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Construction of the JT-60SA tokamak was completed on schedule in March 2020. Manufacture and assembly of all the main tokamak components satisfied technical requirements, including dimensional accuracy and functional performances. Development of the plasma heating systems and diagnostics have also progressed, including the demonstration of the favourable electron cyclotron range of frequency (ECRF) transmission at multiple frequencies and the achievement of long sustainment of a high-energy intense negative ion beam. Development of all the tokamak operation control systems has been completed, together with an improved plasma equilibrium control scheme suitable for superconducting tokamaks including ITER. For preparation of the tokamak operation, plasma discharge scenarios have been established using this advanced equilibrium controller. Individual commissioning of the cryogenic system and the power supply system confirmed that these systems satisfy design requirements including operational schemes contributing directly to ITER, such as active control of heat load fluctuation of the cryoplant, which is essential for dynamic operation in superconducting tokamaks. The integrated commissioning (IC) is started by vacuum pumping of the vacuum vessel and cryostat, and then moved to cool-down of the tokamak and coil excitation tests. Transition to the super-conducting state was confirmed for all the TF, EF and CS coils. The TF coil current successfully reached 25.7 kA, which is the nominal operating current of the TF coil. For this nominal toroidal field of 2.25 T, ECRF was applied and an ECRF plasma was created. The IC was, however, suspended by an incident of over current of one of the superconducting equilibrium field coil and He leakage caused by insufficient voltage holding capability at a terminal joint of the coil. The unique importance of JT-60SA for H-mode and high-β steady-state plasma research has been confirmed using advanced integrated modellings. These experiences of assembly, IC and plasma operation of JT-60SA contribute to ITER risk mitigation and efficient implementation of ITER operation.

Active control and stabilization of locked modes (LM) via rotating external magnetic perturbations are numerically investigated under a realistic low resistivity condition. To explore plasma responses to rotating and/or static external magnetic perturbations, we have developed a resistive magnetohydrodynamic code ‘AEOLUS-IT’. By using AEOLUS-IT, dependencies of mode behavior on frequency and polarity of the rotating magnetic perturbation are successfully clarified. Here, the rotational direction of the rotating magnetic perturbation to the equilibrium plasma rotation in the laboratory frame is referred to as ‘polarity’. The rotating magnetic perturbation acts on the background rotating plasma in the presence of a static field. Under such circumstances, there exist bifurcated states of the background rotating plasma, which should be taken into account when studying the dependence of the mode behavior on the rotating magnetic perturbation. It is found that there exist an optimum frequency and polarity of the rotating magnetic perturbation to control the LM, and that the LM is effectively stabilized by a co- polarity magnetic perturbation in comparison with a counter-polarity one.

In JT-60U tokamak, low frequency modes (LFMs) have been reported recently when the m/n = 2/1 neoclassical tearing mode grows sufficiently. LFMs show the low mode frequencies (< 20 Hz) with the rotation direction to the counter-current direction (ctr-direction) toroidally even if the strong injection power to the co-current direction (co-direction) is applied by tangential neutral beam injections. As a candidate to drive the rotation to the ctr-direction, the neoclassical toroidal viscosity (NTV) torque is investigated in this study. Indeed, when a LFM is observed, the estimated "offset velocity" of the NTV torque is larger than the observed toroidal velocity around the q = 2 surface and results in the torque to the ctr-direction (ctr-torque). The torque balance between the NTV torque and the torque induced by the eddy current on the resistive wall is investigated. It is found that the rotation to the ctr-direction having the low mode frequency is obtained as the result of the ctr-torque to the offset velocity. Our investigation suggests the requirement to include the NTV torque in the modelling on the mode locking of neoclassical tearing modes in tokamaks.

The characteristics of ion heat transport inside the internal transport barrier (ITB), which is sustained by weak magnetic shear, are investigated using a cold pulse induced by supersonic molecular-beam injection (SMBI) in JT-60U. It is known that cold-pulse propagation in ITBs usually significantly reduces the temperature and degrades the temperature gradient. When subsequent SMBIs are launched before the temperature has recovered, it is observed that the temperature gradient of the ITB is not monotonically decreased but alternately decreased and increased. Alternating decreasing and increasing phases of the ITB temperature gradient continue for about 1 s (∼6τ E), and the properties of the cold-pulse propagation and the flux-gradient relations differ according to the phase. The usual transient transport analysis is also provided by inducing a cold pulse in stationary ITBs. A rapid reduction and recovery of the temperature is observed inside the ITBs, and it is found that the ion-heat flux changes without a variation in the local ion-temperature gradient or a change to the other observable local parameters. The flux-gradient relations exhibit significant hysteresis in two ITB cases: (i) ITBs with strong electron-density gradients and (ii) ITBs with weak ion-temperature gradients. In addition, the range in which hysteresis appears is most likely to depend on the width of the ITB.

