Fast and reliable discharge modeling is an essential tool for tokamak scenario development, as extensive simulations are required to explore the feasible operational space and translate physical targets into implementable actuator waveforms. In this work, a fast discharge simulator for Experimental Advanced Superconducting Tokamak (EAST) was developed by coupling the fast integrated tokamak modeling tool METIS with the free-boundary equilibrium code FBT from the MEQ (Matlab EQuilibrium) suite. METIS provides the evolution of the plasma profiles by self-consistently solving heat and particle transport with source profiles. The generated outputs are then used directly in FBT to computes the poloidal field (PF) coils currents needed in order to obtain a given plasma shape. The coupling strategy is based either on matching the free functions ( p '(psi) and FF '(psi)) on the right-hand side of the Grad-Shafranov equation, or on matching global quantities such as the plasma current Ip and stored energy WMHD. Both approaches are effective in ensuring consistency of the pressure and current-density profiles between the two codes. The last closed flux surface (LCFS) curve computed by FBT is fed back into METIS to update the LCFS, and METIS is then run again. The METIS-FBT iteration is repeated until convergence is reached. The results obtained with the equilibrium code FBT are first benchmarked against experimental data and reconstruction results available on EAST. The coupled METIS-FBT workflow is then assessed through a post-shot simulation of an radio-frequency heated EAST discharge, showing its ability to reproduce the main plasma evolution and generate inverse free-boundary PF-coil current trajectories consistent with experimental references. Its predictive scenario-design capability is further demonstrated through the design of a synthetic discharge scenario.

Recent achievements of lower hybrid current drive (LHCD) experiment at 4.6 GHz towards long-pulse and high-power operation on EAST tokamak are reported. The duration of long-pulse plasmas has been extended to 1056 s with 1.1 MW LH power in I-mode and 1066 s with 0.92 MW in H-mode. The operational domains in plasma current and line-averaged density for fully non-inductive discharges with LH alone and with the combined LH and electron cyclotron (EC) waves are presented. The dependence of LHCD efficiency on plasma density in both L- and H-mode discharges characterized by residual loop voltage V _loop = 0 is quantified. It is found that the LHCD efficiency is improved significantly by EC heating due to the increase of electron temperature. The dominant issues in long-pulse and high-power operation are summarized and discussed, including the power coupling imbalance, the hot spot and arc events in front of the antenna, and the deteriorated plasma heating effect with high LH power. Finally, prospects with a new 4 MW LHCD system at 4.6 GHz which is under development are given.

We report the first in-depth comparison of the impact of toroidal magnetic field direction on solid boron injection used for Edge-Localized Mode (ELM) control, power exhaust, and core high- Z impurity control in the Experimental Advanced Superconducting Tokamak. With favorable ion ∇ B drift towards the upper X-point in an upper-single-null configuration, boron injection effectively suppresses ELMs, produces a detachment of the inner divertor target, and leads to improved energy confinement. ELM suppression in this configuration is accompanied by the excitation of an Edge Harmonic Mode. In contrast, with unfavorable ion ∇ B drift away from the upper X-point, boron injection also suppresses ELMs but leads to a more symmetric detachment state of both the inner and outer divertor targets, while plasma energy confinement is slightly degraded despite similar boron injection levels; a different low-frequency coherent mode without multiple harmonics is observed. Measurements from toroidally separated views show that the divertor response to boron injection is essentially toroidally symmetric, supporting the use of two-dimensional SOLPS-ITER modeling with a toroidally uniform impurity source. These experimental observations are qualitatively consistent with SOLPS-ITER simulations, which highlight the critical role of E × B drift effects in setting the Bt -dependent in–out asymmetry of detachment and in asymmetrically transporting particles and injected impurities within the scrape-off layer and private-flux region. These findings underscore the importance of drift physics and real-time wall conditioning in controlling low- Z impurity transport and optimizing edge solutions for integrated, ELM-stable, high-performance tokamak operation.

