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Abstract Experimental Advanced Superconducting Tokamak (EAST) was built to demonstrate high-power, long-pulse operations under fusion-relevant conditions, with major radius R = 1.9 m, minor radius a = 0.5 m, and design pulse length up to 1000s. It has an ITER-like D-shaped cross-section with two symmetric divertors at the top and bottom, accommodating both single null and double null divertor configurations. EAST construction was started in 2000, and its first plasma was successfully obtained in 2006. In the past 15 years, plasma-facing components, plasma heating, diagnostics, and other systems have been upgraded step by step to meet its mission on exploring of the scientific and technological bases for fusion reactors and studying the physics and engineering technology issues with long pulse steady-state operation. An advanced steady-state plasma operation scenario has been developed, and plasma parameters were greatly improved. Meanwhile, front physics on the magnetic confinement plasmas have been systemically investigated and lots of fruitful results were realized, covering transport and confinement, MHD stabilities, pedestal physics, divertor and scrap-off layer (SOL) physics, and energetic particle physics. This brief review of EAST on engineering upgrading, stand-steady operation scenario development, and plasma physics investigation would be useful for the reference on construction and operation of a superconducting tokamak, such as ITER and future fusion reactor.

Since the last IAEA-FEC in 2021, significant progress on the development of long pulse steady state scenario and its related key physics and technologies have been achieved, including the reproducible 403 s long-pulse steady-state H-mode plasma with pure radio frequency (RF) power heating. A thousand-second time scale (∼1056 s) fully non-inductive plasma with high injected energy up to 1.73 GJ has also been achieved. The EAST operational regime of high β _P has been significantly extended ( H _98y2 > 1.3, β _P ∼ 4.0, β _N ∼ 2.4 and n _e / n _GW ∼ 1.0) using RF and neutral beam injection (NBI). The full edge localized mode suppression using the n = 4 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with q _95 ≈ 3.1 on EAST, extrapolating favorably to the ITER baseline scenario. The sustained large ELM control and stable partial detachment have been achieved with Ne seeding. The underlying physics of plasma-beta effect for error field penetration, where toroidal effect dominates, is disclosed by comparing the results in cylindrical theory and MARS-Q simulation in EAST. Breakdown and plasma initiation at low toroidal electric fields (

Tungsten (W) impurity control is critical for plasma performance and a priority for ITER. The simultaneous reduction of W and rotation in the core region induced by resonant magnetic perturbations (RMPs) has been found and understood in EAST. A positive feedback loop between the W and rotation is first proposed, resulting in core W accumulation and high rotation even in low-torque plasma before the RMP application. This cycle can be reversed by the edge rotation braking induced by RMP, causing a significant simultaneous reduction of W concentration and rotation. These new mechanisms are based on several repeatable experiments and confirmed by the modeling results from TGYRO and NTVTOK. It provides a new understanding of the RMP effects on W and rotation and can be used for W and rotation control in future reactors.

Recent experimental results from the EAST superconducting tokamak with a full metal wall (without Li-coating) provide strong support for the ITER startup scheme with electron cyclotron wave (ECW) assistance. Detailed investigations were performed on the impact of different magnetic field geometries, including quadrupolar field configuration (QFC), wide null field configuration (NFC), and trapped particle configuration (TPC). All three configurations enabled the breakdown of neutral gas with ECW pre-ionization. However, in most discharges with QFC, the absence of closed flux surfaces after the breakdown was associated with unsuccessful plasma startups. Conversely, both NFC and TPC achieved successful plasma startups under normal conditions, with TPC demonstrating higher robustness against high impurity levels compared to NFC. Experiments found that the minimum ECW power required for a successful TPC startup is approximately 0.48 MW. Plasma startup failed if the prefill gas pressure exceeded optimal levels for a given ECW power ( $ \unicode{x2A7E} {\text{0}}{\text{.48}}$ MW), with the critical gas pressure linearly dependent on the ECW power. Additionally, higher ECW power was required with increased impurity levels. With ECW assistance, the toroidal electric field at $R = 1.85{\text{ m}}$ in EAST routine operations ( ${{ < }}0.15{\text{ V}}\,{{\text{m}}^{ - 1}}$ ) is significantly lower than ITER’s maximum value (0.3 V m ^−1 ), suggesting the possibility of reducing flux consumption and lowering demands on coils and power supplies for ITER.

Recent results of density ramp-up (< n _e >/ n _GW = 0.5–0.75) experiment in type-I ELMy H-mode on experimental advanced superconducting tokamak are presented, with a focus on pedestal dynamics and turbulence behavior. With the density increase, a general trend is that the pedestal electron pressure ( p _e ^ped ) decreases while the ELM frequency ( f _ELM ) increases. Especially when < n _e >/ n _GW arrives at ∼ 0.68, a sudden decrease of p _e ^ped and an increase of f _ELM are observed. A quasi-coherent mode (QCM) with a frequency of 200–300 kHz is observed in the pedestal region and appears in both density and magnetic fluctuations. The time evolution of p _e ^ped during the inter-ELM phase is analyzed for different densities. The analysis suggests that the pedestal behavior for < n _e >/ n _GW < 0.68 is mainly attributed to pedestal stability but not due to pedestal transport. It is shown that the resistivity effect on the peeling-ballooning (PB) mode (Zhang et al 2017 Phys. Plasmas 24 062108) could be used to explain the present result. But the sudden decrease of p _e ^ped and increase of f _ELM at < n _e >/ n _GW > 0.68 cannot be attributed to the resistivity effect. It is found that for the plasma with < n _e >/ n _GW > 0.68, the QCM disappears just before ELM and the density pedestal becomes steeper. This implies that the QCM can drive outward particle transport. The steeper density pedestal leads to a narrower pedestal width. A narrower pedestal width will lead to a lower pedestal pressure since PB mode limits the pressure gradient and could explain the observed sudden decrease of p _e ^ped and increase of f _ELM .