We report the experimental finding of n = 1 helical cores (HCs) accompanied by saturated m/n = 2/1 tearing modes (TMs) with low mode frequencies in JT-60U. The HCs accompanied by TMs were observed after an increase in the mode amplitude and a decrease in the mode frequency of m/n = 2/1 precursors with tearing parity. The decreased mode frequency is typically lower than 20 Hz. With various diagnostics, the coupling of n = 1 HCs and m/n = 2/1 TMs has been clearly observed. Because the coherent oscillations in the ion temperature are observed in both the core region and the edge region, the flux surfaces including the m/n = 2/1 magnetic island appear to have m = 1 helical deformation. It has also been suggested that the m/n = 2/1 TM and the HC rotate in the electron diamagnetic direction keeping fm/n=1/1(HC) = 2fm/n=2/1(TM) in several plasmas. Here, fm/n=1/1(HC) is the mode frequency of HCs and fm/n=2/1(TM) is the mode frequency of TMs. In addition, the core seems to be shifted to the high-field side when the O-points of the m/n = 2/1 magnetic island line up in the midplane, which is confirmed by reconstructions of magnetohydrodynamic equilibria with motional Stark effect measurement and the MEUDAS code. Our observation of m/n = 2/1 TMs having HCs contributes to the understanding of the excitation mechanism of HCs in tokamak plasmas.

Active control and stabilization of locked modes via rotating external magnetic perturbations are numerically investigated under a realistic low resistivity condition. To explore plasma responses to rotating and/or static external magnetic perturbations, we have developed a resistive magnetohydrodynamic code "AEOLUS- IT". By using AEOLUS-IT, dependencies of mode behavior on frequency and polarity of the rotating magnetic perturbation are successfully clarified. Here, the rotational direction of the rotating magnetic perturbation to the equilibrium plasma rotation in the laboratory frame is referred to as "polarity". The rotating magnetic perturbation acts on the background rotating plasma in the presence of a static field. Under such circumstances, there exist bifurcated states of the background rotating plasma, which should be taken into account when studying the dependence of the mode behavior on the rotating magnetic perturbation. It is found that there exist an optimum frequency and polarity of the rotating magnetic perturbation to control the locked mode, and that the locked mode is effectively stabilized by a co-polarity magnetic perturbation in comparison with a counter-polarity one.

Impact of the electron cyclotron range of frequency wave (ECRF) on the internal transport barriers (ITBs) in a weak shear (WS) plasma has been investigated in JT-60U. The fundamental O-mode ECRF of 110 GHz injected obliquely (co-current drive) from the low field side is used. It is observed that the ion temperature (Ti) ITB in a WS plasma can be degraded by ECRF. It is clarified for the first time that the degradation depends increasingly on the EC power (PEC) but decreasingly on the plasma current (Ip). Moreover it is confirmed that ECRF affects the toroidal rotation (Vt) indirectly and results in the flattening of Vt(ρ) and therefore the radial electric field (Er) profiles regardless of the direction of the target Vt(ρ), peaking co or counter direction (relative to the Ip direction). Furthermore, it is recently found that Ti and Vt in the whole ITB region are affected with almost no delay from the EC onset even with off-axis EC deposition. These results indicate that EC injection unveiled a semi-global structure that characterizes Ti ITB in a WS plasma.

Controllability of plasmas with a large bootstrap current fraction (fBS) has been investigated in JT-60U. Real time control logic for avoidance of collapses by pressure profile control through the toroidal rotation has been newly installed for the long sustainment of reversed magnetic shear plasmas. The new real time control logic can control the injection timing of the neutral beam based on the real time detection of the minimum value of safety factor (qmin) using the motional Stark effect diagnostic. Using the real time control logic, the weak reversed magnetic shear plasma with fBS ~ 70% is sustained for ~8 s. In such a plasma, dynamic change in the current profile, especially qmin, which was induced by the change in the pressure profile at the internal transport barrier (ITB) through the rotation control, was observed. The response of qmin to the change in the ion temperature gradient at the ITB for large fBS plasmas is stronger than that for smaller fBS plasmas.

