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In this work, model-predictive control (MPC) was combined for the first time with singular perturbation theory, and an original plasma kinetic control method based on extremely simple data-driven models and a two-time-scale MPC algorithm has been developed. A comprehensive review is presented in this paper. Slow and fast semi-empirical models are identified from data, by considering the fast kinetic plasma dynamics as a singular perturbation of a quasi-static equilibrium, which itself is governed, on the slow time scale, by the flux diffusion equation. This control technique takes advantage of the large ratio between the time scales involved in magnetic and kinetic plasma transport. It is applied here to the simultaneous control of the safety factor profile, q (𝑥), and of several kinetic variables, such as the poloidal beta parameter, β _p , and the internal inductance parameter, l _i , on the EAST tokamak. In the experiments, the available control actuators were lower hybrid current drive (LHCD) and co-current neutral beam injection (NBI) from different sources. Ion cyclotron resonant heating (ICRH) and electron cyclotron resonant heating (ECRH) are used as additional actuators in control simulations. In the controller design, an observer provides, in real time, an estimate of the system states and of the mismatch between measured and predicted outputs, which ensures robustness to model errors and offset-free control. Based on the observer information, the controller predicts the behavior of the system over a given time horizon and computes the optimal actuation by solving a quadratic programming optimization problem that takes the actuator constraints into account. A number of control applications are described in the paper, either in nonlinear simulations with EAST-like parameters or in real experiments on EAST. The simulations were performed with a fast plasma simulator (METIS) using either two control actuators (LHCD and ICRH) in a low density scenario, or up to four actuators at higher density: LHCD, ECRH, and two NBI systems driven in a on/off pulse-width-modulation (PWM) mode, with different injection angles. The control models are identified with the prediction-error method, using datasets obtained from open loop simulations in which the actuators are modulated with pseudo-random binary sequences. The simulations with two actuators show that various q (𝑥) profiles and β _p waveforms can be tracked without offset, within times that are consistent with the resistive and thermal diffusion time scales, respectively. In simulations with four actuators, simultaneous tracking of time-dependent targets is shown for q (𝑥) at two normalized radii, 𝑥 = 0 and 𝑥 = 0.4, and for β _p . Due to the inherent mismatch between the optimal NBI power request and the delivered PWM power, the kinetic controller performs with reduced accuracy compared with simulations that do not use the NBI/PWM actuators. The first experimental tests using this new control algorithm were performed on EAST when the only available actuator was the LHCD system at 4.6 GHz. The algorithm was thus used in its simplest single-input-single-output version to track time-dependent targets for the central safety factor, q _0 , or for β _p . In the closed loop control experiments, the q _0 targets were tracked in about one second, consistently with the plasma resistive time constant. Excellent tracking of a piecewise linear β _p target waveform was also achieved. When the NBI system became controllable in real time by the EAST plasma control system, new experiments were dedicated to multiple-input-multiple-output MPC control with three actuators: LHCD and two NBI actuators using the PWM algorithm. Given that the minimum time allowed between NBI on/off switching was 0.1 s, i.e. larger than the characteristic time of the fast plasma dynamics, a reduced version of the MPC controller based only on the slow model was used. Various controller configurations were tested during a single experimental session, with up to three controlled variables chosen among q _0 = q ( 𝑥 = 0), q _1 = q ( 𝑥 = 0.5), β _p and l _i . The main difficulty encountered during this session was the unavailability of the full baseline ICRH and ECRH powers that were used in the reference scenario, and from which the plasma model was identified. This often led to the saturation of one or several actuators, which prevented some targets selected in advance from being accessible. Nevertheless, in cases that were free from actuator saturation, q _0 and q _1 targets were successfully reached, in a time that is consistent with the resistive diffusion time of the model and with small oscillations that are characteristic of the PWM operation of the neutral beams. During the simultaneous control of q _0 and β _p , the ICRH power was too low and, in addition, the plasma density was much larger than the reference one. The q _0 targets were not accessible in this high-density/low-power case, but β _p control was successful. Finally, the simultaneous control of q _0 and l _i was satisfactory and, during the simultaneous control of, q _0 , β _p and l _i , the tracking of β _p and l _i was satisfactory but q _0 was too large due to the lack of ICRH power and to NBI saturation. In conclusion, the extensive nonlinear simulations described in this paper have demonstrated the relevance of combining MPC, data-driven models and singular perturbation methods for plasma kinetic control. This technique was also assessed experimentally on EAST, although some tests were perturbed by undesired parameter changes with respect to the reference scenario.