The edge coherent mode (ECM) is considered a highly attractive pedestal mode because it extends the duration of edge-localized modes and increases particle and impurity transport without significantly affecting energy transport. Moreover, it operates compatibly with high-performance plasma discharges. The ECM can also be detected using Langmuir probes on the divertor target plate, indicating that it extends from the pedestal region into the SOL and resulting in a connection to the divertor target plate via magnetic field lines. In this work, the distribution of ECM on divertor target plates is investigated by analyzing 215 upper single null discharges on the EAST tokamak. The coherence analysis of plasma fluctuations between the electron cyclotron emission signal in the pedestal region of the outer midplane and the ion saturation current measured by Langmuir probes in the divertor region reveals that the ECM is hardly detected by the divertor probe close to the outer strike point but can be observed at far SOL. This finding indicates the presence of an ECM quiescent region near the SOL on the divertor plate, and the extent of the quiescent region in the poloidal flux coordinate (Δ) has been statistically analyzed. A pronounced relationship between Δ and triangularity ( δ ) has been observed, that is Δ increasing with δ . Further analysis reveals that this relationship can be attributed to the average magnetic shear in the SOL. This result is consistent with the physical picture, which states that strong magnetic shear close to X -point significantly squeezes the cross-section of flux tubes down to scales dominated by collisions.

I-mode is a promising regime for future fusion reactors due to the high energy confinement and the moderate particle confinement. However, the effect of lithium, which has been widely applied for particle recycling and impurity control, on I-mode plasma is still unclear. Recently, experiments of real-time lithium powder injection on I-mode plasma have been carried out in EAST Tokamak. It was found that the confinement performance of the I-mode can be improved by the lithium powder injection, which can strongly reduce electron turbulence (ET) and then trigger ion turbulence (IT). And it was observed that the ET intensity is inversely proportional to the velocity shear, which suggests that the injection of lithium powder leads to a gradual enhancement of the shear flow, whereby the turbulence is reduced and consequently the confinement is improved. Four different regimes of I-mode have been identified in EAST. The Type I I-mode plasma is characterized by the weakly coherent mode (WCM) and the geodesic-acoustic mode (GAM). The Type II I-mode is featured as the WCM and the edge temperature ring oscillation (ETRO). The Type III I-mode corresponds to the plasma with the co-existence of ETRO, GAM, and WCM. The Type IV I-mode denotes the plasma with only WCM but without ETRO and GAM. It was observed that the WCM intensity is increased with lithium powder injection by the confinement improvement/pedestal temperature increase. EAST experiments demonstrate that lithium powder injection is an effective tool for real-time control and confinement improvement of I-mode plasma.

Plasma electron density is a crucial parameter in plasma studies. Accurately inverting the plasma electron density profile is vital for plasma control experiments and the investigation of plasma physical mechanisms. This paper proposes an integrated data analysis (IDA) method based on Bayesian inference, which integrates polarimetric interferometry, hydrogen cyanide laser interferometer, and microwave reflectometer diagnostics for inverting the plasma electron density profile. To enhance inversion accuracy, a Gaussian prior probability of the non-stationary hyperparameter is used. This prior probability effectively simulates situations where there is a large plasma electron density gradient in the pedestal, especially under the condition of high-confinement mode discharge. Compared to the use of Gaussian prior probability for the stationary hyperparameter, the proposed IDA method based on the non-stationary hyperparameter prior probability achieves higher inversion accuracy.

In the pedestal region of the Experimental Advanced Superconducting Tokamak (EAST) during high confinement mode plasma operations with radio-frequency heating, two distinct fluctuations are observed: high-frequency fluctuations (HFFs) and edge harmonic oscillation-like (EHO-like) modes. The HFFs are characterized by intermittent fluctuations with a broadband frequency range of $1{-}3{\text{ MHz}}$ and a poloidal wave number ( ${k_\theta }$ ) greater than $0.9{\text{ c}}{{\text{m}}^{ - 1}}$ . On the other hand, the EHO-like mode exhibits characteristics similar to magnetohydrodynamics (MHD)-like modes with n = 1−5 and lower poloidal wave numbers ( ${{\text{k}}_{{\theta }}} \unicode{x2A7D} 0.12{\text{ c}}{{\text{m}}^{ - 1}}$ ). During the pedestal establishing phase following the L–H transition, a significant concurrent presence of HFF and EHO-like modes in high-density pedestal regions has been noted. In this phase, the EHO-like mode not only modulates the amplitude of the HFF but also engages in nonlinear interactions. The occurrence of EHO-like mode and HFF is associated with particle transport toward the divertor, though it is notably less than that caused by edge coherent modes. During the inter-edge localized mode (ELM) period, a significant decrease in the ${D_\alpha }$ baseline is observed whenever the low frequency fluctuation (LFF) weakens and the HFF grows, prior to each large ELM. One possible explanation is that the rapid increase of $E \times B$ shear stabilizes the LFF and destabilizes the HFF, which lowers the pedestal transport and enables the further growth of the pedestal until the onset of the ELM.

In the EAST tokamak, we have developed an internal transport barrier (ITB) high-confinement mode (H-mode) scenario characterized by dominant electron heating and centrally peaked electron temperature profiles, facilitated primarily through the combustion of lower hybrid current drive and electron cyclotron radio heating (ECRH). Hard x-ray diagnostics reveal a marked increase in the population of fast electrons with energy from 30 keV to 80 keV, concurrent with augmented ECRH power during H-mode plasma operations. This surge in fast electron population precedes the formation of the electron temperature ITB (Te-ITB). Within the Te-ITB H-mode discharge, a mild and long-lived m / n = 1/1 mode (where m and n denote the toroidal and poloidal mode numbers, respectively) emerges proximal to the ITB region. This mode precipitates a redistribution of fast electrons, contributing to an increase in the safety factor near the magnetic axis and thereby promoting the stability of the Te-ITB. Furthermore, we explore the influence of fast electrons on plasma pressure and examine the effects of the profile of fast electrons on the central Te. Strategies to maintain the m / n = 1/1 mode at a moderate amplitude are also discussed, highlighting their significance in the sustained management of Te-ITB.