A novel approach for simultaneous power exhaust and edge-localized mode (ELM) control is presented in the Experimental Advanced Superconducting Tokamak discharges, which utilize an ITER-like tungsten divertor. Real-time injection of boron (B) powder and neon (Ne) gas overcomes their limitations encountered when used separately. Pure Ne seeding leads to a narrow operational window constrained by core impurity accumulation and H-mode to L-mode back transitions, while pure solid B injection (SBI) is insufficient for effective divertor cooling. In comparison, their combined use achieves a stable, stationary, ELM-suppressed H-mode with adequate power exhaust. This synergistic scenario features partial energy detachment at the outer divertor while maintaining good plasma confinement (H98 similar to 1) with minimal degradation. Two key features of this scenario are: (1) the SBI triggers a persistent Edge Harmonic Mode (EHM), which provides a crucial continuous particle transport channel, preventing Ne and tungsten/molybdenum accumulation without flushing out by ELM, and (2) the B + Ne mixture allows for active optimization of the radiated power profile. Core radiation can be reduced by substituting a portion of the Ne with B, leveraging their complementary non-coronal equilibrium radiation efficiencies. This combined B + Ne injection scheme presents a promising pathway toward integrated core-edge scenarios, offering the potential to minimize total impurity throughput while leveraging an actuator (powder injection) already being considered for ITER.

Recent experiments in EAST have demonstrated the compatibility of tungsten (W) as main wall plasma facing material with high confinement H-mode with low or no boron coverage. The experiments were conducted in plasmas with $q_{95}$ $\approx$ 6.0, which allowed access to both type-I and type-II ELMy H-modes in EAST with a W wall and low normalized input torque similar to ITER. Central electron cyclotron (EC), neutral beam injection (NBI) and lower hybrid waves (LHW) were applied as auxiliary heating in a range of total power injected into the plasma from 3 MW to 5 MW. Small amplitude high frequency type-II ELMs allow maintaining good H-mode energy confinement even when the distance between the separatrix and the main W limiter is as low as 4 cm. The normalized H-mode energy confinement can reach $H_{98}$ factors up to 1.1 for both EC+NBI and EC+LHW power combination in the type-II ELMy H-mode regime. In addition, for these type-II ELMy H-mode conditions, nitrogen puffing from divertor region has been used to achieve partial detachment without significant impact on $H_{98}$ nor on the core W concentration. On the contrary, EAST operation in the type-I ELMy H-mode is strongly affected by the main wall W source. The use of $n$ = 2 resonant magnetic perturbations (RMPs) to achieve type-I ELM suppression in EAST reduces the core W level but at the cost of about 10% reduction on energy confinement. These results complement existing W-wall findings, providing a preliminary foundation from EAST for evaluating the impact of the W first wall in ITER, as proposed in the new ITER baseline, and highlight that the achievement of sufficient ELM control levels with low deterioration of energy confinement is key to minimize this impact.

Helium plasma experiments conducted on different tokamaks all demonstrate that the energy confinement are lower than those of deuterium plasmas under similar operation conditions. However, the origin of the difference in confinement between helium and hydrogen isotopes remains unclear. To understand the influence of helium concentration, which leads to the variation of ion mass and charge number, on energy confinement, new helium experiments were conducted in the EAST tokamak with dominant electron heating and a tungsten divertor in 2025. A notable dependence on both ion mass and charge number is observed in the effective diffusion coefficient, consistent with the trend predicted by the gyro-Bohm scaling ${{\chi }}{}_{{\text{gB}}}^{\text{ }}{\text{ }} \propto {{\rho }}{}_{\text{i}}^{\text{ }}{\text{ }}$ . The higher energy confinement time is observed with a lower ion mass. Conversely, when both the ion mass and charge number are increased, the energy confinement time is found to be similar. Power balance and transport analyses indicate that the ion-electron collision, significantly influenced by the variation of ion mass and charge number, appears to play a dominant role in regulating ion-scale turbulence in helium H-mode discharges. Furthermore, linear electromagnetic gyrokinetic simulation reveals that the linear growth rates of electron temperature gradient (ETG) modes might be reduced by the enhancement of ion-electron energy exchange in L-mode discharges. Consequently, it is crucial to take into account the role of electron-ion collision/energy exchange to understand the influence of ion mass and chargenumber on energy confinement. These results contribute to a better understanding of the transport characteristics in multi-ion-component plasma.

High plasma density operation is crucial for a tokamak to achieve energy breakeven and burning plasma. However, there is often an empirical upper limit of electron density in tokamak operation, namely, the Greenwald density limit [Formula: see text], above which tokamaks generally disrupt. Achieving high-density operation above the density limit has been a long-standing challenge in magnetic confinement fusion research. Here, we report experimental results on the Experimental Advanced Superconducting Tokamak (EAST) achieving line-averaged electron density in the range of (1.3 to 1.65) [Formula: see text], significantly above the typical EAST operational range of (0.8 to 1.0) [Formula: see text]. This is performed with electron cyclotron resonance heating (ECRH)-assisted ohmic start-up and sufficiently high initial neutral density. These experiments are shown to operate in the density-free regime first predicted by a recent plasma-wall self-organization theory. These results suggest a promising scheme for substantially increasing the density limit in tokamaks, a critical advancement toward achieving burning plasma.