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.

After installation of ferritic steel tiles, fast ion losses due to toroidal field ripple have been reduced by 1/2–1/3. The increase in absorbed power at same injection power can reduce the required number of neutral beam injector (NBI) units to sustain a given normalized beta, βN, resulting in a better flexibility of torque input by increasing the available combination of tangential NBI units. By making use of these advantages to sustain an internal transport barrier (ITB), the performance of long-pulse ELMy H-mode plasmas was improved in terms of sustained duration time for both high βN and high thermal confinement enhancement factor (HH98(y,2)). High βN > 2.3 together with HH98(y,2) ~ 1 was sustained for 23.1 s (~12τR, where τR is the current diffusion time) at q95 ~ 3.3, which also provide high βNHH98(y,2) ≥ 2.2 and a bootstrap current fraction of ≥40%. βNHH98(y,2) of 2.0 was sustained for 28.6 s, which is limited by the maximum injection period of 30s for NBI system. These long-pulse plasmas are possible candidates for ITER hybrid operation scenario. Improved confinement is characterized by the larger thermal components at a given density maintained by lower heating power than in previous experiments. The strength of the ITB depends on the pedestal temperature, which varies with edge density while keeping constant the edge pressure (limited by type I ELMs). The fact that co-toroidal rotation as a result of reduced fast ion losses provides better quality of Te-ITB also contributes the improvement of thermal plasma confinement. These long-pulse plasmas indicate that further investigation to establish high performance plasmas longer than the time scale of wall saturation (τW) with active particle control is essential to establish the operational scenarios for next step devices, where the wall pumping does not work.

Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

The m = 2/n = 1 neoclassical tearing mode (NTM) has been observed to strongly degrade confinement and frequently lead to a disruption in high β discharges in DIII-D if allowed to grow to a large size. Stabilization of grown NTMs by the application of a highly localized electron cyclotron current drive (ECCD) at the island location has led to operation at an increased plasma pressure, up to the no-wall kink limit. After the NTM is stabilized by the ECCD, the correct location for the current drive is maintained using information from real-time equilibrium reconstructions which include measurements from the motional Stark effect diagnostic. This same process is used alternatively to prevent the mode from ever growing, leading to performance at the pressure limit in high performance hybrid discharges with β above 4%. Modelling using the modified Rutherford equation shows that the required power is in close agreement with the experimental threshold for the prevention of the 2/1 NTM.

Long-spatial-scale density fluctuations with a wave number of the order of 1 cm−1 were suppressed in the internal transport barrier (ITB) region of a JT-60U reversed magnetic shear plasma, after a cold pulse was induced by the shallow pellet injection at the plasma edge. The particle and ion heat diffusivities are reduced with the suppressed fluctuations. The reduction of the electron heat diffusivity is observed only transiently in the inner ITB portion around the layer of the density fluctuation measurement during the cold pulse propagation and it is slower (~ 15 ms) than the reduction of the density fluctuation level (< 5 ms). In the outer ITB portion, the electron heat diffusivity increases transiently. These results indicate decoupling of the electron heat transport and the long-spatial-scale density fluctuations.

A systematic study of the effect of the level of toroidal plasma rotation at the top of the ion temperature pedestal ( ) 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 (βp) in the range 0.84 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%.

Characteristics of the edge pedestal has been studied for the high βp ELMy H mode and the reversed shear ELMy H mode in JT-60U. The contribution of the pedestal is around 20% for the confinement improvement factor, the normalized beta, and the bootstrap fraction. The upper boundary of the fusion performance measure increases with the pedestal poloidal beta, βp-ped. At high triangularity, δ, βp-ped increases with the total βp (βp − tot) almost linearly for positive shear type I ELMy discharges. This dependence is not due to the profile stiffness, since the dependence is the same for the discharges both with and without an internal transport barrier (ITB). In the reversed shear ELMy H mode, βp-ped increases with βp-tot on the same line as the positive shear cases except at high q95 = 8–9.3. In the high plasma current regime, the final structure with both an ITB and edge transport barrier seems to be determined by the balance between the expanding ITB-foot radius and deepening ELM penetration: the ELM penetration radius deepens with increasing pedestal stored energy and then reaches the ITB-foot radius. The ITB-foot seems to behave as a barrier against the ELM crash penetration, and shrinks after a few ELM attacks, and the ELM penetration follows the shrinking ITB-foot.

