The EAST research program aims to demonstrate steady-state long-pulse advanced high-performance H-mode operations with ITER-like poloidal configuration and RF-dominated heating schemes. Since last IAEA FEC, EAST has been upgraded with all ITER-relevant auxiliary heating and current drive systems, enabling the investigation of plasma profile control by coupling/integration of various combinations. By means of the 4.6 GHz and 2.45 GHz LHCD systems, H-mode can be obtained and maintained at relatively high density, even up to n e ~ 4.5 × 10 19 m-3 , where a current drive effect is still observed. Significant progress has been achieved on EAST, including: i). Demonstration of a steady-state scenario (fully non-inductive with V loop ~ 0.0V at high β P ~ 1.8 and high performance (H 98,y2 > 1.0) in upper single-null (ε ~ 1.6) configuration with the tungsten divertor; ii) Discovery of a stationary ELM-stable H-mode regime with 4.6 GHz LHCD; iii) achievement of ELM suppression in slowly-rotating H-mode plasma with the application of n = 1 and 2 RMPs.

The recent EAST experimental progress since the last IAEA FEC in 2016 is presented. First demonstration of >100 seconds time scale long-pulse steady-state scenario with a good plasma performance (H98(y2) ~ 1.1) and a good control of impurity and heat exhaust with the tungsten divertor has been successfully achieved on EAST using the pure RF power heating and current drive. The extended operation regimes have been obtained (βP~2.5 & βN~1.9 of using RF&NB and βP~1.9 & βN~1.5). High bootstrap current fraction up to 47% was achieved with q95~6.0-7.0. The interaction effect between the ECH and two LHW systems has been investigated for enhanced current drive and improved confinement quality. ELM suppression using the n= 2 RMPs has been achieved at q95 (≈ 3.2-3.7) with standard type-I ELMy H-mode operational window in EAST. Reduction of the peak heat flux on the divertor was demonstrated using the active radiation feedback control. An increase in the total heating power and improvement of the plasma confinement are expected using a 0-D model prediction for higher bootstrap fraction. Towards long pulse, high bootstrap current fraction operation, a new lower ITER-like tungsten divertor with active watercooling will be installed, together with further increase and improvement of heating and current drive capability.

A power-balance model, with radiation losses from impurities and neutrals, gives a unified description of the density limit (DL) of the stellarator, the L-mode tokamak, and the reversed field pinch (RFP). The model predicts a Sudo-like scaling for the stellarator, a Greenwald-like scaling, , for the RFP and the ohmic tokamak, a mixed scaling, , for the additionally heated L-mode tokamak. In a previous paper (Zanca et al 2017 Nucl. Fusion 57 056010) the model was compared with ohmic tokamak, RFP and stellarator experiments. Here, we address the issue of the DL dependence on heating power in the L-mode tokamak. Experimental data from high-density disrupted L-mode discharges performed at JET, as well as in other machines, are taken as a term of comparison. The model fits the observed maximum densities better than the pure Greenwald limit.

The internal transport barrier (ITB) has been obtained in ELMy H-mode plasmas by neutron beam injection and lower hybrid wave heating on the Experimental Advanced Superconducting Tokamak (EAST). The ITB structure has been observed in profiles of ion temperature, electron temperature, and electron density within ρ<0.5. It was also observed that the ITB formation is stepwise. Due to the ITB formation, the confinement quality H 98y2 increases from 1 to 1.1 and the normalized beta, β N , increases from 1.5 to near 2. The fishbone activity observed during the ITB phase suggests the central safety factor q(0)∼1. Transport coefficients are calculated by particle balance and power balance analysis, showing an obvious reduction after the ITB formation.

Aimed at high-confinement (H-mode) plasmas in the Experimental Advanced Superconducting Tokamak (EAST), the effect of local gas puffing from electron and ion sides of a lower hybrid wave (LHW) antenna on LHW–plasma coupling and high-density experiments with lower hybrid current drive (LHCD) are investigated in EAST. Experimental results show that gas puffing from the electron side is more favourable to improve coupling compared with gas puffing from the ion side. Investigations indicate that LHW–plasma coupling without gas puffing is affected by the density near the LHW grill (grill density), hence leading to multi-transition of low–high–low (L–H–L) confinement, with a correspondingly periodic characteristic behaviour in the plasma radiation. High-density experiments with LHCD suggest that strong lithiation gives a significant improvement on current drive efficiency in the higher density region than 2 × 1019 m−3. Studies indicate that the sharp decrease in current drive efficiency is mainly correlated with parametric decay instability. Using lithium coating and gas puffing from the electron side of the LHW antenna, an H-mode plasma is obtained by LHCD in a wide range of parameters, whether LHW is deposited inside the half-minor radius or not, implying that a central and large driven current is not a necessary condition for the H-mode plasma. H-mode is investigated with CRONOS.