The real-time response monitoring model and the physics feature based response optimization model are proposed on EAST using three deep learning models. Compared with the widely used fully connected neural network (FNN) and convolutional neural network-long short-term memory (CNN-LSTM) models, the inverted transformer (iTransformer) model, which uses the attention mechanism to more naturally capture multivariate dependencies, demonstrates the best overall performance. For real-time response monitoring, the iTransformer model is trained on 133157 experimental samples with 4-fold toroidal phase symmetry augmentation. This model establishes a direct nonlinear mapping from resonant magnetic perturbation (RMP) coils and electromagnetic measurement sensors to the plasma response. With an inference time per discharge of less than 10 ms, it can provide real-time response monitoring under flexible three-dimensional (3D) fields. For experimental response optimization, the physics feature based iTransformer model is trained on 12080 filtered experimental samples. This model captures the dependence of the response amplitude on key parameters, including IRMP, phi UL, q95, beta N, ne, Bt, and li. Combined with SHapley Additive exPlanations (SHAP) analysis, it provides interpretable insights into parameter influences and directly guides the optimization of 3D field experimental discharges. This work establishes a comprehensive framework for real-time monitoring and physics feature based optimization of plasma responses based on statistical patterns in large datasets, thereby paving the way for active control of error-field locked mode and edge localized mode (ELM).

Machine learning offers a data-driven approach for rapid pre-shot prediction of key plasma parameters in tokamak experiments. In this study, random forest, support vector regression (SVR), and independent single-output multilayer perceptron (MLP) models were developed to predict macroscopic and dimensionless plasma parameters in EAST under H-mode operation. Ten experimentally accessible parameters-plasma current, toroidal magnetic field, line-averaged density, radiative power loss, confinement enhancement factor, Greenwald density ratio, safety factor, normalized beta, triangularity, and elongation-were used as inputs. Five targets were predicted: normalized ion gyroradius rho*, normalized collisionality nu*, internal inductance li, poloidal beta beta P, and central ion temperature Ti0. To avoid data leakage from correlated samples within the same discharge, a discharge-based splitting strategy was adopted, assigning complete discharges exclusively to either the training or testing set. Performance was evaluated using R2 and mean relative percentage error. SVR provides the most balanced overall performance, achieving high testing accuracy for most targets, especially rho* and beta P. The independent MLP models show strong nonlinear modeling capability, particularly for nu*, whereas RF provides acceptable baseline performance but is limited by weak extrapolation when training and testing distributions differ. Robustness tests with 5% Gaussian input perturbations show that SVR maintains stable predictions for most variables, especially rho*, nu*, and beta P. MLP also remains robust for nu* and beta P, while Ti0 and li are more challenging because of their stronger dependence on current profile, temperature profile, transport behavior, and equilibrium state. Compared with conventional zero-dimensional scaling-law predictions, the machine learning models show improved agreement with experimental measurements and greater nonlinear mapping flexibility. These results indicate that SVR is a suitable and robust model for multi-parameter pre-shot prediction in EAST, while MLP may benefit from larger datasets.

Abstract During the 2025 campaign, helium-3 ( 3 He) minority heating with waves in the ion cyclotron range of frequencies (ICRF) was investigated for the first time on the Experimental Advanced Superconducting Tokamak (EAST). With the lowest available ICRF frequency of f IC = 27 MHz, experiments were conducted at a high toroidal magnetic field of B t = 2.8 T and plasma current I p = 450 kA. To optimize 3 He minority heating, the variation of 3 He concentration was systematically explored. Real-time feedback control of the 3 He concentration was successfully implemented through spectroscopic measurement and closed-loop regulation of the 3 He gas injection, demonstrating the feasibility of the control system functions. The ICRF heating efficiency reached a maximum at a minority concentration of ∼8-9%, with the core electron temperature increasing from approximately 5.0 to 7.0 keV and the ion temperature from approximately 1.3 to 1.9 keV under 2.9 MW of ICRF power. These results are in good agreement with simulations from the two-dimensional full-wave code TORIC. Experiments further indicate that higher plasma density enhances 3He heating efficiency. We also briefly discuss the strategy for future 3 He ICRF experiments on EAST.