In the Large Helical Device, the structure of localized oscillations at the m/n = 2/1 rational surface during neutral beam (NB) injection in the counter direction (counter-NB injection) has been investigated in detail using an electron cyclotron emission (ECE) diagnostic. Here m and n are poloidal and toroidal mode numbers, respectively. Under the condition of increasing plasma current in the counter direction and nearly constant pressure, the oscillations are observed at larger magnitudes of plasma current (−20 kA T−1). In the NB reversal experiments, where the direction of the NB is changed from the co- to the counter-direction and vice versa, the oscillations were observed only for the former case. These results suggest that the profile and magnitude of the plasma current play an important role in the mode onset. Analysis based on the Mercier criteria shows that for a flat or a hollow current profile an interchange mode becomes more unstable at higher negative plasma current, showing a tendency similar to the experimental observations. In addition, radial displacement due to the localized oscillations evaluated from electron temperature perturbation profiles with the ECE diagnostic agrees well with that predicted by a three-dimensional ideal magnetohydrodynamic stability code.

Neoclassical tearing modes (NTMs) will be the principal limit on performance in ITER in the standard scenario, which has beta well below the ideal kink limit. Measurements of island size from ASDEX Upgrade, DIII-D and JET in beta rampdown experiments are used to determine the marginal size for m/n = 3/2 NTM removal. This is compared with data from ASDEX Upgrade, DIII-D and JT-60U with removal of the 3/2 NTM by electron cyclotron current drive (ECCD) at near constant beta. The empirical marginal island size is consistent in both sets of removal experiments and is found to be about twice the ion banana width. A common methodology is developed for fitting the saturated m/n = 3/2 island before (or without) ECCD in all four experimental devices, ASDEX Upgrade, DIII-D, JET and JT-60U. To this is added (and model tested to experiments) the effect of un-modulated co-ECCD on the island width due to replacing the missing bootstrap current and making the tearing stability parameter Δ' more negative. The common model is then used to evaluate the ITER ECCD system, with or without modulation, for both the m/n = 3/2 mode, which is benchmarked here, as well as the m/n = 2/1 NTM. The ITER ECCD top launch system with 20 MW of power is found to be effective in greatly reducing the size of the islands. An m/n = 2/1 mode locking model is used to show that the rotation in ITER should be sufficient for the island reduction by ECCD to avoid locking that causes loss of H-mode and disruption.

Results from stabilization of neoclassical tearing modes (NTMs) in JT-60U are described. NTM stabilization and confinement improvement have been demonstrated by employing a real-time NTM stabilization system, where the identification of the location of an NTM and the optimization of the injection angle of the electron cyclotron wave are performed in real time. Also, a high-beta plasma with the normalized beta of 3 has been sustained by suppressing NTM by applying the electron cyclotron current drive (ECCD) before the onset (`preemptive ECCD'). In addition, a simulation code for analysis of the NTM evolution has been developed by combining the modified Rutherford equation with the transport code TOPICS. It is found that the simulation well reproduces the NTM behavior in JT-60U. The simulation also shows that the ECCD width is also important for NTM stabilization, and that the EC wave power for complete stabilization can be reduced by narrowing the ECCD profile.

Transient transport experiments are performed in plasmas with and without internal transport barriers (ITB) on LHD and JT-60U. The dependence of χe on the electron temperature, Te, and on the electron temperature gradient, ∇Te, is analysed with an empirical non-linear heat transport model. In plasmas without an ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation: the χe depends on Te and ∇Te in JT-60U, while the ∇Te dependence is weak in LHD. Inside the ITB region, there is none or weak ∇Te dependence both in LHD and JT-60U. Growth of the cold pulse driven by the negative Te dependence of χe is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U).