This work presents a robust methodology for effectively distinguishing between the seed island and the onset of the neoclassical tearing mode (NTM) in the EAST tokamak. Unlike previously employed methods, the width of the seed island is carefully regulated by gradually ramping up the currents in the resonant magnetic perturbation (RMP) coils. Remarkably, the seed island phase can be sustained for several hundred milliseconds, providing ample time to examine the nonlinear dynamics of NTM threshold physics. This study investigates the plasma response in terms of various factors, including the plasma rotation, the electron density, and the electron temperature, from the initial formation of the seed island to the saturation of the NTM island. Through extensive statistical analysis of NTM triggering events, it is revealed that both the critical width of the seed island and the critical RMP currents are positively correlated with β _p . Moreover, the relationship of the transition time associated with mode penetration to β _p differs from the dependence of the transition time for NTM triggering on β _p . The growth rates associated with these two nonlinear phenomena show similar characteristics. Finally, reduced magnetohydrodynamic (MHD) modeling replicates the two nonlinear bifurcation states observed in the experiment. It is noteworthy that the RMP-induced NTM islands are locked to the static magnetic perturbation. This scenario differs from the natural excitation of NTMs, which occurs due to transient MHD phenomena, such as sawtooth crashes, where the triggering source is a rotating magnetic perturbation. This leads to the formation of rapidly rotating islands at a natural frequency. Furthermore, the stability of these islands is expected to be related to the polarization current effect. Nevertheless, this experiment sets a new course for the comprehensive investigation and understanding of NTM threshold physics, providing valuable insights for future tokamak design and operational strategies.

The attainment of a reliable equilibrium is a critical aspect of tokamak experiments and physics analysis. A common method for establishing a reliable equilibrium current involves reconstructing it from indirect measurements, such as those obtained from polarimeter-interferometers (POINT) and motional Stark effect (MSE) systems. However, uncertainties still exist in the reconstruction results. For the equilibrium reconstruction on the EAST tokamak, which is based on the POINT system, the primary sources of uncertainty are the limited scope of measurements and the sensitivity of the reconstruction process. This paper proposes an enhanced approach that utilizes current simulation as a constraint to maintain consistency between the initial equilibrium and the simulated results. The radio frequency waves driven current is identified as a particularly influential component due to its interaction with the q profiles of the equilibrium and the deposition region of the waves. Two specific discharges are presented to illustrate how a new equilibrium can be achieved, which enhances consistency between the equilibrium and the simulated current, taking into account the dependencies of various components.

To address the needs for a fusion pilot plan design, DIII-D/EAST joint experiments on DIII-D have demonstrated high normalized beta β _N ∼ 4.2, toroidal beta β _T ∼ 3.3% with q _min > 2, q _95 ⩽ 8 sustained for more than six energy confinement times in high poloidal beta regime. The excellent energy confinement quality ( H _98y2 ∼ 1.8) is achieved with an internal transport barrier at high line-averaged Greenwald density fraction f _Gr > 0.9. The trapped gyro-Landau fluid (TGLF) modeling of the transport characteristics shows that the beam-driven rotation does not play an important role in the high confinement quality. The modeling also captures very well several transport features, giving us confidence in using integrated modeling to project these experimental results to future machines. The high-performance phase is terminated by fast-growing modes triggered near the n = 1 ideal-wall kink stability limit. New radio frequency (RF) capabilities for off-axis current drive could remove the residual ohmic current to achieve a fully non-inductive state, and improve the mode–wall coupling to increase the ideal-wall β _N limit, enabling sustainment of the fully non-inductive high performance plasma in stationary conditions.

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.

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.

