Recent Publications

A novel high-performance confinement regime in a tokamak driven by tearing mode locking and rotation shear

Qinghao Yan, Da Li, Deliang Yu, B Li, Shuosu Yang, Rui Ke, ShaoBo Gong, Yihang Chen, Junzhao Zhang, Yi Zhang, et al.

In magnetically confined fusion plasmas, tearing mode (TM) locking is conventionally regarded as detrimental, often halting rotation and triggering disruptions. Contrary to this expectation, we demonstrate that locked TMs can enhance confinement by inducing an internal transport barrier (ITB). Specifically, the interaction between a locked outer plasma region and neutral beam injection (NBI)-driven core rotation generates a broad and intense toroidal velocity shear layer, which suppresses turbulence and improves core confinement. Beam emission spectroscopy and charge exchange recombination spectroscopy confirm a substantial reduction in turbulent transport and a rise in core ion temperature exceeding $T_i > 9\,\mathrm{keV}$. This phenomenon—termed the Locking-Induced Flow and Transport (LIFT) mode—transforms a traditionally disruptive mechanism into a confinement enhancement strategy.

Progress in reactor-core fusion technologies presented at the 30th IAEA Fusion Energy Conference 2025

Jiming Chen, Jianbao Wang, Qixiang Cao, Shen Qu, Zongyu Yang

This paper provides an overview of the recent progress in reactor-core fusion technologies within the vacuum vessel presented at the 30th International Atomic Energy Agency (IAEA) Fusion Energy Conference held in October 2025 in Chengdu, China, including plasma control, neutronics and nuclear data, tritium technologies, design and development of in-vessel components, related materials and intense neutron sources for material irradiation. Plasma control focusing on achievements for ITER and major fusion research devices are covered and the application of digital twin and Artificial Intelligence are highlighted. Research and Development (R&D) results from the International Thermonuclear Experimental Reactor (ITER) and current devices to future Demonstration Fusion Power Plant (DEMO) reactors are summarized, particularly new findings, data validation, design and analysis tool improvement, properties assessment and technical solutions for critical issues.

Effects of strike point location on upstream electron density in HL-3 by SOLPS-ITER modeling

Haopan Du, Chen Zhang, Na Wu, Hailong Du, Qingquan Yang, Xin Lin, Kaiyang Yi, Xuele Zhao, Yilin Wang, ChangJiang Sun, et al.

Experiments in L-mode on the HL-3 tokamak show that, under similar discharge conditions, the divertor strike point (SP) location significantly affects upstream edge and divertor plasma parameters. When the SP is located at the corner rather than the bottom, the upstream electron density (at the outer midplane ne,OMP) is substantially higher. In this work, simulations were conducted using the edge plasma code SOLPS-ITER with the experimental parameters, which well reproduce the experimentally observed trends and provide insight into the mechanisms by which the SP location influences ne,OMP. The results show that when the SP is located at the corner, D atoms reflected from the right baffle accumulate in the downstream divertor region, enhancing the flux of D atoms entering the core region near the X-point. This strengthens the ionization source in the core, promotes radial transport of D+, and ultimately increases nD+,OMP. Moreover, the simulations indicate that the difference in ne,sep between the corner and bottom cases increases with input power. Besides stronger core ionization enhancing radial D+ transport in the corner case, two additional effects contribute: (1) the peak D+ flux incident on the corner divertor target (ϕ_(D⁺,OT)^peak) sputters more C impurities, which are then transported upstream, resulting in a higher upstream carbon density nC,OMP and consequently increasing ne,OMP; (2) in the corner case, the D+ radial flux through the outer divertor entrance to the downstream region is reduced, further increasing the difference in ne,OMP.

Neural network surrogate model for drift-kinetic energy perturbation for internal kink instability

Shuo Jiang, Lina Zhou, Yueqiang Liu, Yan Qiao, Zongqiang Li, Guangzhou Hao, Yong Wang, Xu Yang, Shuo Wang, Yutian Miao, et al.