Scrape-off layer (SOL) profiles and turbulence in ion cyclotron range of frequency (ICRF)-heated plasmas are investigated by the reciprocating probe diagnostic system (FRPs) and gas puff imaging (GPI) diagnostic in EAST. A radio-frequency (RF) sheath potential reaching up to 100 V is identified proximate to the ICRF antennas. Notably, the amplitude of this RF sheath potential escalates in response to rising ICRF power and inversely with plasma density. When a RF sheath is present in the far SOL, a pronounced density ‘shoulder’ forms in front of the ICRF antennas, while the ‘shoulder’ fade away as the antenna and associated RF sheath shift outwards. A strong E _r shear is revealed by measurements from both FRPs and GPI. Analysis of the poloidal wave number-frequency spectrum reveals suppression of high-frequency turbulence in the far SOL due to the RF sheath. This effect is manifested in the reduced autocorrelation time τ _c and reduced average blob size δ _blob of the SOL plasma. Intriguingly, the poloidal propagation direction of the low-frequency turbulence reverses from the electron to the ion diamagnetic drift direction at the RF sheath location. A surge of tungsten impurity is potentially attributed to the heightened interaction between the SOL plasmas and the wall material. Shifting the ICRF antennas outward, to alleviate heat spots, results in the relocation of the RF sheath to the shaded region of the main limiter. This shift amplifies the radial velocity of blobs in the far SOL and concurrently diminishes the SOL density when compared to conditions without ICRF injection. The properties of ion saturation current fluctuations are consistent with the stochastic model predictions.

Effective management of hydrogen isotopes retention in plasma-facing materials (PFMs) is crucial, particularly when utilizing tritium (T) as fuel, for the success of burning plasma operations. Boronization, a widely employed technique for controlling fuel recycling and mitigating impurity influx from plasma-surface interactions into the core of burning plasma, significantly influences hydrogen isotopes retention in PFMs. In this study, boronization films were generated by ion cyclotron range of frequency (ICRF) discharge assisted boronization with C _2 B _10 H _12 as boron source on tungsten substrates at Experimental Advanced Superconducting Tokamak (EAST) which employing ITER-like water-cooled W monoblock PFMs and components (PFCs), followed by in-situ glow discharge (GD) cleaning for 2 h and 20 shots (180 s) edge-plasma exposure. Employing the Material and Plasma Evaluation System (MAPES), representative samples were analyzed after each process. The resultant carbon–boron films, dense and continuous, exhibited thickness up to 120 nm and were identified as amorphous in structure. It was observed that the D _2 -GD cleaning resulted in a significant isotopic exchange effect, effectively reducing the hydrogen (H) retention in the carbon–boron films. This hydrogen isotope replacement efficiency was found to be influenced by the thickness of the films. Notably, after boundary plasma exposure, samples with thicker films demonstrated an enhanced capacity to capture deuterium (D), adsorbing 10 times more D than bare tungsten (W). Our findings offer transformative insights for T recycling analysis and the plasma operation of devices like International Thermonuclear Experimental Reactor (ITER), highlighting the impact of boronization and subsequent treatments on hydrogen isotope retention behavior in PFMs.

To sustain good lower hybrid wave (LHW)–plasma coupling for long-pulse plasma operation, for the first time, coupling feedback control is designed and realized in EAST using a proportion integration differentiation method by choosing the reflection coefficient (RC) of LHW power as the reference for gas-puffing feedback, and including one pulse test and multi-pulse experiments. Experiments show that such feedback control can work correctly and maintains good LHW–plasma coupling effectively for a long time, suggesting the possibility of feedback control application on LHW–plasma coupling in long-pulse plasma. Furthermore, during the feedback control process of multi-pulse supersonic molecular beam injection (SMBI), the stored energy changes from 29 kJ to 58 kJ, and the energy confinement factor ( H _89 ) increases from 0.98 to 1.45, implying a positive effect of coupling feedback on plasma performance. Experiments between SMBI puffing and the gas puffing system, fed by a piezoelectric valve near the antenna, are further investigated, showing that the response time of the RC with SMBI is faster than that by the piezoelectric valve. In addition, SMBI puffing on the electron-drift side of the LHW antenna is a little quicker than that on the ion-drift side. Studies suggest that such feedback control is effective for long-pulse LHW–plasma coupling, and the gas puffing by SMBI on the electron-drift side of the LHW antenna could offer an effective way to sustain good LHW coupling in steady-state operation in the future. Further optimization will be continued at a later date.

In the ITER and future fusion devices, W/Cu monoblocks will be used as divertor target which are exposed to both steady state heat load and transient heat flux. Especially, the transient heat flux up to 10 GW m ^−2 during plasma disruption, is expected to induce the shallow surface damages, such as melting, and even boiling of W/Cu monoblocks. Thus, the performance of damaged W/Cu monoblocks under subsequent long-term plasma discharges is a key concern that needs to be verified and tested on existing tokamaks. Since 2022, a new type of main limiter composed of ITER-like W/Cu monoblocks has been installed and tested in EAST. The surface of W/Cu monoblocks of the limiter was damaged by the transient heat flux during the early stages of plasma construction. Subsequently, they were subjected to long-term plasma discharges over 2600 shots in normal plasma discharge conditions. This circumstance conveniently facilitates the discussion of the performance of W/Cu monoblocks with damaged surfaces especially a melting edge with hill structure under prolonged exposure to plasma. In general, the shallow damage resulting from transient heat flux on W/Cu monoblocks appears to have minimal impact on the heat exhaust capacity under steady-state heat loads, as indicated by both experimental monitoring and numerical simulation results. However, shallow melting, leading to a change in surface structure and the formation of hills, could theoretically increase local temperatures, creating potential hot spots. This phenomenon requires further validation through dedicated experiments. Moreover, the brittleness of the near-surface layer may give rise to brittle destructions, such as cracks and even dust particles, posing an additional concern. These findings yield unique qualitative conclusions that can be referenced for ITER and other fusion devices.

