waves is observed at the plasma centre when the antenna phasing is Δ = π. The absorption efficiency rises with decreasing phase velocity of the excited waves. The experimental results obtained in the coupling experiment are consistent with theoretical predictions.

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.

The ion Bernstein wave (IBW) is launched into the JFT-2M tokamak plasma in order to investigate the antenna loading resistance and the heating effect. Antenna loading resistance increases as the scrape-off electron density increases, in the same manner as that of fast wave antennas. Although the hydrogen ion temperature increases due to the IBW, most of the wave power is absorbed by the peripheral electrons. This is mainly due to the antenna location, which is at an oblique angle to the midplane. Peripheral electron heating induces the enhanced radiation loss, and IBW power of only about 120 kW can be launched because of the radiation limit.

Current drive by lower hybrid waves that have a phase velocity five times the electron thermal velocity has been investigated in the JFT-2M tokamak. Two different modes of tokamak operation, the automatic current regulation (ACR) and the automatic voltage regulation (AVR) modes, were employed to study the plasma control during the current drive. In the ACR mode, which is the ordinary tokamak operation in JFT-2M, the current of the poloidal-field coils which produce the main magnetic flux outside the torus is controlled by thyristor power supplies, while in the AVR mode their voltage is controlled. The results indicate that control of the poloidal field determines the time evolution of the plasma current during current drive whereas operation in the AVR mode suggests the sustainment of a steady-state plasma current through recharging of the Ohmic transformer. A high current drive efficiency was obtained which increased with the plasma current as up to Ip = 150 kA. The dependence of the current drive efficiency on the plasma current is related to the electron temperature and to propagation effects of the lower hybrid waves. The interaction of the excited lower hybrid waves with the electrons was investigated by pulse height analysis of the soft-X-ray emission. It was shown that the measured soft-X-ray energy spectrum is consistent with the theoretical spectrum of the launcher. The characteristic times for the modification of the electron distribution function and for the drop in loop voltage are proportional to the electron density.

Transitions of L-mode plasmas to the H-mode have been induced by an electron cyclotron heating (ECH) pulse. The transitions occur when ECH is applied to plasmas preheated either by a neutral beam or by waves in the ion cyclotron range of frequency with power levels well below their own threshold power for the H-mode transition. The position of the electron cyclotron resonance layer has been scanned and it has been shown that edge heating rather than central heating is effective in inducing the transition to the H-mode.

Reduction of impurity line emissions associated with the ion cyclotron range of frequency (ICRF) heating is achieved by the phase control of a loop antenna array. Reduction in metal impurity emissions and radiation loss is closely correlated with the amount of power radiated from the antennae with parallel wave number near k∥ = 0. The maximum density attainable without disruption is increased over that in the Ohmic heating phase, by reduction of radiation loss.

A new mode of improved energy confinement, for which the confinement time is not worse or is sometimes even better than that for the well known H-mode and the density of which is in a quasi-stationary state, has been obtained in neutral beam heating experiments on JFT-2M. The new mode is different from the H-mode in many respects. The central electron temperature is higher in the new mode than in the H-mode. Radiation loss and density are reduced in the peripheral region but not in the central region. Therefore, the density and radiation profiles are highly peaked in the new mode, in contrast to the broad profiles in the H-mode. Particle confinement in the peripheral region seems to be worse in the new mode than in the H-mode. The new mode can be obtained in both divertor configurations and limiter discharges in JFT-2M.

Significant improvement of energy confinement has been observed on JFT-2M during ICRF heating. This improvement is preceded by a sudden drop in the Hα/Dα emission and a successive increase in stored plasma energy, electron density and radiation loss. This is believed to be the same phenomenon as the H-mode transition observed in ASDEX, and in PDX divertor experiments with neutral beam injection. However, in JFT-2M, this transition is observed both in limiter discharges and in open divertor configurations.

A Monte Carlo simulation on sloshing ion formation and the associated potential dip is described. It is shown in the GAMMA 6 tandem mirror machine that the sloshing ion distribution can be sustained for about 500 µs. The warm passing ions have a density depression in the region of the sloshing-ion density dip because of the acceleration due to the potential. This effect is included in the evaluation of the potential dip self-consistently. The potential dip is sustained for the duration of sloshing ion formation because the trapped cold ions obtain energy from the passing ions and escape from the potential dip.

As a step towards the creation of a thermal barrier in a tandem mirror, formation of sloshing ion distribution is studied by oblique injection of neutral beams into one of the minimum B mirrors of GAMMA 6. Formation of the sloshing ion distribution is verified from the observations of the pitch angle distribution. The distribution is kept in a quasi-steady state, against relaxation of sloshing ions, to form a broad distribution. Potential dip is also observed in association with the sloshing ion distribution. The potential dip is sustained because the density of cold ions trapped in the dip is small.