Multi-layer perceptron based neural networks (NNs) are trained to predict the ideal internal kink (IK) instability in tokamak plasmas, with particular emphasis on the drift-kinetic energy perturbation associated with thermal particles. The NN surrogate model is trained and validated on a semi-analytic database and tested on both held-out semi-analytic data and five EU DEMO design cases. A well-trained NN predicts the amplitude and phase of the perturbed drift-kinetic potential energy with the coefficient of determination (R²) reaching 0.88, and 0.98, respectively. The NN also achieves high R² values of 0.94 and 0.92 for the fluid growth rate and fluid perturbed potential energy, respectively. These results, together with low mean absolute errors, demonstrate the effectiveness of the semi-analytic training approach and establish a foundation for real-time IK stability monitoring and sawtooth period estimation in tokamak experiments. The physics-prediction reliability is further confirmed through case study of an EU DEMO negative-triangularity equilibrium, where the NN successfully reproduces the theoretical linear relationship between the fluid growth rate and the fluid perturbed potential energy and captures the dependence of IK instability on multiple plasma parameters. Collectively, these results confirm that the trained NN effectively captures essential physical characteristics of the IK instability.

Radial propagation of ELM-filaments and boundary turbulence characteristics with divertor heat flux mitigation in the small ELM regime on HL-2A

Ting Wu, Lin Nie, Zhanhui Wang, Jinming Gao, Yukun Shu, Yi Zhang, Yu Zhou, Yujie Zhou, Yiren Zhu, Yihang Chen, et al.

This paper investigates the radial propagation of ELM-filaments and boundary turbulence characteristics with divertor heat flux mitigation in the small ELM regime on HL-2A. Divertor nitrogen seeding mitigates divertor heat flux while maintaining high plasma performance (H_98>1.0 and β_N>2.0) and negligible degradation of total stored energy and energy confinement time. In the pedestal, the pressure gradient and ELM amplitudes decrease after divertor heat flux mitigation. The magnetic perturbation diminishes while electrostatic turbulence increases, with poloidal velocity shear remaining unchanged in the pedestal. Turbulent particle flux and perpendicular heat flux decrease in the far SOL region. The ELM-filament burst rates and conditionally averaged amplitudes from the pedestal to the near SOL and the far SOL are systematically investigated before and after heat flux mitigation. ELM-filaments burst rates increase and ELM-filaments amplitudes decrease in the pedestal after heat flux mitigation. The evolution of near-SOL filaments could be partially reflected by that at the outer divertor target (ρ≈0.99-1.09). At the divertor target, the filaments burst rates are larger at ρ≈1.01 but become smaller at ρ≈1.09 The burst rates of filaments in the far SOL (ρ≈1.16) decrease dramatically, resulting in a significant reduction in filament transport after divertor heat flux mitigation. These results provide improved insight into the radial propagation of ELM-filament, and benefit the understanding of power deposition on the first wall in the small ELM regime.

Design and commissioning of a tangential Gas Puff Imaging diagnostic on the HL-3 tokamak

W.C. Wang, Z.H. Huang, Z.B. Shi, Y. Hu, J. Cheng, K.Y. Yi, W.M. Xie, Y. He, N. Wu, W. Zhao, et al.

Design of a Pop-Up Divertor Langmuir Probe System for the HL-3 Tokamak

Jichan Xu, Shuang Xu, Qinglai Qiu, Yunfei Dai, Yang Zhang, Xiaodong Zhang, Tengfei Sun, Jinming Gao, Yuxiang Liu, Zhihui Huang, et al.

The influence of magnetic geometry on ion temperature gradient modes in the CFQS

Moshan Li, Jie Huang, Yuhong Xu, Akihiro Shimizu, Mitsutaka Isobe, Shoichi Okamura, Haifeng Liu, Xian-Qu Wang, Y Q Wu, X Zhang, et al.

