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Lower Hybrid Current Drive (LHCD) experiments performed at density close to that required for the steady state scenario are reported. On C-Mod, FTU and Tore Supra, a strong decay of the brehmsstrahlung emission is observed when the density is increased, much faster that the prediction of LHCD modelling. On JET, LH power deposition is also found to be sensitive to the plasma density: LH power modulation indicates that the power deposition moves to the very edge of the plasma (r/a ~ 0.9) when the density approaches the requirement of the JET SS scenario. From this experiment but also from the reconstruction of the electron cyclotron emission spectrum, the decrease of the LHCD efficiency with density is also found. From LHCD modelling of different JET pulses performed at different densities and wave parallel refraction indexes, it is concluded that the wave accessibility condition is not the key parameter for explaining the decrease of the efficiency. C-Mod, FTU and Tore Supra experiments indicate that the plasma edge parameters, namely density and temperature but also fluctuations, are affecting the efficiency via loss mechanisms which are likely to be collisional damping (C-Mod), parametric decay instabilities or wave scattering (FTU/ Tore Supra).

Abstract. This paper summarizes highlights of research results from the Alcator C-Mod tokamak covering the period 2006 through 2008. Active flow drive, using mode converted waves in the ion cyclotron range of frequencies (ICRF), has been observed for the first time in a tokamak plasma, using a mix of D and 3He ion species; toroidal and poloidal flows are driven near the location of the mode conversion layer. ICRF induced edge sheaths are implicated in both the erosion of thin boron coatings and the generation of metallic impurities. Lower Hybrid RF has been used for efficient current drive, current profile modification, and toroidal flow drive. In addition, LHRF has been used to modify the H-mode pedestal, increasing temperature, decreasing density, and lowering the pedestal collisionality. Studies of hydrogen isotope retention in solid metallic plasma facing components reveal significantly higher retention than expected from ex-situ laboratory studies; a model to explain the results, based on plasma/neutral induced lattice damage has been developed and tested. During gaspuff mitigation of disruptions, induced MHD causes the magnetic field to become stochastic, resulting in reduction of halo currents, spreading of plasma power loading, and loss of run-away electrons before they cause damage. Detailed pedestal rotation profile measurements have been used to infer ER profiles, and correlation with global H-mode confinement. An improved L-mode regime, obtained at q 95 ≤3 with ion drift away from the active x-point, shows very good confinement without a strong density pedestal, and no evidence of particle or impurity accumulation without the need for ELMs or any other edge density regulation mechanism.

Since the first H-mode discharges in 2010, the duration of the H-mode state has been extended and a significantly wider operational window of plasma parameters has been attained. Using a second neutral beam (NB) source and improved tuning of equilibrium configuration with real-time plasma control, a stored energy of W tot ∼ 450 kJ has been achieved with a corresponding energy confinement time of τ E ∼ 163 ms. Recent discharges, produced in the fall of 2012, have reached plasma β N up to 2.9 and surpassed the n = 1 ideal no-wall stability limit computed for H-mode pressure profiles, which is one of the key threshold parameters defining advanced tokamak operation. Typical H-mode discharges were operated with a plasma current of 600 kA at a toroidal magnetic field B T = 2 T. L-H transitions were obtained with 0.8-3.0 MW of NB injection power in both single-and double-null configurations, with H-mode durations up to ∼15 s at 600 kA of plasma current. The measured power threshold as a function of lineaveraged density showed a roll-over with a minimum value of ∼0.8 MW at ne ∼ 2×10 19 m −3 . Several edge-localized mode (ELM) control techniques during H-mode were examined with successful results including resonant magnetic perturbation, supersonic molecular beam injection (SMBI), vertical jogging and electron cyclotron current drive injection into the pedestal region. We observed various ELM responses, i.e. suppression or mitigation, depending on the relative phase of in-vessel control coil currents. In particular, with the 90 • phase of the n = 1 RMP as the most resonant configuration, a complete suppression of type-I ELMs was demonstrated. In addition, fast vertical jogging of the plasma column was also observed to be effective in ELM pace-making. SMBI-mitigated ELMs, a state of mitigated ELMs, were sustained for a few tens of ELM periods. A simple cellular automata ('sand-pile') model predicted that shallow deposition near the pedestal foot induced small-sized high-frequency ELMs, leading to the mitigation of large ELMs. In addition to the ELM control experiments, various physics topics were explored focusing on ITER-relevant physics issues such as the alteration of toroidal rotation caused by both electron cyclotron resonance heating (ECRH) and externally applied 3D fields, and the observed rotation drop by ECRH in NB-heated plasmas was investigated in terms of either a reversal of the turbulence-driven residual stress due to the transition of ion temperature gradient to trapped electron mode turbulence or neoclassical toroidal viscosity (NTV) torque by the internal kink mode. The suppression of runaway electrons using massive gas injection of deuterium showed that runaway electrons were avoided only below 3 T in KSTAR. Operation in 2013 is expected to routinely exceed the n = 1 ideal MHD no-wall stability boundary in the long-pulse H-mode ( 10 s) by applying real-time shaping control, enabling n = 1 resistive wall mode active control studies. In addition, intensive works for ELM mitigation, ELM dynamics, toroidal ro...