The stabilization of the m / n = 2/1 neoclassical tearing mode (NTM) by electron cyclotron current drive (ECCD) has been carried out in EAST H-mode discharges, where m / n is the poloidal/toroidal mode number. The experimental results are reported for the first time in this paper. To facilitate the experimental study, the magnetic island (NTM) is generated by a sufficiently large amplitude of the externally applied resonant magnetic perturbation (RMP). After switching off the RMP, the NTM exists due to the bootstrap current perturbation, with the magnetic island width being about 5 cm for the local equilibrium bootstrap current fraction being larger than 10%. By applying the localized ECCD later, the NTM is fully suppressed if the radial misalignment between the magnetic island and the ECCD location is sufficiently small. The stabilizing effect depends on both the radial misalignment and the applied electron cyclotron wave power. More importantly, it is found that the NTM can be avoided when applying ECCD earlier during the ramp-up phase of the RMP amplitude, if ECCD is localized around the O-point of the magnetic island, indicating an efficient way for avoiding locked modes that can lead to the major disruptions of tokamak plasmas.

Magnetic field measurement is the basic diagnosis to obtain the physical engineering parameters of magnetic confinement fusion device and the macro information of plasma discharge. The real-time diagnosis of magnetic field distribution near the plasma-facing components (PFCs) surface provides the important information on the migration and transport model of key elements. In this work, a remote, in-situ approach for the magnetic field measurement near the surface of PFCs by the polarization-resolved laser-induced breakdown spectroscopy (LIBS) based on Zeeman effect is proposed and implemented. The Zeeman characteristics of the emission spectra of laser-induced W, Mo and C plasmas were verified in the laboratory by using different magnetic field configurations. According to the polarization characteristics of the Zeeman sublines of the LIBS spectrum, the intensity and direction of the external local magnetic field were successively identified by using a linear polarizer. Subsequently, a linear array fiber was utilized to determine the polarity of the external magnetic field. And finally, the magnetic field intensity near the lower edge surface of the tungsten baffle of the Experimental Advanced Superconducting Tokamak (EAST) upper divertor was measured when the field coils were demagnetized. This method can supplement the experimental data near the PFCs for the magnetic field configuration of the magnetic confinement fusion device and provide a reference for the wall element analysis model diagnosed by LIBS.

Experiments at EAST demonstrate effective modulation of the stationary heat flux to the secondary lobes of the magnetic footprint induced by the resonant magnetic perturbations (RMPs) by slightly varying the equilibrium q _95 , consistent with prior numerical modeling. During the small q _95 variation, the edge localized mode control is well maintained, and the position of the secondary heat flux peak is effectively shifted, thus avoiding a specific location heat flux accumulation. As the divertor heat load is one of the significant concerns in tokamaks, these results provide a promising choice, varying magnetic equilibrium periodically to shift stationary heat load deposition position during static n = 4 ( n is the toroidal mode number) RMP condition, for further fusion devices. In this respect, the use of this technique for n = 4 RMPs is advantageous because the q _95 range that needs to be covered to spread the divertor heat load is reduced due to the smaller toroidal extent of the off-separatrix heat deposition zones compared to lower n ’s.

A type of actively cooled W/Cu flat-type component with high heat exhaust capacity was installed as limiter during the EAST spring plasma campaign in 2022, aiming to support the long pulse operation. Unfortunately, severe melting phenomena with obvious droplets ejection of the flat-type W/Cu limiter was repeatedly monitored by CCD and IR cameras, which not only induced the failure of component but also seriously influenced plasma operation. The high temperature around midplane of W/Cu flat-type main limiter is identified to be closely connected with ICRF induced fast ions loss. Indeed, the surface temperature of W/Cu flat-type main limiter was too high to ignore. The damage of flat-type structure, however, usually started between the joint interface of W plates and CuCrZr heat sink material. Such damage would in turn lead to the gradual increase of surface temperature which eventually would cause melting of W plates. Once melting events occurred on the W/Cu flat-type main limiter, the vast majority of cases would result in plasma disruption. Moreover, the damage of main limiter would rapidly deteriorate. Hence, more attention should be paid to how to improve the fatigue lifetime of the joint interface. Such test results of flat-type W/Cu component for limiter are important references for the improvement and application of W/Cu flat-type component for high heat flux area in fusion devices.

Efficient ion cyclotron range of frequencies (ICRF) wave heating requires good wave coupling at the plasma edge and good radio frequency power absorption in the plasma core. This study reviews recent progress in improving these two aspects of ICRF heating with the new two-strap antennas through various experiments and simulations on the Experimental Advanced Superconducting Tokamak (EAST). Our study shows that the ICRF coupling can be significantly improved by decreasing the parallel wave number, increasing the local scrape-off layer (SOL) density by midplane gas puffing, and increasing the global SOL density by decreasing the separatrix–antenna distance. It can also be improved by increasing the core plasma density, changing the divertor strike point position, and optimizing the antenna phasing. The core ICRF power absorption can be increased by optimizing the cyclotron resonance position and minority ion concentration and by applying new heating schemes such as three-ion heating. Although some of the methods have been previously studied on other devices, improving ICRF coupling by shifting the divertor strike point was tested on EAST for the first time. Quantitative characterization of these methods and the conclusions drawn from this study can provide important insights for achieving more efficient ICRF heating in current and future fusion machines.

