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.

A long pulse electron cyclotron resonance heating (ECRH) system has been developed to meet the requirements of steady-state operation for the EAST superconducting tokamak, and the first EC wave was successfully injected into plasma during the 2015 spring campaign. The system is mainly composed of four 140 GHz gyrotron systems, 4 ITER-Like transmission lines, 4 independent channel launchers and corresponding power supplies, a water cooling, control & inter-lock system etc. Each gyrotron is expected to deliver a maximum power of 1 MW and be operated at 100-1000 s pulse lengths. The No.1 and No.2 gyrotron systems have been installed. In the initial commissioning, a series of parameters of 1 MW 1 s, 900 kW 10 s, 800 kW 95 s and 650 kW 753 s have been demonstrated successfully on the No.1 gyrotron system based on calorimetric dummy load measurements. Significant plasma heating and MHD instability suppression effects were observed in EAST experiments. In addition, high confinement (H-mode) discharges triggered by ECRH were obtained.

Recent EAST experiment has successfully demonstrated long pulse steady-state high plasma performance scenario and core-edge integration since the last IAEA in 2018. A discharge with a duration over 60s with βP ~2.0, βN ~1.6, H98y2~1.3 and internal transport barrier on electron temperature channel is obtained with multi-RF power heating and current drive. A higher βN (βN~1.8, βp~2.0, H98y2~1.3, ne/nGW~0.75) with a duration of 20s is achieved by using the modulated neutral beam and multi-RF power, where several normalized parameters are close or even higher than the phase III 1GW scenario of CFETR steady-state. High-Z impurity accumulation in the plasma core is well controlled in a low level by using the on-axis ECH. Modelling shows that the strong diffusion of TEM turbulence in the central region prevents tungsten impurity to accumulate. More recently, EAST has demonstrated compatible core-edge integration discharges in the high βp scenario: high confinement H98y2>1.2 with high βP~2.5/βN~2.0 and fbs~50% is sustained with reduced divertor heat flux at high density ne/nGW~0.7 and moderate q95~6.7. By combining active impurity seeding through radiative divertor feedback control and strike point splitting induced by resonant perturbation coil, the peak heat flux is reduced by 20-30% on the ITER-like tungsten divertor, here a mixture of 50% neon and 50% D2 is applied.

Ion cyclotron emissions (ICE) driven by high energy ions in the Experimental Advanced Superconducting Tokamak (EAST) are reported. The classic edge emissions driven by energetic neutral beam injection (NBI) deuterium ions at low magnetic field side, near the plasma pedestal region, are often observed. In addition, the ICE, the spectral peaks of which match the cyclotron frequencies of fusion ions and NBI deuterium ions near the magnetic axis are also detected. In ICE experiments, deuterium plasma is heated with deuterium NBI, without any ion cyclotron resonance heating. In H-mode discharges, edge localized mode can not only increase the intensity of the edge ICE, but also broaden its spectrum. ICE also appears at the time of plasma disruption in the EAST tokamak, this radiation seems to be driven by deuterium ions at the radial location of R  =  1.85 m–2.05 m. These results lead to a practical way to monitor fusion alpha particles in deuterium–tritium devices such as ITER, DEMO and CFETR in the future.

The auxiliary heating power absorption of 2.45 GHz and 4.6 GHz lower hybrid (LH) waves in EAST is analysed in order to improve the determination of the energy confinement time. The LH power absorption coefficient is obtained by using the time derivative of the total stored energy including the plasma kinetic energy and the poloidal magnetic field energy. Results on EAST have shown that, for both frequencies, the fraction of LH power absorbed in the plasma bulk decreases with the increase in the plasma density, and with the decrease in the toroidal magnetic field. Scaling laws for the density dependence of the LH absorption coefficient at different fields and frequencies have been statistically obtained. Comparison of the confinement improvement factor H89 obtained with these new scaling laws against those assuming an absorption coefficient equal to the launched k//wave spectrum directivity is made.

Upstream density profiles in the scrape-off layer (SOL) have been examined in low-confinement mode (L-mode) and high-confinement mode (H-mode) plasmas in the EAST superconducting tokamak. A weak density shoulder forms in the near SOL region in upper single-null configurations when the neutral pressure measured at the lower divertor exceeds a threshold value of 2 × 10−2 Pa in L-mode plasmas. When the neutral pressure is below this threshold, the weak density shoulder is absent and the sidebands of the lower hybrid waves associated with SOL parametric instabilities are reduced. Active detachment control with neon–deuterium seeding demonstrate that the weak density shoulder can form before the onset of the outer divertor detachment as long as the neutral pressure is above the threshold. Furthermore, no remarkable expansion of a shoulder is observed during divertor detachment, suggesting that divertor detachment is not a necessary condition for the formation or growth of a density shoulder. Through the increase in neutral pressure in the lower divertor by an order of magnitude, the weak shoulder was observed to expand into the far SOL and reach the leading edge of the limiter. The results in L-mode discharges identified the neutral pressure in the lower divertor as a primary factor for the formation of an SOL density shoulder in the upper single-null discharges. For the type-I ELMy H-mode plasmas, a similar density shoulder was detected during the inter-ELM phase when the neutral pressure in the lower divertor exceeded a threshold value of 4 × 10−2 Pa. On the other hand, the shoulder was absent when the divertor neutral pressure went below this threshold even though the plasma discharge was conducted with a higher core line-averaged density and divertor collisionality. This is consistent with the observations in L-mode plasmas. The neutral particle ionization of the working gas is thus believed to play a key role during the formation of the SOL density shoulder in the EAST tokamak.

Significant improvement of plasma performance in high-confinement mode (H-mode) discharges with favourable toroidal field B t, i.e. the ion ∇B drift towards the primary X-point, has been widely observed in the EAST tokamak with pure radio-frequency heating in contrast to that with the unfavourable B t. Statistical analysis indicates that plasma in the favourable B t has higher core electron temperature, similar core ion temperature and relatively steeper pedestal density compared with that in the unfavourable B t. The improvement in plasma performance is mainly contributed by the increase of core electron temperature in the favourable B t. Further analysis indicates that the plasma with favourable B t has much lower density and recycling in the scrape-off layer (SOL). Lower SOL density and recycling benefit the mitigation of parametric instability activity of lower hybrid wave (LHW), and thus facilitate the increase of core electron temperature in the favourable B t. The performance improvement in the favourable B t demonstrates to be more evident with high LHW power. Divertor local E r × B drift which can increase the backflow particles from the divertor region to the upstream region could be partly responsible for the much higher SOL plasma density in unfavourable B t. These findings could facilitate the application of LHW power on future large fusion devices, such as the China Fusion Engineering Test Reactor, to achieve high-performance steady-state operation.

Lower hybrid wave (LHW)-plasma coupling and lower hybrid current drive (LHCD) experiments in divertor, including single-null and double-null, and limiter configurations were conducted systematically in EAST. A maximum power for launched LHW is 1.4 MW and the plasma current with LHCD is about 1 MA. It is indicated that the coupling is best in limiter configuration, then in single-null one, while worst in double-null one. Study in current drive efficiency by a least squares fit shows that there is no obvious difference in drive efficiency between the double-null and the single-null cases, whereas the efficiency is a slightly lower in the limiter case. The effect of plasma density on the current drive efficiency is due to the influence of density on impurity concentration.