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Mastering nuclear fusion, which is an abundant, safe, and environmentally competitive energy, is a great challenge for humanity. Tokamak represents one of the most promising paths toward controlled fusion. Obtaining a high-performance, steady-state, and long-pulse plasma regime remains a critical issue. Recently, a big breakthrough in steady-state operation was made on the Experimental Advanced Superconducting Tokamak (EAST). A steady-state plasma with a world-record pulse length of 1056 s was obtained, where the density and the divertor peak heat flux were well controlled, with no core impurity accumulation, and a new high-confinement and self-organizing regime (Super I-mode = I-mode + e-ITB) was discovered and demonstrated. These achievements contribute to the integration of fusion plasma technology and physics, which is essential to operate next-step devices.

High power experiments, up to 9.2 MW with LHCD and ICRH, have been carried out in the full tungsten tokamak WEST. Quasi non inductive discharges have been achieved allowing to extend the plasma duration to 53s with stationary conditions in particular with respect to tungsten contamination. Transitions in H mode are obtained lasting up to 4s with weak energy increment at the power crossing the separatrix is close to the threshold. Hot L mode plasmas (Te(0)>3keV) with a confinement time following the ITER L96 scaling are routinely obtained. The weak aspect ratio dependence of this scaling law is confirmed. Tungsten accumulation is generally not an operational issue on WEST. Difficulty of burning through tungsten can prevent from accessing to a hot core plasma in the ramp-up phase or can lead to rapid collapse of the central temperature when radiation is enhanced by a slight decrease of the temperature. Apart few pulses post-boronization, the plasma radiation is rather high (Prad/Ptot~50%) and is dominated by tungsten. This fraction does not vary as the RF power is ramped up and is quite similar in ICRH and/or LHCD heated plasmas. An estimate of the contribution of the RF antennas to the plasma contamination in tungsten is given

Upgrades to the motional Stark effect (MSE) diagnostic on Alcator C-mod have enabled accurate measurement of the current profile using pitch angle constrained magnetic equilibrium reconstructions. The MSE diagnostic utilizes an intrashot calibration technique along with kinetic profiles to constrain the reconstruction throughout the discharge. The system was used to study the current profile in plasmas with substantial Lower Hybrid Current Drive (LHCD), including fully non-inductive discharges sustained for several current relaxation times. The current profile evolution is reconstructed using the MSE data with 100ms time resolution as LHCD is applied to the plasma and as the plasma current profile relaxes back to inductively driven after LHCD ceases. The LH driven current is observed to be off-axis and significantly broadens the plasma current profile, resulting in q o above 1 and sawtooth suppression. In fully noninductive discharges the reconstructed q profile is flat or non-monotonic. In plasmas with significant LHCD the plasma evolves towards a non-inductive equilibrium but in certain discharges large MHD instabilities coincide with a significant reduction in current drive. The reconstructed driven current profile was measured over a variety of parameters including plasma current, launched n || and LH power. The plasma density was varied over a range where previous work shows the current drive efficiency decreases precipitously from the classic efficiency scaling proportional to 1/n e . As the density is increased the LHCD modification of the current profile is observed to decrease, consistent with decreases in non-thermal ECE emission and hard X-ray production indicating negligible current drive.