The operational space and global performance of plasmas with edge-localized modes (ELMs) suppressed by resonant magnetic perturbations (RMPs) are surveyed by comparing AUG, DIII-D, EAST, and KSTAR stationary operating points. RMP-ELM suppression is achieved over a range of plasma currents, toroidal fields, and RMP toroidal mode numbers. Consistent operational windows in edge safety factor are found across devices, while windows in plasma shaping parameters are distinct. Accessed pedestal parameters reveal a quantitatively similar pedestal-top density limit for RMP-ELM suppression in all devices of just over $3\times 10^{19}$ m ^−3 . This is surprising given the wide variance of many engineering parameters and edge collisionalities, and poses a challenge to extrapolation of the regime. Wide ranges in input power, confinement time, and stored energy are observed, with the achieved triple product found to scale like the product of current, field, and radius. Observed energy confinement scaling with engineering parameters for RMP-ELM suppressed plasmas are presented and compared with expectations from established H and L-mode scalings, including treatment of uncertainty analysis. Different scaling exponents for individual engineering parameters are found as compared to the established scalings. However, extrapolation to next-step tokamaks ITER and SPARC find overall consistency within uncertainties with the established scalings, finding no obvious performance penalty when extrapolating from the assembled multi-device RMP-ELM suppressed database. Overall this work identifies common physics for RMP-ELM suppression and highlights the need to pursue this no-ELM regime at higher magnetic field and different plasma physical size.

The improved energy confinement mode (I-mode) is widely considered as an important operation regime for ITER. I-mode implementation depends on the specified basic plasma parameters and certain operation conditions, which are discovered by statistical plasma characteristics from a large number of I-mode discharges on a tokamak. The extraction process of I-mode plasma characteristics is complicated, time-consuming, and limited to the sampling rate of the measured signals. Experimental observation of the I-mode is accompanied by the appearance of a weakly coherent mode (WCM). However, it takes much time to accurately scan and quantify WCM characteristics when analyzing many I-mode discharges. Recently, a neural network identification method was developed as an I-mode detector to traverse a whole database as a replacement for manual identification. Two fully connected neural network models were trained with the spectrum of propagation velocity of density perturbation from Doppler backward scattering and the electron density measured by a polarimeter-interferometer system with the experimental advanced superconducting tokamak I-mode database. An accuracy of 98.30% in identifying WCMs in I-mode discharges is achieved with the WCM classification model. In addition, the regime classification model was also utilized to successfully distinguish between the low confinement mode (L-mode), I-mode, and high confinement mode (H-mode) with 96.03% accuracy. Finally, ablation experiments were performed on the regime classifiers, showing that there is potential for further performance improvement with future use of RNN model.

Improved confinement caused by Ne injection is investigated in EAST. The safety factor q profile and electron-scale turbulence are studied with a polarimeter-interferometer system and CO _2 laser collective scattering system in a series of Ne injection experiments in EAST. Ne with a suitable injection rate can lead to negative magnetic shear in the plasma core, which promotes electron-scale turbulence suppression. The continuous injection of Ne under the reversed shear condition further reduces the electron-scale turbulence intensity. Core electron temperature increase and confinement improvement are observed. QuaLiKiz is applied to model the influence of an impurity and reversed q on turbulence. The combination of stabilization effects of negative magnetic shear and Ne on turbulence is much stronger than the impurity’s own effect, which could explain the experimentally observed electron temperature ( T _e ) increase and turbulence suppression. Besides, the relationship between pedestal structure induced by Ne injection and lower hybrid wave (LHW)-driven current deposition is discussed. These results provide a new perspective to improve the confinement in the case of Ne injection, i.e. to adjust the Ne injection to modify the ideal plasma equilibrium.

The scrape-off layer (SOL) power width ( λ _q ) is an important parameter for predicting the heat load on divertor targets for future magnetically confined devices. Currently, the underlying physics for λ _q scaling is not fully understood. This paper extends the previous SOL particle flux width ( λ _js , λ _q ≈ λ _js is assumed, j _s is the measured ion current density used as a proxy for particle flux) scaling measured by the inboard divertor Langmuir probes (Div-LPs) in deuterium plasmas (Liu et al 2019 Plasma Phys. Control. Fusion 61 045001) to the λ _js scalings measured by the outboard Div-LPs in both deuterium and helium plasmas on EAST. A systematic method has been used to correct the upper-outer Div-LP measurements to reduce the measurement uncertainty of λ _js . About 520 discharges are selected for the construction of six databases (H-mode, L-mode, and Ohmic for deuterium and helium plasmas) to scale λ _js . Since the published λ _q scalings do not fit EAST λ _js well in the databases, four separate λ _js scalings are proposed using statistical method to minimize the scaling parameters. It is found that λ _js has a robust scaling dependence on the plasma line-averaged density ( ${\bar n_{\text{e}}}$ ) for both deuterium and helium plasmas and a strong positive scaling dependence on the divertor leg length ( L _div,leg ). The scaling exponent of the stored energy agrees well with the previous λ _q / λ _js scalings for deuterium plasmas. The near unity scaling exponents of edge safety factor and L _div,leg reveal the importance of parallel connection length in determining λ _js . The positive scaling dependence of λ _js on ${\bar n_{\text{e}}}$ and the 2 ∼ 4 times larger scaling amplitude compared with the published λ _q scalings probably indicate the turbulence broadening of λ _js on EAST. The comparison of λ _js for deuterium and helium plasmas using the constructed databases shows no significant difference. Statistical comparison of the normalized λ _js with different ELM types suggests that the grassy-ELM λ _js is slightly larger than the inter large-ELM (type-I ELM) λ _js for both deuterium and helium plasmas.

An innovative divertor concept, the fish tail divertor, is proposed in this paper, aimed at reducing the surface temperature on the tokamak divertor plate as well as that due to the edge localized modes. This new concept has been implemented in experiments to demonstrate its capability of strike point sweeping on the plate at a frequency range from 10 to 100 Hz by using an oscillating magnetic field. A strike point movement of 5–6 cm is achieved by applying a coil current of several percent of plasma current, leading to a significant reduction of divertor surface temperature. The result indicates a possible application in a fusion reactor.