This paper reports on the experimental observation and numerical reconstruction of island-like structures in synchrotron imaging emitted by runaway electrons (REs) on Experimental Advanced Superconducting Tokamak. During low-density ohmic discharges with external resonant magnetic perturbations, island-like structures were observed in the synchrotron radiation emitted by REs. Through simulations based on relativistic guiding-center motion equations and a cone radiation model, it was found that these island-like structures in synchrotron imaging are related to the island-like spatial distribution of REs. The island-like spatial distribution might be caused by the sticky regions in the stochastic field. Appropriate projection is necessary for island-like structures in synchrotron imaging. The study provides insights into the behavior of REs in perturbation magnetic fields and offers a sign for the existence of sticky regions within stochastic fields. Besides, discrepancies between simulation and experimental results suggest that non-linear plasma response should be considered. This research improves our understanding of RE behavior in perturbed fields and has implications for the development of effective RE mitigation strategies in large-scale tokamaks.

A primary objective of the experimental advanced super-conducting tokamak (EAST) is to demonstrate steady-state long-pulse high-performance plasma operation for future large-scale devices like ITER and CFETR (Wan et al 2017 Nucl. Fusion 57 102019). The formation of internal transport barriers (ITBs) is one of the key issues to achieve high-performance plasma operation. Optimizing the current density profile is a promising way to improve plasma confinement, which is beneficial to the formation of ITBs. A lot of effort has been dedicated to optimizing the current density profile at EAST over recent years. In this paper, the authors discuss the formation of ITBs leading to a significant improvement of plasma confinement by optimizing the injection time of electron cyclotron resonance heating (ECRH) power at EAST. After ECRH delayed injection, poloidal beta and thermal energy confinement ( H98y2) was observed to increase. The current density distribution and the power deposition distribution were changed. The analysis of the soft x-ray (Xu et al 2020 Phys. Scr. 95 055603) and electron cyclotron emission (Liu et al 2016 Plasma Sci. Technol. 18 1148-54) diagnostics data showed that the structural strength of magnetohydrodynamics was significantly decreased in the core region in this case. The turbulence growth rate calculated by the tokamak global linearized fusion code (Kinsey et al 2008 Phys. Plasmas 15 055908) also shows that trapped electron mode turbulence is stabilized.

Achieving sustained high plasma confinement both in the core and at the edge, without crashes due to edge instabilities, is highly desirable for efficient fusion energy production in a tokamak device. However, control of edge instabilities usually results in core plasma confinement degradation. Here we demonstrate a new improved core confinement plasma regime in EAST, characterized by a sustainable internal transport barrier (ITB) and effective control of edge instabilities, both achieved by applying small resonant magnetic field perturbations (RMPs) at just 0.1% of the equilibrium field amplitude. The key is to localize the perturbations at the edge, which can be effectively realized by use of RMPs with high toroidal mode numbers. Nonlinear interaction between the perturbed magnetic fields and plasma, especially tungsten impurities, on different scales plays a key role in lowering the threshold heating power by a factor of 2 for triggering the formation of an ITB to improve core confinement. More importantly, the improved core confinement can be actively controlled by tuning the perturbed field profile. This provides an attractive method and a physical basis for an integrated solution of keeping high core confinement during edge stability control for a future fusion reactor.

In the Experimental Advanced Superconducting Tokamak (EAST), a transition from electromagnetic to electrostatic turbulence is observed in the pedestal region as plasma density ramps up. This transition is manifested by the suppression of magnetic fluctuations and the presence of broadband electrostatic turbulence. The frequency domain of the electrostatic turbulence is typically beyond 300 kHz. It leads to a rapid build-up of density gradient and a sharp degradation of energy confinement. By reducing the gas puffing rate, a prolonged intermediate transition phase is observed, and the confinement improved with increasing density in this phase. The emergence of broad electrostatic turbulence is associated with the enhanced turbulence control parameter α _t , together with a weaken edge radial electric field. Furthermore, the impact of the turbulence transition to scrape-off layer (SOL) transport is evaluated. Measurements suggested that radial particle flux and intermittent structures are strengthened after the transition. Moreover, the profiles in the far SOL are broaden with increase of α _t . The properties of the magnetic fluctuations are consistent with the nature of the magnetic coherent mode previously observed in EAST, while the broad electrostatic turbulence is proposed to be an ion temperature gradient mode by the Gyrokinetic Electromagnetic Numerical Experiment simulation. It is clarified that the energy confinement degradation in high-density regimes is primarily driven by the broadband turbulence rather than divertor detachment. These findings advance our understanding of high-density H-mode plasmas and provides additional insights into the interplay between edge turbulence and global confinement properties.