First principle modeling of the lower hybrid (LH) current drive in tokamak plasmas is a longstanding activity, which is gradually gaining in accuracy thanks to quantitative comparisons with experimental observations. The ability to reproduce simulatenously the plasma current and the non-thermal bremsstrahlung radial profiles in the hard x-ray (HXR) photon energy range represents in this context a significant achievement. Though subject to limitations, ray tracing calculations are commonly used for describing wave propagation in conjunction with Fokker-Planck codes, as it can capture prominent features of the LH wave dynamics in a tokamak plasma-like toroidal refraction. This tool has been validated on several machines when the full absorption of the LH wave requires the transfer of a small fraction of power from the main lobes of the launched power spectrum to a tail at a higher parallel refractive index. Conversely, standard modeling based on toroidal refraction only becomes more challenging when the spectral gap is large, except if other physical mechanisms may dominate to bridge it, like parametric instabilities, as suggested for JET LH discharges (Cesario et al 2004 Phys. Rev. Lett. 92 175002), or fast fluctuations of the launched power spectrum or 'tail' LH model, as shown for Tore Supra (Decker et al 2014 Phys. Plasma 21 092504). The applicability of the heuristic 'tail' LH model is investigated for a broader range of plasma parameters as compared to the Tore Supra study and with different LH wave characteristics. Discrepancies and agreements between simulations and experiments depending upon the different models used are discussed. The existence of a 'tail' in the launched power spectrum significantly improves the agreement between modeling and experiments in plasma conditions for which the spectral gap is large in EAST and Alcator C-Mod tokamaks. For the Alcator C-Mod tokamak, the experimental evolution of the HXR profiles with density suggests that this model is valid up to a line-averaged density of × + n 1.0 10 e 20 m −3 , a statement that is confirmed by simulations of the HXR scaling law with density. While simulations with GENRAY/CQL3D codes have ascribed the fast decrease of the HXR emission with density to parasitic absorption in the scrape-off layer by collisional damping, an alternative interpretetation based on an enhanced refraction as the LH wave propagates in the vicinity of the X-point is provided by C3PO/LUKE codes. The consequences for the predictions of LH current in ITER are discussed.

A 4.6 GHz lower-hybrid current drive (LHCD) system has been firstly commissioned in EAST in the 2014 campaign. The first LHCD results with 4.6 GHz show that LHW can be coupled to plasma with a low reflection coefficient, drive plasma current and plasma rotation, modify the plasma current profile, and heat plasma effectively. By means of configuration optimization and local gas puffing near the LHW antenna, good LHW-plasma coupling with a reflection coefficient less than 5% is obtained. The maximum LHW power coupled to plasma is up to 3.5 MW. The current drive (CD) efficiency is up to 1.1 × 10 19 A m −2 W −1 and the central electron temperature is above 4 keV, suggesting that LH power could be mainly deposited in the core region, which is in agreement with code simulation. Experiments show that the current profile is effectively modified and toroidal rotation in the co-current direction is driven by the LHCD. Also, the CD efficiency and current profile depend on the launched wave spectrum, suggesting the possibility of controlling the current profile by changing the phase difference. Repeatable H-mode plasma is obtained by either the 4.6 GHz LHCD system alone, or together with a 2.45 GHz LHCD system, the NBI (neutral beam injection) system. The different ELM features of H-mode between the different heating methods are under investigation.

Using a 2 MW 2.45 GHz lower hybrid wave (LHW) system installed in experimental advanced superconducting tokamak, we have systematically carried out LHW-plasma coupling and lower hybrid current drive experiments in both divertor (double null and lower single null) and limiter plasma configuration with plasma current (I p ) $ 250 kA and central line averaged density (n e ) $ 1.0-1.3 Â 10 19 m À3 recently. Results show that the reflection coefficient (RC) first is flat up to some distance between plasma and LHW grill, and then increases with the distance. Studies indicate that with the same plasma parameters, the best coupling is obtained in the limiter case (with plasma leaning on the inner wall), followed by the lower single null, and the one with the worst coupling is the double null configuration, explained by different magnetic connection length. The RCs in the different poloidal rows show that they have different coupling characteristics, possibly due to local magnetic connection length. Current drive efficiency has been investigated by a least squares fit with N peak == ¼ 2:1, where N peak == is the peak value of parallel refractive index of the launched wave. Results show that there is no obvious difference in the current drive efficiency between double null and lower single null cases, whereas the efficiency is somewhat small in the limiter configuration. This is in agreement with the ray tracing=Fokker-Planck code simulation by LUKE=C3PO and can be interpreted by the power spectrum up-shift factor in different plasma configurations. A transformer recharge is realized with $0.8 MW LHW power and the energy conversion efficiency from LHW to poloidal field energy is about 2%.