High performance reversed shear (RS) discharges with strong internal transport barrier (ITB) and flat pressure profile in the plasma core region disrupt frequently even with low beta. We analysed MHD instabilities leading to low beta disruption with measuring fluctuations and current profile with MSE measurement. We mainly observed two types of disruptions. One is the disruption without precursor at surface q ~ integer. The other is the disruption with n = 1 precursor of γ > 100 ms. The poloidal mode number of the n = 1 mode is equal to the outermost integer of q. The n = 1 mode exists continuously from the peripheral region to the ITB layer or the n = 1 modes exist separately at the peripheral region and at the ITB layer. The phase has a difference of 180° between at the peripheral region and at the ITB layer. To explain these characteristics of disruption, we introduce a simple model where disruption occurs when both MHD instabilities at the plasma surface and at the q = m/n surface in the RS region, where m and n are the poloidal and the toroidal mode numbers of the surface MHD instability, are simultaneously unstable. This simple model can explain almost all observed disruption by two processes. One is the surface mode triggered disruption, which occurs when surface q changes; the corresponding q = m/n surface at ITB layer changes discretely. The other is the internal mode triggered disruption, which occurs when the corresponding q = m/n surface becomes unstable gradually.

Stabilization of an m = 3/n = 2 neoclassical tearing mode (NTM) by a local electron cyclotron current drive (ECCD) has been studied in high βp ELMy H-mode plasmas in the JT-60U tokamak. In the stabilization experiment, the electron cyclotron (EC) injection timing is scanned from the 'before mode onset' phase (preemptive ECCD) to the mode saturation phase (late ECCD). It has been demonstrated that the ECCD is more effective when it is applied before the mode onset and the critical timing for effective ECCD is related to the mode excitation phase. The EC power to prevent the 3/2 NTM excitation by preemptive ECCD is lower by about 20% compared with complete stabilization by late ECCD. Even when the 3/2 NTM is excited, the stabilization effect is stronger than for late ECCD. For a high performance plasma of the normalized beta, βN = 3.0, the H-factor H89PL improves from 1.9 to 2.1 when the 3/2 NTM is suppressed. A numerical study has been made based on the modified Rutherford equation coupled with a 1.5D transport code and an EC code. The simulation reproduces well the time evolution of the magnetic island both at the growing and stabilizing phases. The simulation results show that it is important to adjust not only the EC current position but also its width to stabilize the NTM effectively. The reduction of EC power for NTM prevention in experiments for preemptive ECCD can be explained by the increase of the minimum seed island size for NTM excitation.

The Large Helical Device is a heliotron device with L = 2 and M = 10 continuous helical coils with a major radius of 3.5–4.1 m, a minor radius of 0.6 m and a toroidal field of 0.5–3 T, which is a candidate among toroidal magnetic confinement systems for a steady state thermonuclear fusion reactor. There has been significant progress in extending the plasma operational regime in various plasma parameters by neutral beam injection with a power of 13 MW and electron cyclotron heating (ECH) with a power of 2 MW. The electron and ion temperatures have reached up to 10 keV in the collisionless regime, and the maximum electron density, the volume averaged beta value and stored energy are 2.4 × 1020 m−3, 4.1% and 1.3 MJ, respectively. In the last two years, intensive studies of the magnetohydrodynamics stability providing access to the high beta regime and of healing of the magnetic island in comparison with the neoclassical tearing mode in tokamaks have been conducted. Local island divertor experiments have also been performed to control the edge plasma aimed at confinement improvement. As for transport study, transient transport analysis was executed for a plasma with an internal transport barrier and a magnetic island. The high ion temperature plasma was obtained by adding impurities to the plasma to keep the power deposition to the ions reasonably high even at a very low density. By injecting 72 kW of ECH power, the plasma was sustained for 756 s without serious problems of impurities or recycling.

Evolution of the current density profile associated with magnetic island formation in a neoclassical tearing mode plasma is measured for the first time in JT-60U by using a motional Stark effect diagnostic. As the island grows, the current density profile turns flat at the radial region of the island and a hollow structure appears at the rational surface. As the island shrinks the deformed region becomes narrower and finally diminishes after the disappearance of the island. In a quiescent plasma without magnetohydrodynamic instabilities, on the other hand, no deformation is observed. The observed deformation in the current density profile associated with the tearing mode is reproduced in a time dependent transport simulation assuming the reduction of the bootstrap current in the radial region of the island. Comparison of the measurement with a calculated steady-state solution also shows that the reduction and recovery of the bootstrap current at the island explains the temporal behaviours of the current density and safety factor profiles. From the experimental observation and simulations, we reach the conclusion that the bootstrap current decreases within the island O-point.