Integration of transient and steady-state divertor heat fluxes control with a high-performance core is necessary for future fusion reactors. In recent DIII-D high- ${\beta _{\text{p}}}$ experiments, divertor detachment and simultaneous edge localized mode (ELM) suppression are demonstrated while the plasma confinement quality is maintained high in ITER-similar shape. By optimizing the neon injection in high- ${\beta _{\text{p}}}$ scenario with ITER-similar shape, deep detachment and ELM suppression are achieved with a high-performance core ( ${\beta _{\text{N}}}$ ∼ 2.8, ${\beta _{\text{p}}}$ ∼ 2.3) at ${q_{95}}$ ∼ 7.5. Partial divertor detachment and suppression of large ELMs are achieved at ${q_{95}}$ ∼ 6. The stability analyses suggest that with low neon injection, the density pedestal becomes higher and steeper and the ${T_{\text{i}}}$ profile also increases, therefore the increased edge pressure and higher current density destabilize the Peeling-Ballooning mode (PBM), which would lead to a large ELM collapse. With strong neon gas puffing, the significantly reduced pedestal pressure and current density, due to the degraded ${T_{\text{e}}}$ pedestal, lead to the stabilization of PBM and ELMs are suppressed. For both cases, the coupling between the large radius internal transport barrier (ITB) and edge pedestal is the key reason for maintaining high global performance. The formation of large radius ITB compensates for pedestal degradation. Such results could provide an attractive scenario to well control the transient and steady-state heat flux onto the divertor plates while maintaining good plasma performance, which is an important step toward the steady-state operation of future fusion reactors.

Enhanced confinement is observed in EAST low ${q_{95}}$ ( ${q_{95}} < 3.5$ ) plasmas with the temperature ratio ( ${T_i}/{T_e}$ ) increasing. A statistically significant positive correlation between ${H_{98y2}}$ and ${T_i}/{T_e}$ has been established. Increased ${T_i}\,$ triggerd by Neutral Beam Injection in experiments maintaining constant ${T_e}$ leads initially to an improved ${T_i}/{T_e}$ ratio, which subsequently mitigates heat transport in the ion channel. The Trapped Gyro–Landau Fluid transport model was applied to analyse the dependence of the turbulent transport on ${T_i}/{T_e}$ . Simulation results demonstrate that the Ion Temperature Gradient mode is stabilized in high ${T_i}/{T_e}$ plasma, with ion flux results from predictive modelling further validating this effect. These findings corroborate the Weiland model and complement one another. This study contributes to a deeper understanding of how the dimensionless parameter ${T_i}/{T_e}$ influences plasma transport, and helps demonstrate the feasibility of operations under the low ${q_{95}}$ condition required for the ITER-inductive scenario in the new ${T_i}/{T_e}$ regime.

This paper presents the first results of a passive active multijunction (PAM) launcher at 2.45 GHz during the lower hybrid current drive (LHCD) experiments on EAST. Good coupling performance with a power reflection coefficient ( RC ) ∼3% has been achieved at the plasma–antenna distance up to ∼11 cm in L-mode edge plasmas without local gas puffing near the PAM launcher. Reliable power coupling of this PAM during the edge perturbations induced by type I edge localized modes (ELMs) has been successfully demonstrated. Compared with the old full active multijunction (FAM) launcher, the new PAM can be placed ∼2 cm further away from the plasma in normal operations, which is in good agreement with the previous prediction (Li et al 2019 Fusion Eng. Des. 147 111 250), by the linear wave–plasma coupling code ALOHA (Hillairet et al 2010 Nucl. Fusion 50 125 010). The flexibility of the power spectrum by changing the phase difference between adjacent modules was validated and ray-tracing/Fokker–Planck simulations can reproduce the experimental features. The achievable power handling is as high as 25 MW m ^−2 , although with a shot pulse length of ∼10 s. The first experiment successfully demonstrated the coupling performance of a PAM launcher at low density and this launcher construction provides helpful engineering experience for the 4.6 GHz PAM development in the near future on EAST.