A key feature of the stellarator is its inherent three-dimensional magnetic configuration, which leads to numerous differences in plasma physics between stellarators and tokamaks. The influence of this three-dimensional geometry on ion temperature gradient (ITG) modes in the Chinese First Quasi-axisymmetric Stellarator (CFQS) is investigated through linear gyrokinetic simulations using the gyrokinetic Vlasov (GKV) code. Across different radial and toroidal positions, the ITG mode exhibits a strong toroidal dependence and a weak radial dependence. The eigenmode localization is primarily determined by the perpendicular wavenumber k_{\perp }^{2} and the normal curvature {\kappa }_{y}: the ITG mode peaks where {\kappa }_{y} is negative and k_{\perp }^{2} is minimized. For ITG modes with k_{x}=0, the growth rate mainly exhibits a toroidal dependence. As the toroidal angle ϕ increases, the peak growth rate decreases and shifts to the low k_{y}{\rho }_{i} region. For modes with k_{x}\neq 0, the most unstable modes acquire a finite k_{x} at all three radial positions for \phi ={90}^{\circ }, and at ρ=0.8 for \phi ={45}^{\circ }. This finite k_{x} alters the value of k_{\perp }^{2} and modifies its distribution along the field line, thereby changing the eigenmode structure. By combining the eigenmode structure with the overall distributions of magnetic geometry parameters, the variation of the growth rate across different spatial positions can be qualitatively estimated.

Stress corrosion mechanism of CLF-1 RAFM steel in flowing Pb-17Li at 450℃: synergistic effects of tensile stress and microstructural evolution

Zhenchao Sun, Yu Guo, Xiujie Zhang, Teng Zhang, Wei Qian, Yao Zhao, Lei Wang, Xinting Lv, Yiming Wang, Zhengdong Li

The compatibility of RAFM steel with liquid Pb-17Li is a critical concern for its application as a structural material in liquid metal blankets of fusion reactors. This study investigated the stress corrosion mechanism of CLF-1 RAFM steel under a tensile stress of 250 MPa for 5000 hours in flowing Pb-17Li at 450 °C. The results indicate that stress-induced strain disrupts the continuity of the protective oxide layer, shortening the incubation period and increasing corrosion initiation sites, thereby intensifying the corrosion severity and susceptibility. The corrosion morphology is jointly influenced by the flow velocity of Pb-17Li and the steel’s microstructure. At high flow velocities, the martensitic lath structure is eroded, resulting in an etched surface. Conversely, lower velocities preserve the more corrosion-resistant laths, producing lamellar corrosion structures. Furthermore, shear stress promotes microstructural coarsening, which eliminates certain grain and sub-grain boundaries. Although this locally improves corrosion resistance, the resulting long and straight grain boundaries instead facilitates the penetration of liquid metal, thereby accelerating corrosion failure. These findings clarify the stress corrosion mechanism of RAFM steel in Pb-17Li and provide new insights for the modeling of liquid Pb-17Li corrosion and design of liquid metal blankets.

Analysis of neutron emission during NBI–ICRF synergistic heating in EAST high neutron rate high β p discharges

Andong Xu, Mingyuan Xu, Yunhe Li, Tao Yu, Jiayi Zhang, Yubo Zhang, Yongqiang Zhang, Chenyu Pan, Baolong Hao, Pan Li, et al.