After installation of tungsten divertor in EAST, impurity accumulation of tungsten ions has been frequently observed in H-mode discharge with internal transport barrier (ITB) due to an enhancement of the impurity confinement inside the ITB. A strong plasma cooling induced by the tungsten impurity ions caused a collapse of the ITB formation. To study the impurity transport in high β _N discharges with ITB, temporal behaviors and radial profiles of spectral lines emitted from low- and high-Z impurity ions were analyzed. Line emissions from moderately ionized ions, e.g. O ^6+ , Fe ^17+ and Mo ^25+ locating outside the ITB ( ρ ⩾ 0.4), maintain low intensities and remain unchanged during the ITB formation. However, line emissions from highly ionized high-Z impurity ions such as Fe ^22+ , Cu ^21+ , Cu ^25+ , Mo ^30+ and W ^26+ -W ^37+ locating inside the ITB ( ρ < 0.4) are strongly influenced by peaking effects of electron density and ion temperature profiles. The impurity screening effect due to the ion temperature peaking is dominant during T _i -ITB phase because the line intensities of high-Z impurity ions are reduced and the radial high-Z impurity profiles are flattened. In contrast, during n _e -ITB phase accompanied by electron and ion temperature ITB, an increase in the electron density gradient from R / L _n _e = 3.4–4.9 results in a significant increase in the high-Z impurity density, and leads to the impurity accumulation. Statistical analysis on the tungsten impurity density ( I _W-UTA / n _e ) with toroidal rotation velocity ( V _t0 ) and ion temperature gradient ( R / L _T _i ) suggests that the tungsten impurity accumulation can be effectively mitigated and the I _W-UTA / n _e can be reduced to below 18 (phs·m·s ^−1 ·Sr ^−1 ), when R / L _n _e < 3.4, V _t0 < 150 km·s ^−1 and R / L _T _i > 2.5. During two ITB phases, however, low-Z impurity ions like O ^7+ locating at edge of the ITB appear to be sensitive to only the electron density gradient.

Controlling the splitting divertor heat flux caused by resonant magnetic perturbations (RMPs) is a topic of concern for fusion devices. As a fundamental prerequisite, it is necessary to understand the characteristics of the heat flux distribution under the applied RMP field, which will be studied in this paper. The nonlinear phenomenon of strike point splitting was found to be strongly dependent on the plasma response under RMP. These splitting heat flux distributions are qualitatively explained by the simulated magnetic footprints. The RMP phase scanning experiment shows that scanning in a certain range of relative phase can maintain good edge-localized mode (ELM) mitigation and simultaneously sweep the striations of heat flux on divertor target. Additionally, even in upper single null (USN) configurations, heat stripes are observed on the lower outer divertor (LO-div), attributed to RMP-induced additional magnetic connections to the LO-div, as confirmed by magnetic topology simulations. A dedicated investigation into the impact of the discrepancy between lower and upper separatrix radii mapped to the low field side mid-plane ( dRsep) reveals its significant influence on both ELM control and heat flux distribution. Within a certain range of dRsep, the heat flux distribution is improved while ELM suppression is maintained. These findings contribute to divertor heat flux understanding under RMP conditions in tokamak operations.

This paper validates for the first time the predictive capability of the DYON code for plasma initiation in EAST, which has metallic wall, superconducting coils and conventional tokamak shape, like ITER. The model accurately reproduced the operating spaces of loop voltage and prefill gas pressure for ohmic discharges, demonstrating its validity in predicting the required operating parameters for successful inductive plasma initiation in EAST. The role of wall conditioning on plasma initiation was investigated with the newly developed physical sputtering models of Boron and Lithium. In EAST experiments, it was observed that the discharges after boronisation of the wall are much more vulnerable to plasma burn-through failure than after lithium-coating. The simulation results revealed that despite the similar physical sputtering yield in Boron and Lithium, the radiative energy loss rates for the boron-coated wall are significantly higher than those for the lithium-coated wall, due to the much higher radiative power coefficients of Boron. Parametric scans of initial Boron content in ohmic discharge at the typical prefilled gas pressure in EAST (0.8 mPa) showed that even 1.5% of Boron content in the prefilled gas, possibly remaining after boronisation of the wall, could lead to excessive radiation energy losses and failure of plasma burn-through. For successful plasma burn-through with 1.5% initial boron content, the modelling indicates 10 kW absorption of EC power is required, and it increases with more initial boron e.g. 50 kW for 3% initial boron content.