This paper describes scalings of the H-mode confinement database of neutral beam injection (NBI)-lower hybrid wave (LHW) and electron cyclotron resonant heating-LHW plasmas in the EAST tokamak. Fundamental information of the EAST H-mode database and the details of calculation of the thermal energy confinement time in EAST are presented. The result of the scaling model ${\tau _{{\text{E,EASTS,H,NBI - LHW}}}}$ of the thermal energy confinement time in NBI-LHW plasmas with root-mean-square-error ${\text{RMSE}} = 5.5\% $ , compared with the multi-machine scaling model ${\tau _{{\text{th,98}}y{\text{2}}}}$ derived from the multi-machine database (mainly NBI heating) with ${\text{RMSE}} = 12.9\% $ , is obtained. The scaling ${\tau _{{\text{E,EAST,H,NBI - LHW}}}}$ presents a similar dependency on plasma current and power loss and a lower dependency on plasma density and elongation, but stronger dependency on toroidal field than in ${\tau _{{\text{th,98}}y{\text{2}}}}$ . The scaling ${\tau _{{\text{E,EAST,H,ECRH - LHW}}}}$ demonstrates a similar dependency on plasma current and power loss and lower dependency on plasma density, but a negative dependency with large error bars on elongation compared with ${\tau _{{\text{E,EAST,H,NBI - LHW}}}}$ in NBI-LHW plasmas. The comparison of EAST models with the more recent multi-tokamak scaling ITPA20-IL shows a similar engineering dependency on plasma current, power loss and plasma density, but opposite dependency on toroidal field.

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.

This paper presents the progress in the long—pulse operation of the electron cyclotron (EC) system and the achievements in high—electron temperature plasmas by the combined EC and lower hybrid (LH) waves heating since the EC system was built in 2015. An electron temperature of up to 12 keV with a duration over 100 s was realized by the simultaneous heating of EC and LH waves at the line-averaged density ${\bar n_{\text{e}}}$ ∼ 1.8 × 10 ^19 m ^−3 . The plasma heating effect strongly depends on the location of EC power deposition. H-mode plasmas with solely EC wave auxiliary heating have been obtained on EAST for the first time. These H-mode discharges show an enhanced confinement factor H _98(y,2) around 1.0, which is higher than the previous H-mode using LH power alone (Xu et al 2011 Nucl. Fusion 51 072001). The total heating power is very close to the threshold value for L–H transition according to the international tokamak scaling. In addition to the temperature effects inside the separatrix, higher electron temperature produced by EC wave is found to reduce the LH power loss in the scrape-off layer due to collisional absorption, which is beneficial to further increase the LH current drive efficiency. Ray-tracing/Fokker–Planck modeling results indicate that higher electron temperature can shorten the LH wave propagation on EAST in a multiple-pass regime, thus decreasing the collisional dissipation.

A newly developed radial directional electron probe (DEP) has been applied to the unidirectional electron current measurement on EAST tokamak. The DEP consists of two radial arrays of channels which have opposite directions and align along the local magnetic field line. Each radial array has 6 holes with a radial interval of 5 mm. Every channel has a hole with 0.5 mm radial width, 3 mm depth and 15° poloidal opening angle. The graphite collector embedded inside the hole is biased to positive potential to repel low energy ions, and high energy ions are blocked by the hole surface because their Larmor radii are larger than the radial width of hole. In consequence, the ion current collected by the DEP collector can be ignored in contrast with the electron current, as demonstrated by the I-V characteristics in a DEP commissioning experiment. The difference of collected current between two opposite channels signifies the unidirectional electron current in the flux tube. In a lower hybrid wave (LHW) modulation experiment, the amplitude and radial structure of unidirectional electron current induced by LHW is measured directly by this radial DEP array, and the LHW filament current covers over 20 mm radial region with a maximum of 20 A/cm2.

Hot spot is a serious challenge limiting long pulse operation with LHCD (lower hybrid current drive) in tokamak. In order to mitigate the hot spot in guard limiter and improve LHCD capability, effect of edge plasma density on inducing hot spot and current drive has been studied in EAST. The temperature in the guard limiter of the LH antenna, inducing hot spot directly, increases with edge density and LH power. Studies show that the hot spot is mainly ascribed to the heat flux in front of LH antenna. Further simulation indicates that such spot corresponds to the peak position of edge density due to local LH electric field. In addition, due to the stronger parametric instability (PI) behavior in the case of higher edge density, the current drive capability decreases with edge density. Strike-point splitting behaviors appear with density increase, in agreement with current profiles in the edge region and the reduction of total driven current, suggesting that more power is deposited in edge region, which does more contribution to hot spot. Above studies offer one possible idea to optimize the edge density so as to satisfy the coupling, mitigate the heat flux in the guard limiter, and improve current drive capability for fusion device.