The energy loss for grassy edge localized modes (ELMs) has been studied to investigate the applicability of the grassy ELM regime to ITER. The grassy ELM regime is characterized by high frequency periodic collapses of 800–1500 Hz, which is ~15 times faster than that for type I ELMs. The divertor peak heat flux due to grassy ELMs is less than 10% of that for type I ELMs. This smaller heat flux is caused by a narrower radial extent of the collapse of the temperature pedestal. The different radial extent between type I ELMs and grassy ELMs agrees qualitatively with the different radial distribution of the eigenfunctions as determined from ideal MHD stability analysis. The dominant ELM energy loss for grassy ELMs appears to be caused by temperature reduction, and its ratio to the pedestal stored energy was 0.4–1%. This ratio is lower by a factor of about 10 than that for type I ELMs, which typically have between 2–10% fractional loss of the pedestal energy. A systematic study of the effects of counter (CTR) plasma rotation on the ELM characteristics has been performed using a combination of tangential and perpendicular neutral beam injections (NBIs) in JT-60U. In the high plasma triangularity (δ) regime, ELM characteristics (e.g. amplitude, frequency and type) can be changed from type I ELMs to high frequency grassy ELMs as the CTR plasma rotation is increased. On the other hand, in the low δ regime, complete ELM suppression (QH-mode) can be sustained for long periods up to 3.4 s (~18τE or energy confinement times), when the plasma position in terms of the clearance between the first wall and the plasma separatrix is optimized during the application of CTR-NBIs. In JT-60U, a transient QH phase was also observed during the CO-NBI phase with almost no net toroidal rotation at the plasma edge.

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-βp ELMy H-mode discharges have been optimized under nearly full non-inductive current drive conditions by the large bootstrap current fraction (fBS ~ 45%) and the beam driven current fraction (fBD ~ 50%), which was sustained for 5.8 s in the stationary condition. This duration corresponds to ~26τE and ~2.8τR, which was limited by the pulse length of negative-ion-based neutral beams. The high confinement enhancement factor H89 ~ 2.2 (HH98y2 ~ 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 (fBS ~ 75%) has been sustained for 7.4 s under nearly full non-inductive current drive conditions. This duration corresponds to ~16τE and ~2.7τR. The high confinement enhancement factor H89 ~ 3.0 (HH98y2 ~ 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.

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.

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-?p plasmas. ITB events are observed as the simultaneous rise and decay of Te 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 Pnbi (above ~8?MW for the 1.2?1.5?MA/3.8?T pulses described below), ITB events were previously detected at various qmin values. However, the role of qmin 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 qmin. 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 Te and Ti at sufficient power is highlighted (even when the variation is significant and complicated in space and time). Within errorbars, ITB splitting occurs as qmin passes through 2.5. The similarity of space?time evolution of Te and Ti 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 qmin passes through 3 (after periodical improvements and degradations via ITB events with 8?ms period) in H-mode, with Pnbi = 8?MW, is described. At lower powers, ITB events are observed only near rational values of qmin. In weak RS shots with Pnbi = 4?MW, transport is reduced via ITB events during 0.08?s at qmin = 3.5, and repetitive short-term phases of reduced transport are observed as qmin passes through 3. The behaviour of Ti looks different. The difference in Te and Ti evolution, which was detected regularly under low power, probably indicates a decoupling of Te and Ti transport.

The results of studies on heating and current drive by the electron cyclotron (EC) waves in JT-60U are presented. An electron temperature of up to 26 keV, as measured by ECE diagnostics, was achieved by injecting EC waves into the centre of a reversed shear plasma produced by the lower hybrid (LH) waves. The electron temperature Te exceeds 24 keV in the region ρ < 0.3, where ρ is the normalized minor radius. The EC driven current profile was measured at high Te up to 21 keV without using LH waves, and was found to be spatially localized. The ECCD (current drive) efficiency ηCD increased with Te and reached 0.42 × 1019 A W−1 m−2 at Te = 21 keV. The dependence of normalized CD efficiency on deposition location was also studied to optimize the CD efficiency, since the trapped particle effect, which depends strongly on deposition location, is expected to reduce ζ. The effect was detected from a significant decrease in ζ in the lower magnetic field deposition, which is consistent with a linearized Fokker–Planck calculation. The measured CD efficiency ζ increased with electron density ne for the same Te, which showed a stronger dependence on ne compared with the linear calculation. Further off-axis ECCD profile at about half of the minor radius was measured, showing fairly good agreement with linear calculation.