During the discharge process, thermal events are common and destructive in Tokamak experiments. High energy particles in plasma can collide with device components such as limiters and diverters, which can lead to overheating, material cracking, and even damage to the structure of the device. Therefore, we need efficient detection methods to monitor thermal events in real-time. To address the challenge of identifying thermal damage to internal components of the first wall during the Experimental Advanced Superconducting Tokamak (EAST) discharges, we introduce the YOLOv8 model for hotspot detection on EAST and present the customized EAST- You Only Look Once (YOLO) algorithm, derived from an enhanced YOLOv8 framework. YOLOv8, known for its strong performance in real-time object detection, serves as a robust base model for this task. However, its performance on small object detection, such as early-stage thermal damage, is limited. Improvements are needed for specialized tasks, particularly for early warning and precise identification of small internal component damage during the initial stages of EAST discharges. we enhance the YOLOv8 model by incorporating specialized layers for detecting small targets and integrating the CBAM attention mechanism. These adjustments result in a network model capable of sensitively detecting internal component damage in EAST. Experimental results demonstrate that EAST-YOLO surpasses several versions of traditional YOLOv8 models in model evaluation metrics, achieving a precision of 97.5%, mAP50 of 97.7%, and a Recall of 94.0%. This problem- oriented approach significantly improves the operational safety and stability of the EAST device by enabling early detection of thermal events. The AI-based detection method provides a new solution to safeguarding the fusion device, while also offering potential avenues for integrating artificial intelligence technologies into EAST feedback control and operating systems in the future.

High plasma density operation is crucial for a tokamak to achieve energy breakeven and a burning plasma. However, there is often an empirical upper limit of electron density in tokamak operation, namely the Greenwald density limit $n_G$, above which tokamaks generally disrupt. Achieving high-density operations above the density limit has been a long-standing challenge in magnetic confinement fusion research. Here, we report experimental results on EAST tokamak achieving the line-averaged electron density in the range of 1.3 $n_G$ to 1.65 $n_G$,while the usual range in EAST is (0.8-1.0)$n_G$. This is performed with ECRH-assisted Ohmic start-up and a sufficiently high initial neutral density. This is motivated by and consistent with predictions of a recent plasma-wall self-organization (PWSO) theory, that increasing ECRH power or pre-filled gas pressure leads to lower plasma temperatures around divertor target and higher density limits. In addition, the experiments are shown to operate in the density-free regime predicted by the PWSO model. These results suggest a promising scheme for substantially increasing the density limit in tokamaks, a critical advancement toward achieving the burning plasma.

Experimental observations in EAST tokamak have revealed a notable suppression of tungsten accumulation during the lower hybrid wave (LHW) injection in the neutral beam injection (NBI)-heated H-mode plasma. The variation in tungsten concentration during the LHW phase is quantified using the intensity of tungsten unresolved transition array (W-UTA), as measured by the extreme ultraviolet spectrometer. After the LHW is turned on, the tungsten concentration, C _W = n _W / n _e , decreased by approximately 45%, from 9.5 × 10 ^−5 to 5.3 × 10 ^−5 . Additionally, the peaked tungsten profile is flattened, with the peak position shifted outward, implying a significant alteration in tungsten transport. The observed results suggest that reduced toroidal rotation and increased electron temperature are responsible for the variation in tungsten transport during the LHW heating. Modeling using a simulation code of Transport in Gyrokinetic Plasmas with Rotation and Optimization (TGYRO) indicates that, after the LHW injection, the turbulent diffusion of tungsten ions is enhanced and the neoclassical convection is weakened. Moreover, comparisons of the transport coefficients of tungsten ions under different plasma parameters reveal that the LHW heating enhances the turbulent diffusion of tungsten ions by increasing the electron temperature gradient. It also decreases the plasma toroidal rotation velocity, which in turn reduces the inward neoclassical convection of tungsten ions. These findings provide a feasible solution for the tungsten accumulation induced by the NBI heating, supporting EAST in achieving long-pulse high-performance plasma discharge. This work offers an important reference for the operation of ITER and the design of future fusion reactors.