Joint DIII-D/EAST experiments in the radio-frequency (RF) heated H-mode scheme with comparison to that of neutral beam (NB) heated H-mode scheme were carried out on DIII-D and EAST under similar conditions to examine the effect of heating scheme on scrape-off layer (SOL) width in H-mode plasmas for application to ITER. A dimensionally similar plasma equilibrium was used to match the EAST shape parameters. The divertor heat flux and SOL widths were measured with infra-red camera in DIII-D, while with divertor Langmuir probe array in EAST. It has been demonstrated on both DIII-D and EAST that RF-heated plasma has a broader SOL than NB-heated plasma when the edge electrons are effectively heated in low plasma current and low density regime with low edge collisionality. Detailed edge and pedestal profile analysis on DIII-D suggests that the low edge collisionality and ion orbit loss effect may account for the observed broadening. The joint experiment in DIII-D has also demonstrated the strong inverse dependence of SOL width on the plasma current in electron cyclotron heated (ECH) H-mode plasmas. Keywords: SOL width, Radio-frequency heating, Neutral beam heating, Low collisionality

The Experimental Advanced Superconducting Tokamak (EAST) has recently achieved a variety of H-mode regimes with different edge-localized mode (ELM) dynamics, including type-I ELMs, compound ELMs, which are manifested by the onset of a large spike followed by a sequence of small spikes on Da emissions, usual type- III ELMs, and very small ELMs. This newly observed very small ELMy H-mode appears to be similar to the type-II ELMy H-mode, with higher repetition frequency (~1 kHz) and lower amplitude than the type-III ELMy H-mode, exhibiting an intermediate confinement level between type-I and type-III ELMy H-modes. The energy loss and divertor power load are systematically characterized for these different ELMy H-modes to provide a physics basis for the next-step high-power long-pulse operations in EAST. Both type-I and compound ELMs exhibit good confinement (H98y,2) ~ 1). A significant loss of the plasma stored energy occurs at the onset of type-I ELMs (~8%) and compound ELMs (~5%), while no noticeable change in the plasma stored energy is observed for the small ELMs, including both type-III ELMs and very small ELMs. The peak heat flux on divertor targets for type-I ELMs currently achieved in EAST is about 10 MW m-2, as determined from the divertor-embedded triple Langmuir probe system with high time resolution. As expected, type-III ELMs lead to much smaller divertor power loads with a peak heat flux of about 2MWm-2. Peak power loads for compound ELMs are between those for type-I and type-III ELMs. It is remarkable that the new very small ELMy H-modes exhibit even lower target power deposition than type-III ELMs, with the peak heat flux generally below 1 MWm-2. These very small ELMs are usually accompanied by broadband fluctuations with frequencies ranging from 20 to 50 kHz, which may promote particle and power exhaust throughout the very small ELMy H-mode regime. [ABSTRACT FROM AUTHOR]

The attainment of long-pulse, high-performance, fully non-inductive plasma is one of the major scientific objectives of EAST, using the ITER-like tungsten upper divertor. Understanding and optimizing the fast-ion behaviors is the critical issue to extending the performance of EAST. Recently, using both neutral beam injection (NBI) and radio frequency (RF; low hybrid, electron cyclotron, and ion cyclotron) heating, fully non-inductive high-βP scenarios with extension of fusion performance at high density and low rotation have been achieved, with βP up to 2.5, βN up to 2.0, H98y 2  >  1.1, and bootstrap current fraction (f BS) up to 50%. For previous long-pulse H-mode plasma at medium density, when NBI is added into RF plasma, βp is increased from 1.2 to 2.0 compared with RF-only discharges. In fact, f BS for both discharges is nearly the same, at ~22%. Analysis shows that the increase in βp is mostly due to fast ions which do not contribute significantly to the neoclassical bootstrap current. Thus, to obtain high-performance plasmas with improved bootstrap current fraction, key parameters (e.g. density, beam energy, etc.) must be further optimized. Experimental results show that high density improves bootstrap fraction also by reducing fast-ion slowing-down time and loss. The lower beam energy also mitigates fast-ion loss, which is better for heating and CD performance. The extension of high-performance, fully non-inductive experiments on EAST at high density and zero/low NBI torque can potentially offer unique contributions towards ITER and CFETR. [ABSTRACT FROM AUTHOR]