This paper reports the analysis of neutron emission characteristics in high poloidal beta (β p ) discharges on the EAST tokamak, where a record fusion neutron rate of S n = 3.9×10 14 s -1 was achieved with β p ∼2.8, β N ∼2.2, and H 98,y2 ∼1.3. Statistical analysis reveals that while ion cyclotron range of frequencies (ICRF) heating significantly boosts both the neutron rate and plasma stored energy, the neutron rate scales sub-linearly with neutral beam injection (NBI) power (S n ∝P NBI 0.92 ). Interpretive TRANSP simulations demonstrate that the NBI--ICRF synergistic effect directly contributes approximately 30% to the total neutron rate through the formation of a high-energy fast-ion tail. However, this enhancement is partially offset by NBI-induced profile degradation, including fuel dilution, impurity accumulation, and core electron temperature reduction. The generation of the fast-ion tail in velocity space is validated by multi-sightline neutron emission spectroscopy. Furthermore, orbit topology analysis using the ORBIT code reveals that the synergistic effect drives suprathermal fast ions into smaller orbits, such as stagnation orbits, leading to a spatial redistribution of fast ions and a consequent peaking of the neutron emissivity profile. These findings provide critical insights into the complex interplay between auxiliary heating, fast-ion behavior, and neutron emission, offering valuable insights for achieving higher beam--thermal fusion rates in future deuterium--deuterium and deuterium--tritium experiments.

Integrated modeling study on sustaining high-current reversed magnetic shear configurations in the HL-3 tokamak via combined NBI and ECW

Hai Li, Yijun Zhong, Junbo Zhang, Shoulong Xu, Ziqi Fan, Haozhen Gu, Xinliang Xu, Shuo Wang, Xueyu Gong

Investigation on fast ion losses induced by neoclassical tearing mode in HL-3 high- β plasmas

Zhiyi Yin, Min Jiang, Baolong Hao, Ruirui Ma, Yuxiao Han, Yi Liu, Linge Zang, Sen Xu, Shuosu Yang, Deliang Yu, et al.

The fast ion losses (FIL) induced by m = 2 , n = 1 neoclassical tearing mode (NTM) are investigated dedicatedly in HL-3 high- β N plasmas, where m ( n ) is the poloidal (toroidal) mode number. The spectrogram of FIL fluctuation measured by FIL detector (FILD) shows a dominant characteristic frequency of the NTM rotation frequency, indicating that the NTM rotation modulates the fast ion (FI) loss. The lost FI population peaks at pitch-angle of 112 ∘ − 118 ∘ and energy ranges of 40 − 60 keV , consistent with the full energy of neutral beam. The intensity of the FILD signal is proportional to the square of island width, implying dominantly convective FIL. In addition, the NTM island structure passes through the lost orbit of the trapped FI, which is calculated backward in time from the FILD measurement, suggesting there exists an interaction between FI orbit and the NTM island. The calculation using ORBIT code reveals the wave-particle resonant condition is fulfilled between the NTM and the trapped FIs, further confirming the convective loss mechanism. The interaction primarily affects the toroidal canonical momentum rather than the energy of FIs due to the relatively low NTM frequency. These results enhance our understanding of NTM induced FI transport in high- β plasmas, providing valuable insight for optimizing the high-performance operation scenario for future fusion reactors.

Impact of low-Z impurity injection on the post-disruption runaway electron current in the J-TEXT tokamak

Wei Yan, Zhongyong Chen, Xun Zhou, Yuan Sheng, Yuwei Sun, Kaiyin Peng, You Li, Zhifang Lin, Nengchao Wang, Zhoujun Yang, et al.

Major disruptions in tokamak plasmas pose a severe threat to the safe and stable operation of the device, and the runaway current formed by high-energy runaway electrons is one of the hazardous consequences. Massive impurity injection serves as a primary means of mitigating runaway current, where high-Z impurities can effectively dissipate it, while low-Z deuterium enables benign termination of runaway current. On the J-TEXT tokamak, experiments on mitigating runaway current have been conducted using shattered pellet injection (SPI) and massive gas injection (MGI) with low-Z deuterium/neon mixtures. The results indicate that both SPI and MGI with large amounts of low-Z mixed impurities can extend plateau duration of runaway current and reduced RE energy following mixture impurity injection, and a higher proportion of deuterium in the mixture impurities plays a dominant role. Furthermore, based on the injection characteristics of SPI and MGI, it is inferred that penetration depth of mixed impurities in the runaway current core region and a larger injection quantity are main factors influencing the phenomenon of runaway current. These findings provide important references for extrapolating mitigation schemes for runaway current to future tokamak fusion reactors.