The q _95 window for Type-I edge localized modes (ELMs) suppression using n = 4 even parity resonant magnetic perturbations (RMPs) has been significantly expanded to the ranges [3.9, 4.1] and [4.2, 4.8] in EAST while maintaining good confinement, which is demonstrated to be reliable and repeatable over the last two years. This window is significantly wider than the previous one achieved using n = 4 odd parity RMPs, which is around $q_{95} = 3.7\pm0.1$ . Here, n represents the toroidal mode number of the applied RMPs and q _95 is the safety factor at $95\%$ of the normalized poloidal magnetic flux. During ELM suppression, there is only a slight drop in the plasma stored energy and density ( ${\leqslant}10\%$ ). The comparison of changes in the pedestal profiles suggests that ELM suppression is achieved when the pedestal gradient is kept lower than a threshold. This wide q _95 window for ELM suppression is consistent with the prediction made by MARS-F modeling prior to the experiment, which located it at one of the resonant q _95 windows for plasma response. The Chirikov parameter taking into account plasma response near the pedestal top, which measures plasma edge stochasticity, significantly increases when q _95 exceeds 4, mainly due to the denser neighboring rational surfaces. The modeling of plasma response reveals a strong coupling between resonant and non-resonant components across the pedestal region, which is a characteristic of the kink-peeling like response observed during RMP-ELM suppression in previous studies on EAST. These promising results demonstrate the reliability of ELM suppression using the n = 4 RMPs in EAST and expand the physical understanding on ELM suppression mechanism.

An experiment was conducted to study the mode penetration of $n = 1$ resonant magnetic perturbation (RMP) in EAST under low neutral beam injection torque in the co-current direction (Co-NBI). The experimental results indicate that the threshold current ${I_{{\text{RMP,th}}}}$ for field penetration decreases with higher input torque ${T_{{\text{NBI}}}}$ . Furthermore, it is observed that the plasma mode frequency $| {{f_{{\text{MHD}}}}} |$ at counter-current direction is greatly reduced when the plasma toroidal rotation frequency ${f_\phi }$ increases. The theoretical scaling of mode frequency $( {{I_{{\text{RMP,th}}}} \propto {{| {{f_{{\text{MHD}}}}} |}^{0.70}}} )$ predicted by the field penetration theory is in good agreement with the experimental observation ( ${I_{{\text{RMP,th}}}} \propto {| {{f_{{\text{MHD}}}}} |^{0.53}}$ ). The role of $| {{f_{{\text{MHD}}}}} |$ and ${f_\phi }$ on the mode onset threshold was separately investigated using the full toroidal geometry initial value code MARS-Q (Liu et al 2013 Phys. Plasmas 20 042503). The numerical scaling based on the experimental mode frequency is consistent with the experimental and theoretical ones. Numerical results suggest that evaluating the total mode frequency $| {{f_{{\text{MHD}}}}} |$ is crucial in field penetration analysis, in contrast to toroidal rotation frequency ${f_\phi }$ . With the increase of ${T_{{\text{NBI}}}}$ , the decreasing $| {{f_{{\text{MHD}}}}} |$ leads to a reduction in the field penetration threshold. This suggests that more attention should be paid to error field tolerance in low Co-NBI torque scenarios, where the electron diamagnetic frequency may be canceled out by NBI-driven toroidal plasma rotation.

Alfvén eigenmodes (AE) activity is observed in the EAST high β _N and low B _T discharge 93910, operation scenario dedicated to explore the ITER baseline scenario. AEs are triggered after the plasma is contaminated by Tungsten that causes an abrupt variation of the thermal plasma and energetic particles (EPs) profiles. The aim of the present study is to analyze the AE stability in the 93910 discharge using the gyro-fluid code FAR3d, identifying the AE stability trends by comparing the plasma before and after the Tungsten contamination. Tungsten contamination causes the destabilization of Toroidal AEs (TAE) and Energetic particle modes (EPMs) in the same frequency range and radial location with respect to the experimental observation and M3D-K/GTAW code results. Next, a set of parametric studies are performed to analyze the effect of the thermal plasma and EP parameters on the AE stability. The analysis indicates a lower EP β threshold for the AEs destabilization if the EP energy increases, an improved AE stability of on-axis NBI configurations due to the stronger continuum damping in the inner plasma region as well as a large enhancement of the EP drive as the thermal ion density increases due to a higher ratio of the EP and Alfven velocities. Consequently, the simulations indicate the increment of the thermal ion density after the Tungsten contamination could be the main cause of the AE/EPM destabilization.

By combining the X-mode polarized lower and upper cut-off reflections obtained from the density profile reflectometer, we have successfully attained a comprehensive density profile spanning from the edge to the core region in pellet injection (PI) experiments on EAST. During the homogenization process after PI on EAST, an innovative method was introduced to quantify the local density increase. This approach employed the distinctive ‘dual-reflection’ phenomenon observed in the EAST microwave reflectometer, encompassing measurements of both the X-mode lower and upper cut-off frequencies. Furthermore, experimental investigations were carried out on EAST to comprehensively explore the parallel and poloidal expansion of the high-density pellet cloud. Notably, this study marks the first instance of measuring expansion velocities of pellet materials in both parallel and poloidal directions on EAST. A comparative analysis was performed initially between these experimental measurements and simulation results obtained from the HPI2 code, marking a pivotal stride towards enhancing its applicability in EAST.