Since the last IAEA Fusion Energy Conference in 2016, the EAST physics experiments have been developed further in support of high-performance steady-state operation for ITER and CFETR. First demonstration of a  >100 s 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 upper tungsten divertor has been achieved on EAST using the pure radio frequency (RF) power heating and current drive. The EAST operational domain has been significantly extended towards a more ITER and CFETR related high beta steady-state regime (βP ~ 2.5 and βN ~ 1.9 of using RF and NB and βP ~ 1.9 and βN ~ 1.5 of using pure RF). A large bootstrap current fraction up to 47% has been achieved with with q95 ~ 6.0–7.0. The interaction effect between the electron cyclotron resonant heating and two lower hybrid wave systems has been investigated systematically, and applied for the improvement of current drive efficiency and plasma confinement quality in the steady-state scenario development on EAST. Full edge-localized mode (ELM) suppression using the n  =  2 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with a range of the edge safety factor of q95  ≈  3.2–3.7 on 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 0D model prediction for a higher bootstrap fraction. Towards a long-pulse, high bootstrap current fraction operation, a new lower ITER-like tungsten divertor with active water-cooling will be installed, together with further increase and improvement of heating and current drive capability. [ABSTRACT FROM AUTHOR]

Experimental and modeling investigations on the Experimental Advanced Superconducting Tokamak (EAST) show attractive confinement and stability properties in fully non-inductive, high poloidal beta plasmas. In the 2018 EAST experimental campaign, extended operation regimes of steady-state scenario were obtained (βP ~ 1.9 & βN ~ 1.5 & H98y 2 ~ 1.3 of using only RF heating) with a high bootstrap current fraction (f BS ~ 47%) and ne/nGW ~ 70%. The confinement quality, H98y 2 ~ 1.3, is much better than standard H-mode, and stationary peaked electron temperature profiles and peaked current density profile when ~1 MW of ECH and ~2.6 MW of LHW are both deposited in the core region. The observed improvement in plasma confinement is much better (H98y 2 ~ 1.3) when compared with the RF-dominant heating experiments in the EAST 2016–2017 experimental campaign (H98y 2 ~ 1.1). Integrated modeling prediction suggests that high electron density would increase the plasma performance and bootstrap current fraction, which is consistent with the general experimental trend. Linear analysis shows that the high-k (ky   >  1) modes instability (ETG) is suppressed in the core region. Also, the Shafranov shift is shown to play a role in the suppression of the electron turbulent energy transport. Besides the modeling predictions, the validation of the predicted of the effect of ECH on the plasma confinement in recent experiments was done and the experimental results were consistent with the modeling results. The validation results also suggest that when ECH is deposited in the core region in the RF heating experiments, increasing the ECH heating power from 0.5 MW to 1.0 MW does make a small improvement in the bootstrap current fraction. The high bootstrap fraction scenario realized on EAST and the investigation to achieve higher-performance plasma would help expanding the operation regime on EAST. [ABSTRACT FROM AUTHOR]

An internal transport barrier (ITB) and double tearing modes (DTM) have been observed during the off-axis sawteeth in EAST. The ITB of electron temperature Te is modulated by the sawteeth oscillation, and the formation of an ITB can be divided into three stages: (i) the transport produced by a sawteeth final crash is suppressed at the first stage with a steep gradient of Te; (ii) the micro-instability is developed at the second stage for further increasing the gradient of Te; (iii) the ITB is formed eventually after the transition from beta-induced Alfvén eigenmodes (BAEs) to reversed shear Alfvén eigenmodes (RSAEs), where the BAEs-RSAEs pair enables the tracking of directly from the experiment. The off-axis sawteeth final crash is triggered by DTM: (i) the DTM can be excited by the redistribution of the profile of thermal particles, where the downward transport of energetic ions is detected indirectly by the soft x-ray arrays for the first time; (ii) the DTM can be excited by the transformation from the kink instability. [ABSTRACT FROM AUTHOR]