Design of a high-β N operation scenario for the HL-3 tokamak

Guanqi Dong, Guangzhou Hao, Jiaxian Li, Yiren Zhu, Shuo Wang, Yunpeng Zou, Zhuo Wang, Zhengji Li, Hailong Du, Yujie Zhou, et al.

Future fusion reactors require high normalized beta ($\beta_N$) operation to achieve economically attractive power density. High-$\beta_N$ operation is also a key objective for the HL-3 tokamak. Designing a stable and robust operational scenario prior to experiments is therefore essential for the successful execution of high-$\beta_N$ campaigns. This study presents the design of an H-mode scenario with $\beta_N > 3$ for HL-3 at a plasma current of $2$~MA and a toroidal magnetic field of $2.2$~T. A multi-level integrated modeling approach is employed, progressing from macroscopic parameter determination and zero-dimensional scans to 1.5-dimensional dynamic scenario construction. Using the OMFIT framework, modules including EFIT, ONETWO, TGYRO, and EPED1-NN are coupled to obtain self-consistent plasma equilibria and radial profiles. Two heating schemes are considered: pure neutral beam injection (NBI, $10$~MW) and NBI combined with electron cyclotron resonance heating (ECRH, $8$~MW + $2$~MW). Both scenarios achieve $\beta_N > 3$, with a plasma stored energy of approximately $2.9$~MJ and a confinement factor $H_{98} \approx 1.8$. A complete discharge waveform is developed, covering null-field breakdown, current ramp-up, and flat-top phases. Divertor heat flux analysis using SOLPS-ITER indicates that without impurity seeding, the heat load on the lower divertor target exceeds the material limit ($7$~MW/m$^2$); neon seeding at $\sim 1\times10^{20}$~s$^{-1}$ can reduce the peak heat flux to below $6$~MW/m$^2$, approaching detachment conditions. Magnetohydrodynamic stability analysis confirms resilience to vertical displacement events (VDEs), resistive wall modes (RWMs), and neoclassical tearing modes (NTMs), though edge localized modes (ELMs) are expected due to peeling-ballooning instability, requiring active mitigation. This work provides a physics foundation and engineering roadmap for high-$\beta_N$ experiments on HL-3.

Surrogate model of transport quantities of gyrokinetic simulations on HL-2A

Hongjian Zhao, Shengming Li, Jiquan Li, Yixiong Wei, Yong Xiao

Artificial intelligence techniques, particularly machine learning, are increasingly being employed to reduce the computational cost of high-fidelity fusion simulations. In this work, nonlinear gyrokinetic simulations are performed to systematically scan the plasma gradient parameter space relevant to HL-2A plasmas, generating a database of electrostatic drift-wave turbulence for the development of transport surrogate models. A local surrogate model for turbulent transport is first constructed using a feedforward neural network (FNN) trained on 4267 nonlinear local simulations, achieving coefficients of determination of up to R2 ≃ 0.9 for key transport coefficients. Building upon this, a second-stage global surrogate model based on support vector regression (SVR) is developed to map local transport predictions to global turbulent transport, trained on 40 global gyrokinetic simulations. The resulting global surrogate model is subsequently coupled to an integrated transport solver, which successfully reproduces both the temporal evolution and steady-state plasma profiles in good agreement with experimental observations. This study provides a practical workflow for predicting global turbulent transport using local surrogate models and machine learning techniques. The proposed approach offers an accurate and computationally efficient strategy for integrated transport simulations, with potential applications in scenario development and control-oriented real-time simulations for present and future fusion devices.

