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The several hundred keV neutral beamline which is now operating on JT-60U marks the first usage of negative ion sources to produce high power neutral beams. Because this was such a large advance in the state of the art with respect to all system parameters, many new physical processes appeared during the earlier phases of the beam injection experiments. The physical mechanisms responsible for these processes have been explored, and solutions have been implemented for some of them, in particular excessive beam stripping, the secular dependence of the arc and beam parameters, and the non-uniformity of the plasma illuminating the beam extraction grid. This has reduced the percentage of beam heat loading on the downstream grids by roughly a third and has permitted longer beam pulses at higher powers. Progress is being made in improving the negative ion current density and in coping with the sensitivity of the caesium in the ion sources to oxidation by tiny air or water leaks, and the cathode operation is being altered.

Formation of a high performance plasma with βp = 3.6, βN = 1.6, HH98y2 = 1.6 without the use of the Ohmic heating solenoid was demonstrated for the first time. Plasma start-up and plasma current ramp-up were accomplished by vertical field and shaping coils, combined with heating and current drive by EC, LH and NBI. The q profile is deeply reversed with qmin = 5.6 at r/a = 0.7, and the bootstrap current fraction was at least 90%. This result opens up the possibility of eliminating the Ohmic heating solenoid, which has a great impact on improving tokamak and ST reactor designs.

In order to prepare adequate current ramp-up and ramp-down scenarios for ITER, present experiments from various tokamaks have been analysed by means of integrated modelling in view of determining relevant heat transport models for these operation phases. A set of empirical heat transport models for L-mode (namely, the Bohm-gyroBohm model and scaling based models with a specific fixed radial shape and energy confinement time factors of H96-L = 0.6 or HIPB98 = 0.4) has been validated on a multi-machine experimental dataset for predicting the li dynamics within ±0.15 accuracy during current ramp-up and ramp-down phases. Simulations using the Coppi-Tang or GLF23 models (applied up to the LCFS) overestimate or underestimate the internal inductance beyond this accuracy (more than ±0.2 discrepancy in some cases). The most accurate heat transport models are then applied to projections to ITER current ramp-up, focusing on the baseline inductive scenario (main heating plateau current of Ip = 15 MA). These projections include a sensitivity study to various assumptions of the simulation. While the heat transport model is at the heart of such simulations (because of the intrinsic dependence of the plasma resistivity on electron temperature, among other parameters), more comprehensive simulations are required to test all operational aspects of the current ramp-up and ramp-down phases of ITER scenarios. Recent examples of such simulations, involving coupled core transport codes, free-boundary equilibrium solvers and a poloidal field (PF) systems controller are also described, focusing on ITER current ramp-down. © 2011 IAEA, Vienna.

We have carried out experiments to clarify the stabilizing effect of a plasma rotation on the resistive wall mode (RWM) that could limit the achievable-βN in high-βN plasmas above the no-wall ideal βN-limit. On JT-60U plasma rotations are controlled using neutral beams with varying combinations of net torque input while keeping βN constant. The RWM is destabilized as the plasma rotation is being reduced. Detailed measurements of the mode structure revealed that the RWM has a global structure that rotates with the order of the resistive wall time. In these experiments, it is found that the critical toroidal rotation speed for the RWM stabilization is less than 1% of the Alfven speed. Moreover, the critical rotation does not strongly depend on βN. The results suggest that high-βN operation up to the ideal wall βN-limit could be possible by suppressing the RWM with a slow plasma rotation in fusion reactors.

The results of stabilizing neoclassical tearing modes (NTMs) with electron cyclotron current drive (ECCD) in JT-60U are described with emphasis on the effectiveness of the stabilization. The range of the minimum EC wave power needed for complete stabilization of an m/n = 2/1 NTM was experimentally identified for two regimes using unmodulated ECCD to clarify the NTM behaviours with different plasma parameters: 0.2 < j EC/j BS < 0.4 for W sat/d EC ∼ 3 and W sat/W marg ∼ 2, and 0.35 < j EC/j BS < 0.46 for W sat/d EC ∼ 1.5 and W sat/W marg ∼ 2. Here, m and n are the poloidal and toroidal mode numbers; j EC and j BS the EC-driven current density and bootstrap current density at the mode rational surface; W sat, W marg and d EC the full island width at saturation, marginal island width and full-width at half maximum of the ECCD deposition profile, respectively. Stabilization of a 2/1 NTM using modulated ECCD synchronized with a mode rotation of about 5 kHz was performed, in which it was found that the stabilization effect degrades when the phase of the modulation deviates from that of the ECCD at the island O-point. The decay time of the magnetic perturbation amplitude due to the ECCD increases by 50% with a phase shift of ±50° from the O-point ECCD, thus revealing the importance of the phasing of modulated ECCD. For near X-point ECCD, the NTM amplitude increases, revealing a destabilization effect. It was also found that modulated ECCD at the island O-point has a stronger stabilization effect than unmodulated ECCD by a factor of more than 2.

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

The effect of the toroidal field (TF) ripple and edge toroidal rotation (VT) on H-mode and pedestal performance as well as on ELM characteristics are investigated both in JET and JT-60U using matched plasma shape. In JT-60U, the amplitude of TF ripple (δr) was reduced from ~1.2% to ~0.5% (averaged value at BT = 2.2 T) after the installation of ferritic steel tiles (FSTs). In addition with FSTs, δr can also be varied by changing the toroidal magnetic field for a given separatrix position. In JET, the TF system can be configured to feed different currents to the odd and even coil sets out of 32 TF coils. In this operation mode, δr can actively be varied by selecting the appropriate differential current between each coil set. In both devices, it is observed that the edge VT, which was in the same direction as the plasma current (co-VT), reduced with increasing δr. Even at the same δr of 0.5%, the achievable edge VT in JT-60U was still lower than that in JET due to higher neutral beam fast ion losses in JT-60U. A series of power and density scans performed at several δr indicated that plasmas with smaller ripple amplitude and/or larger co-VT are favourable to achieve higher pped and HH factor in both devices. As for ELM characteristics, larger co-VT seems to increase the ELM energy loss together with the reduction of the ELM frequency.

Wall conditioning is essential in tokamak and stellarator research to achieve plasma performance and reproducibility. This paper presents an overview of recent conditioning results, both from experiments in present devices and modelling, in view of devices with superconducting coils, with focus on W7-X, JT-60SA and ITER. In these devices, the coils stay energised throughout an experimental day or week which demands for new conditioning techniques that work in presence of the nominal field, in addition to the proven conditioning methods such as baking, glow discharge conditioning (GDC) and low-Z wall coating through GDC-plasma, which do not work under such condition. The discussed techniques are RF conditioning without plasma current, both in the ion cyclotron and electron cyclotron range of frequencies, and diverted conditioning plasmas with nested magnetic flux surfaces. Similarities and differences between tokamaks and stellarators are highlighted. Finally a conditional tritium recovery strategy for ITER is proposed based on Ion Cyclotron Wall Conditioning and L-mode plasma results from JET, equipped with an ITER-like wall (beryllium main chamber wall and tungsten divertor).