The plasma current profile modification in the edge region is observed in LHCD (lower hybrid current drive) plasmas for the first time in EAST, correlated with an increased level of PI (parametric instability) activities with the decrease in LH (lower hybrid) source frequency or the increase in plasma density. With the increase in the PI strength, an increase in the edge current is observed despite of a reduction in current drive capability. In these plasmas, strike-point splitting behaviors are also observed, implying the modification of the magnetic geometry in the scrape-off-layer. Results show the effects of LH frequency and plasma density on power dissipation in the edge region, suggesting a role of anomalous wave power damping on the modification of current profile at the plasma edge. These novel results are significant in that they give insight for the first time into how nonlinear wave-plasma interactions such as PI may directly impact the edge current profile, the control of which is critical in order to achieve optimized modes of operation in a steady state fusion reactor. [ABSTRACT FROM AUTHOR]

A toroidal Alfvén eigenmode (TAE) excited by barely trapped energetic electrons during the application of a static magnetic perturbations (MPs) is observed for the first time in tokamak ohmic heating plasmas. This TAE appears when the current of a static n  =  2 MPs exceeds a threshold value, at which a forced n  =  2 magnetic reconnection happens. Here, n is the toroidal mode number. This TAE is located near the plasma edge, which agrees with the calculation of the TAE gap. It propagates in the ion diamagnetic direction and has a dominant toroidal mode number n  =  2. The frequency of the TAE is consistent with the precessional frequency of the barely trapped energetic electrons with energy around 150 keV. An increase of the hard x-ray emission intensity in this energy range is observed just after the n  =  2 magnetic reconnection. The energetic electrons may be accelerated by the magnetic islands formed after reconnection. The TAE is stabilized after a reduction in the contents of energetic electrons. This observation suggests that the application of static MPs may provide us with a novel method to actively control Alfvén eigenmodes in fusion plasmas. [ABSTRACT FROM AUTHOR]

Alfvén eigenmodes are destabilized at the DIII-D pedestal during transient beta drops in high poloidal β discharges with internal transport barriers (ITBs), driven by n  =  1 external kink modes, leading to energetic particle losses. There are two different scenarios in the thermal β recovery phase: with bifurcation (two instability branches with different frequencies) or without bifurcation (single instability branch). We use the reduced MHD equations in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles as well as the geodesic acoustic wave dynamics, to study the properties of the instabilities observed in the DIII-D high poloidal β discharges and identify the conditions to trigger the bifurcation. The simulations suggest that instabilities with lower frequency in the bifurcation case are ballooning modes driven at the plasma pedestal, while the instability branch with higher frequencies are low n (n  <  4) toroidal Alfvén eigenmodes nearby the pedestal. The reverse shear region between the middle and plasma periphery in the non-bifurcated case avoids the excitation of ballooning modes at the pedestal, although toroidal Alfvén eigenmodes and reverse shear Alfvén eigenmodes are unstable in the reverse shear region. The n  =  1 and n  =  2 Alfvén eigenmode activity can be suppressed or minimized if the neutral beam injector (NBI) intensity is lower than the experimental value (). In addition, if the beam energy or neutral beam injector voltage is lower than in the experiment (), the resonance between beam and thermal plasma is weaker. The and 6 AE activity can not be fully suppressed, although the growth rate and frequency is smaller for an optimized neutral beam injector operation regime. In conclusion, AE activity in high poloidal β discharges can be minimized for optimized NBI operation regimes. [ABSTRACT FROM AUTHOR]