High-heat-flux performance of monoblock target prepared with advanced W-K plate

Fan Feng, Youyun Lian, Jianbao Wang, Jiupeng Song, Mengxia Liang, Yuzhong Jin, Xiang Liu

Potassium-doped tungsten (W-K) is a promising plasma-facing material because nanoscale K bubbles may improve microstructural stability without introducing solid second phases. In this work, large-scale rolled W-K plates containing ~90 ppm K were fabricated by powder metallurgy, hot rolling and stress-relief annealing, and were machined into ITER-like water-cooled monoblock mock-ups. The rolled plates showed a tensile strength of 1225 MPa at 50 °C , ductility exceeding 20% at 200 °C, and a recrystallization temperature of ~1500 °C . Transient electron-beam thermal-shock tests on the RD-TD plane demonstrated that the as-rolled and 1400 °C -annealed W-K remained crack-free after 100 pulses of 1 ms up to 0.66 GW/m², whereas specimens annealed at ≥1500 °C exhibited reduced cracking thresholds. Under steady-state high-heat-flux fatigue, W-K monoblocks maintained structural integrity at 20 MW/m² for up to 1500 cycles, although surface roughening, intergranular fissures and local melting developed with increasing cycle number. At 25 MW/m², severe roughening/erosion occurred after 500 cycles as the apparent surface temperature exceeded 2300 °C . A key mechanistic finding is that crack density and crack depth are governed by different factors: crack density increased mainly with accumulated thermal cycles, whereas crack depth was controlled predominantly by peak surface temperature. Stable K-bubble dispersion is suggested to retard grain-boundary migration and suppress microcrack nucleation, contributing to the high thermal-shock and HHF tolerance of rolled W-K monoblocks.

Edge turbulence spreading and blob transport broaden the heat flux width approaching the density limit

Ting Wu, Patrick H Diamond, Lin Nie, Rui Ke, Zhipeng Chen, Qinghu Yang, Wenjing Tian, Zhoujun Yang, Zhongyong Chen, Min Xu

bstract This paper investigates how edge turbulence spreading and blob transport broaden the heat flux width in Ohmic-plasma approaching the operational density limit of the J-TEXT tokamak. At the plasma edge, E_r×B shear flow collapses while turbulent transport and spreading are significantly enhanced when approaching the density limit. The heat flux widths correlate positively with edge radial flux of turbulence internal energy through the LCFS as well as the energy production ratio (the ratio of turbulence spreading from the edge into the SOL to the net local production of turbulence in the SOL). The energy production ratio model combining experimental data shows that turbulence spreading at the LCFS is likely the origin of the SOL turbulence. The mechanism of the heat flux width broadening in high density operation may be the stronger edge turbulence spreading across the LCFS to increase the SOL turbulence as the plasma approaches the density limit. How blob transport (turbulent particle flux and/or radial flux of turbulence internal energy) broadens the heat flux width is investigated in detail. The blob-induced turbulent particle flux fraction (Γ_blob⁄Γ_total ) is 0.2–0.4 while blob-induced turbulence spreading fraction (〖Sp〗_blob/〖Sp〗_total) is 0.5–0.9, suggesting that blob-induced spreading is more important than blob-induced turbulent particle flux. Blobs with larger radial scales induce stronger edge spreading into the SOL, thus dominating the SOL turbulence. These results suggest that edge turbulence spreading and blob transport play crucial roles in broadening the heat flux width as the plasma approaches the density limit.