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 (kθ0). 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 kθ0 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.

The influence of dominant electron heating on internal transport barriers (ITBs) was investigated in JT-60U. The electron cyclotron waves were mainly utilized in order to extend the experimental space into a dominant electron heating regime. The experiments were carried out on both reversed magnetic shear (RS) and positive magnetic shear (PS) plasmas. It was found that in RS plasmas clear ITBs in Te, Ti and ne profiles remained even under dominant electron heating. Energy confinement in excess of twice the H-mode scaling has been obtained in the regime. On the other hand, it was found that in PS plasmas the Ti ITB was degraded by dominant electron heating.

A review of magnetohydrodynamic limits to tokamak operation in terms of current, density and pressure is given. Although the current and density limits in a tokamak usually lead to disruptive termination of the discharge, it is argued that these can be avoided by staying away from the respective limits. This is especially true since operation close to these limits is not really desirable, due to the decreased confinement at very high density and the high disruptivity at low q. On the other hand, the limit to plasma pressure set by neoclassical tearing modes (NTMs) and resistive wall modes (RWMs) is too low to guarantee economic operation of future fusion reactors. Therefore, active control of these two instabilities is now being studied. Noticeable progress has been made by NTM stabilization with ECCD. Avoidance of NTMs and RWMs by tailoring sawteeth and spinning the plasma, shows promising results. Also, experiments on direct RWM stabilization by active coils are showing their first encouraging results.

This paper reports results on the progress in steady-state high-βp ELMy H-mode discharges in JT-60U. A fusion triple product, nD(0)τETi(0), of 3.1 × 1020 m−3 s keV under full non-inductive current drive has been achieved at Ip = 1.8 MA, which extends the record value of the fusion triple product under full non-inductive current drive by 50%. A high-beta plasma with βN ~ 2.7 has been sustained for 7.4 s (~60τE), with the duration determined only by the facility limits, such as the capability of the poloidal field coils and the upper limit on the duration of injection of neutral beams. Destabilization of neoclassical tearing modes (NTMs) has been avoided with good reproducibility by tailoring the current and pressure profiles. On the other hand, a real-time NTM stabilization system has been developed where detection of the centre of the magnetic island and optimization of the injection angle of the electron cyclotron wave are done in real time. By applying this system, a 3/2 NTM has been completely stabilized in a high-beta region (βp ~ 1.2, βN ~ 1.5), and the beta value and confinement enhancement factor have been improved by the stabilization.

To measure the electron temperature profile of tokamak plasma in a high confinement mode with edge localized mode (ELMy H-mode), a method of data analysis is presented on the fast-scanning Fourier transform spectrometer (FTS), and results in the JT-60U tokamak are described. In the ELMy H-mode, where the FTS interferogram of electron cyclotron emission (ECE) becomes incomplete due to nonthermal emission pulses of ELM, the spectrum of ECE is estimated by using the maximum entropy method (MEM) of nonparametric type. The optimization of the regularization parameter is studied in terms of adopting the generalized cross validation (GCV) from the linear regularization of Tikhonov–Phillips. A study is also performed on a method to remove the nonthermal emission pulses from interferogram data before the MEM analysis, and on an extension of minimum cross entropy in order to meet with the missing of interferogram values in the region of zero path difference. From the application to experimental data of many shots, it is found that the electron temperature profiles can be automatically measured using the FTS during the ELMy H-mode operation of the tokamak.

Stabilization of an m = 3/n = 2 neoclassical tearing mode (NTM) has been studied experimentally by applying a local electron cyclotron current drive (ECCD) in the JT-60U tokamak. The EC power is injected before the mode onset, and its effect is compared with the ECCD applied at the saturation phase. The experimental results show that the ECCD applied at the growth phase is more effective than that applied at the saturation phase. The necessary EC power for the suppression is reduced and the mode onset is delayed, indicating the hysterisis characteristics of the NTM on the ECCD stabilization. The dependence on the EC power and injection angle is also shown.