Tungsten (W) is one of the most promising plasma-facing materials for future fusion devices. Although its melting point is the highest among all metals, it still has great risk of melting under extremely high plasma heat fluxes, which is a big concern for ITER and future reactors. Actively cooled W plasma-facing components (PFCs) with both monoblocks and flat-type structure have been successfully installed in the lower divertor of the EAST tokamak since 2021, and provide a good opportunity for direct comparison of the damage mechanism for the two types of PFCs. Various in situ melting phenomena on the lower divertor have been observed by CCD cameras, which have been further verified by post-mortem inspections. Severe melting and even exfoliation of the edge-beveled W plates were observed on some W/Cu flat-type components at horizontal outer targets. Many droplets were ejected during long-pulse operations, which induced a significant increase of W impurities and total irradiation in the core plasma, and thus greatly deteriorated the plasma performance and even caused disruptions. Two different shaping structures of flat-type PFCs show different positions of melting and the corresponding mechanisms. Slight melting was found on the sharp leading edges of W/Cu monoblocks between cassette modules (inter-CM) for horizontal targets with small droplet ejection, which was much improved compared to that observed on the upper W divertor, illustrating that the application of a large-sized bevel chamfer inter-CM was generally effective. In addition, an unexpected melting phenomenon on the dome plate was attributed to the extreme transient heat flux during disruption with runaway electrons. The application of both types of W/Cu PFCs for the divertor provides important experiences and lessons for the engineering design and optimization of divertor PFCs in future fusion devices.

The edge turbulence characteristics and the induced radial transport have been investigated in edge localized mode (ELM) suppression by using the $n = $ 4 resonant magnetic perturbation coils on EAST, with ${q_{95}} = $ 3.6 and the electron collisionality $\nu _{\text{e}}^* \approx $ 0.5. During ELM suppression, the edge turbulence is enhanced dramatically, as measured by the reciprocating probe and the poloidal correlation reflectometry. In the near SOL, the low frequency turbulence (250 kHz), which could be beneficial to form the observed turbulent transport. The estimated upstream cross-field particle flux is consistent with the total particle flux deposited on divertor targets, demonstrating that the enhanced radial turbulent particle transport is an important mechanism for particle exhaust in ELM suppression.

Ion Cyclotron Range of Frequencies (ICRF) heating and Neutral Beam Injection (NBI) can have synergy due to the acceleration of NBI beam ions by ICRF wave fields at their harmonics. To understand the influence of ICRF-NBI synergy on fast ion distribution and plasma performance, dedicated experiments and TRANSP simulations have been carried out on EAST. The simulation results are consistent with the experimental results. They show that the ICRF-NBI synergy not only accelerates the NBI beam ions with energy lower than 80 keV to energy larger than 300 keV, but also generates fusion neutrons with energy larger than 3 MeV. Moreover, ICRF-NBI synergy improves the plasma performance by increasing the poloidal beta, plasma stored energy, core ion temperature, total neutron yield and kinetic pressure. In a typical H-mode plasma with 1.0 MW NBI and 1.5 MW ICRF power, it was observed that ICRF-NBI synergy increases the poloidal beta, plasma stored energy, core ion temperature and neutron yield by ∼35%, 33%, 22% and 80%, respectively. Various parameter scans show that the ICRF-NBI synergetic effects can be enhanced by decreasing the minority ion concentration or the distance between the harmonic resonance and magnetic axis, or by increasing the ICRF heating power or NBI beam energy. Consequently, this leads to a generation of fast ions with higher energy. For instance, the maximum energy of the fast ion tail increases from 300 to 600 keV as n(H) decreases from 5% to 0.1%.

The plasma-beta effect on the n = 1 resonant magnetic perturbation (RMP) field penetration in purely radio-frequency (RF) wave heated discharges has been investigated in EAST. The experimental results show that the dependence of the threshold RMP coil current for field penetration, $I_{\textrm {RMP,th}}$ , on the total absorbed power P _tot scales as approximately $I_{\textrm {RMP,th}} \propto P_{\textrm {tot}}^{0.30}$ , indicating that the error-field tolerance is improved with increasing RF power. This is benefited by the increased electron perpendicular flow dominated by a counter-current electron diamagnetic flow with increasing RF power. However, theoretical scaling in cylindrical geometry overestimates the power index. Assuming an additional term $\beta_\textrm N^{\alpha_{\beta_\textrm N}}$ for the normalized beta in the scaling, it is shown that the fitted $\alpha_{\beta_\textrm N}$ from the experimental observation is around −1, indicating a degradation effect of plasma beta. To clarify the underlying physics of the plasma-beta effect that was not included in the theoretical scaling in cylindrical geometry, the MARS-Q code with full toroidal geometry is employed for simulation of nonlinear field penetration (Liu et al 2013 Phys. Plasmas 20 042503). The MARS-Q simulation results reproduce the $\beta_\textrm N$ dependence well, and hence the P _tot scaling of the threshold current in experimental observations. The main reason for this is that the net total torque, which is mainly contributed by the neoclassical toroidal viscosity (NTV), increases with increasing plasma $\beta_\textrm N$ . The results demonstrate that the nonlinear toroidal coupling effect via NTV torque plays an important role in determining field penetration, even in cases with relatively low $\beta_\textrm N \in [0.3,0.6]$ , which is far less than the no-wall beta limit.

The spectrum effects on toroidal rotation braking, induced by n = 1 resonant magnetic perturbations (RMPs) in the discharges with $q_{95} = 4.1$ and $q_{95} = 5.1$ , are studied in the EAST tokamak. Here n is the toroidal mode number, RMP spectrum is varied by scanning $\delta\phi_\mathrm{UL} $ , the phase difference between the upper and lower rows of RMP coils. The toroidal rotation changes periodically with the periodic $\delta\phi_\mathrm{UL} $ scanning and such an effect is stronger in the discharge with lower $q_{95} = 4.1$ . The spectrum dependence of the neoclassical toroidal viscosity (NTV) torque, modeled by NTVTOK based on the magnetic perturbation obtained from MARS-F calculation, agrees well with that of the experimentally observed braking torques in both discharges. The modeled NTV torque is stronger in the discharge with lower q _95 , which also agrees with the observations. The comparisons between the spectrum dependence of the NTV and magnetic perturbations show that the resonant mode of magnetic perturbations near the plasma edge mainly contribute the NTV torque. These agreements between modeling and experiments highlight the capability of NTV theory in explaining the experimental observation in the EAST tokamak.