In the 2017 EAST experimental campaign, a steady-state long-pulse H-mode discharge lasting longer than 100 s has been obtained using only radio frequency heating and current drive, and the confinement quality is slightly better than standard H-mode, H98y2 ~ 1.1, with stationary peaked electron temperature profiles. Integrated modeling of one long-pulse H-mode discharge in the 2016 EAST experimental campaign has been performed with equilibrium code EFIT, and transport codes TGYRO and ONETWO under integrated modeling framework OMFIT. The plasma current is fully-noninductively driven with a combination of ~2.2 MW LHW, ~0.3 MW ECH and ~1.1 MW ICRF. Time evolution of the predicted electron and ion temperature profiles through integrated modeling agree closely with that from measurements. The plasma current (Ip ~ 0.45 MA) and electron density are kept constantly. A steady-state is achieved using integrated modeling, and the bootstrap current fraction is ~28%, the RF drive current fraction is ~72%. The predicted current density profile matches the experimental one well. Analysis shows that electron cyclotron heating (ECH) makes large contribution to the plasma confinement when heating in the core region while heating in large radius does smaller improvement, also a more peaked LHW driven current profile is got when heating in the core. Linear analysis shows that the high-k modes instability (electron temperature gradient driven modes) is suppressed in the core region where exists weak electron internal transport barriers. The trapped electron modes dominates in the low-k region, which is mainly responsible for driving the electron energy flux. It is found that the ECH heating effect is very local and not the main cause to sustained the good confinement, the peaked current density profile has the most important effect on plasma confinement improvement. Transport analysis of the long-pulse H-mode experiments on EAST will be helpful to build future experiments. [ABSTRACT FROM AUTHOR]

High pressure noble gas injection is a promising technique to mitigate the effect of disruptions in tokamaks. In this paper, results of mitigation experiments with low-Z massive gas injection (helium) on the EAST tokamak are reported. A fast valve has been developed and successfully implemented on EAST, with valve response time  ⩽150 μs, capable of injecting up to particles, corresponding to 300 times the plasma inventory. Different amounts of helium gas were injected into stable plasmas in the preliminary experiments. It is seen that a small amount of helium gas () can not terminate a discharge, but can trigger MHD activity. Injection of 40 times the plasma inventory impurity () can effectively radiate away part of the thermal energy and make the electron density increase rapidly. The mitigation result is that the current quench time and vertical displacement can both be reduced significantly, without resulting in significantly higher loop voltage. This also reduces the risk of runaway electron generation. As the amount of injected impurity gas increases, the gas penetration time decreases slowly and asymptotes to (∼7 ms). In addition, the impurity gas jet has also been injected into VDEs, which are more challenging to mitigate that stable plasmas. [ABSTRACT FROM AUTHOR]

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 the 2014 IAEA FEC, EAST has been upgraded with all ITER-relevant auxiliary heating and current drive systems, enabling the investigation of plasma profile control by the coupling/integration of various auxiliary heating combinations. Fully non-inductive steady-state H-mode plasma (H98,y2  >  1.1) was extended over 60 s for the first time with sole RF heating plus good power coupling and impurity and particle control. 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 ne ~ 4.5  ×  1019 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 Vloop ~ 0.0 V at high βP ~ 1.8 and high-performance in upper single-null (ε ~ 1.6) configuration with the tungsten divertor; (ii) discovery of a stationary H-mode regime with no/small ELM using 4.6 GHz LHCD, and; (iii) achievement of ELM suppression in slowly rotating H-mode plasma with n  =  1 and 2 RMP compatible with long-pulse operations. The new advances in scenario development provide an integrated solution in achieving long-pulse steady-state operations on EAST. [ABSTRACT FROM AUTHOR]

An efficient technique for building plasma equilibria with preset or constrained q profiles by solving the nonlinear Grad–Shafranov equation iteratively is proposed. The technique uses a set of dynamic virtual poloidal field probes in the plasma core region. Numerical examples are shown for monotonic q profiles and negative central magnetic shear equilibria which are obtained with this technique. As a proof of principle, the technique is applied on the EAST tokamak, to constrain the plasma equilibrium reconstruction with the experimental q  =  1 surface information from the ECE diagnostic in addition to the external magnetic measurements. [ABSTRACT FROM AUTHOR]