A very high core electron temperature ( T _e0 ∼ 10 keV) plasma has been established and stably sustained by applying both lower hybrid wave (LHW) and on-axis electron cyclotron resonance heating (ECRH) in the Experimental Advanced Superconducting Tokamak (EAST). In this work, power balance analysis shows that the increase of ECRH power can increase the normalized T _e gradient significantly at the plasma core region ( ρ < 0.6), but does not change the T _e profile stiffness in the low-density L-mode plasmas. This has been considered to be due to a strong synergistic effect between ECRH and LHW. Furthermore, three distinguishable stages characterized by different T _e profile stiffnesses can be identified from the density ramp-up in the electron-heated plasma on EAST. A stronger T _e profile stiffness at ρ = 0.3 has been observed in the Stage-II, where the LHW power deposition gradually moves away from the plasma core region, following the electron density increases. Furthermore, the formation of an internal plasma density transport barrier inside ρ ∼ 0.6, accompanied by a sudden drop in core T _e and a rise in both core plasma density and ion temperature, has been observed for the first time during the transition from the Stage-II to the Stage-III when the central line-averaged plasma density reaches a threshold of 2.2 × 10 ^19 m ^−3 . This finding strongly affects further development of high-performance gas-fueled electron-heated plasma scenarios in EAST and suggests an advanced operational regime with a wide internal plasma density transport barrier.

In a reversed shear discharge in the Experimental Advanced Superconducting Tokamak (EAST), an internal transport barrier (ITB) was formed. For the first time, the turbulence in the plasma core before and after the formation of the ITB in EAST was investigated by microwave reflectometry. It was found that during the formation of the ITB, the turbulence inside the barrier was not completely suppressed. The density fluctuation level decreased in the early phase of ITB and significantly increased later. It was found that the density fluctuation level increased with the density gradient after the appearance of reversed shear Alfvén eigenmodes (RSAEs). The change in turbulence, measurable by the reflectometer, did not affect the rate of increase in core density. Through the measurement of RSAEs, it was found that the formation of this ITB may be related to the minimum q ( q _min ). q _min was close to 2.

In toroidal magnetic fusion devices, fast-ion D-alpha diagnostic (FIDA) is a powerful method to study the fast-ion feature. The fast-ion characteristics can be inferred from the Doppler shifted spectrum of Dα light according to charge exchange recombination process between fast ions and probe beam. Since conceptual design presented in the last HTPD conference, significant progress has been made to apply FIDA systems on the Experimental Advanced Superconducting Tokamak (EAST). Both co-current and counter-current neutral beam injectors are available, and each can deliver 2-4 MW beam power with 50-80 keV beam energy. Presently, two sets of high throughput spectrometer systems have been installed on EAST, allowing to capture passing and trapped fast-ion characteristics simultaneously, using Kaiser HoloSpec transmission grating spectrometer and Bunkoukeiki FLP-200 volume phase holographic spectrometer coupled with Princeton Instruments ProEM 1024B eXcelon and Andor DU-888 iXon3 1024 CCD camera, respectively. This paper will present the details of the hardware descriptions and experimental spectrum.

To investigate the fast ion behavior, a fast ion D-alpha (FIDA) diagnostic system has been installed on EAST. Fast ion features can be inferred from the Doppler shifted spectrum of Balmer-alpha light from energetic hydrogenic atoms. This paper will focus on the validation of FIDA measurements performed using MHD-quiescent discharges in 2015 campaign. Two codes have been applied to calculate the Dα spectrum: one is a Monte Carlo code, Fortran 90 version FIDASIM, and the other is an analytical code, Simulation of Spectra (SOS). The predicted SOS fast-ion spectrum agrees well with the measurement; however, the level of fast-ion part from FIDASIM is lower. The discrepancy is possibly due to the difference between FIDASIM and SOS velocity distribution function. The details will be presented in the paper to primarily address comparisons of predicted and observed spectrum shapes/amplitudes.

EAST was equipped with W-coated divertor plates and dome for the upper divertor, the remaining plasma facing components include the plasma facing components at lower divertor consisted of Carbon materials, at the baffles and main chamber walls consisted of Molybdenum materials. In order to study the control of W impurity on EAST with W divertor by an additional radiation losses, the radiative tungsten divertor with Ar impurity seeding for the control of tungsten impurity is studied by SOLPS-ITER simulation. The computational region includes the core, SOL and private flux region. The real geometry and material of plasma facing components (PFCs) are taken into account. The plasma species in the simulation include D0, D+1, e-, He0-He+2, C0-C+6, Mo0-Mo+42, W0-W+74 and Ar0-Ar+18. The simulation results show that with the increase of Ar seeding rate at the dome the plasma temperature at the target plates and the concentration of W impurity in the plasma decrease. The simulation plays an important role in the study of radiative tungsten divertor towards the higher heating power and steady state operation.

To study the dynamic characteristics of filamentary structure during the edge-localized mode (ELM) burst on the experimental advanced superconducting tokamak, the BOUT++ six-field two-fluid model is used to simulate the temporal and spatial evolutions of the ELM filaments. In the peak gradient region at the outer mid-plane, the radial speed of filaments is changed frequently and varied from 0.11 km/s to 0.79 km/s, and the poloidal speed oscillates in the range of 0.61–0.85 km/s. The width of the ELM filaments is increased from 7.76 mm to 26.7 mm and grows rapidly to 41.99 mm from t = 0.2717 ms to 0.2831 ms after the start of the burst. After that, the width decreases gradually and stabilizes at 25–30 mm when it comes to the saturation phase. Further research indicates that the filament propagation speed is increased during their outward movement. The width gradually decreases when the filaments move outward, and it is proportional to electron temperature (Te) perturbation. All the results are consistent with the experimental data presented by Chen et al., IEEE Trans. Plasma Sci. 47, 799 (2019).