Integrated modelling of electron cyclotron wave control of core impurity accumulation in HL-3 tokamak

ziqi Fan, Yijun Zhong, Junbo Zhang, Shoulong Xu, Hai Li, Haozhen Gu, xinliang xu, Shuo Wang, Xue-yu Gong, Wulyu Zhong

Core impurity accumulation can severely degrade plasma confinement and may even lead to disruptions; therefore, its mitigation is essential for achieving stable high-performance operation in tokamak devices. In this work, based on the OMFIT integrated modelling platform, the effects of electron cyclotron waves (ECWs) on core tungsten impurity transport are investigated in the HL-3 tokamak under an NBI-heated background plasma. The simulation results show that, in the pure NBI case, the W impurity density profile exhibits pronounced central peaking, whereas ECW injection can effectively mitigate tungsten accumulation, with the control efficiency depending strongly on the power deposition location, injected ECW power, and current-drive scheme. Near-axis ECW deposition at ρ ≈ 0.1 and off-axis deposition at ρ ≈ 0.5 are found to be favorable impurity control schemes, whereas intermediate-radius deposition at ρ ≈ 0.3 provides a relatively weak suppression effect. The two effective deposition regions correspond to different dominant mechanisms: near-axis ECW deposition mainly alleviates W peaking by enhancing core turbulent transport and weakening inward convection, whereas off-axis deposition suppresses W accumulation primarily by flattening the main-ion density profile and thereby reducing the density-gradient-driven neoclassical inward pinch. For near-axis ECW deposition, the current-drive scheme has a significant influence on W transport. In the counter-current drive case, the increase in the local safety factor is accompanied by enhanced neoclassical inward convection, resulting in only limited mitigation of W peaking; by contrast, when ECW is deposited farther off axis, the W transport behaviour is only weakly sensitive to the current-drive scheme. These results provide a physics basis for understanding impurity transport under high-power NBI heating and offer useful guidance for impurity control in the HL-3 tokamak.

Integrated framework for unstable event identification of disruption on the HL-3 tokamak

Junru Wen, Yi Yu, Zongyu Yang, Yi Ling, B Li, Yihang Chen, Da Li, Wulyu Zhong

For tokamaks like the HL-3, which operates at reactor-grade parameters, major disruptions under high-performance conditions are intolerable. High-parameter tokamaks require not only algorithms and engineering solutions for disruption mitigation but also systematic analysis of disruption causes. Traditional disruption analysis relies on various diagnostic data, requiring both temporal evolution analysis and diagnostic cross-channel comparison, which demands substantial expert knowledge and manual effort. To address these limitations, a Plasma Event Identification System have been developed to detect key plasma events in the HL-3 tokamak. This system integrates artificial intelligence, threshold-based judgments to identify disruptions and pre-disruption plasma events including VDE, MHD instability, low-q disruption and high density. The performance of each module in the system was evaluated: The disruption identification module achieved 99% accuracy; The disruption time identification module reached 95% accuracy; The tearing mode and locked mode identification attained 95% overall accuracy, demonstrating the system’s robust performance. The system enables rapid statistics and analysis for HL-3 experiments, including disruption cause analysis based on expert-logic judgments and 200 ms pre-disruption event-chain. Furthermore, this system has been deployed in the HL-3 plasma display for disruption analysis. The analysis results are also upload to the Dig Data analysis database to support the research needs of engineering and physics researchers.

Magnetic topology discontinuous changes and discrete shifts of divertor heat flux induced by edge-localized modes in the HL-2A tokamak

Jinming Gao, Xiao Lan Zou, Guoliang Xiao, Wulyu Zhong, Tianbo Wang, Dongmei FAN, Tengfei Sun, Y B Dong, Min Jiang, Zhou Yulin, et al.

Magnetic reconnection is a fundamental physical process in plasma physics. Edge-localized mode (ELM) physics and its impact on divertor heat loads are critical issues for fusion plasmas. We report the first observation of discontinuous changes in magnetic topology—manifested as sudden shifts of the divertor heat load peak—during ELM crashes in tokamak plasmas. Nonlinear ELM evolution drives magnetic reconnection, generating energetic electrons that form helical current filaments. These filaments induce an abrupt poloidal extension of topological fractures, evidenced by non-gradual changes in the poloidal correlation length of ELM-induced magnetic perturbations. These results present a new aspect of ELM physics, which should be taken into account in the modelling. Also our findings on nonlinear interactions of electron dynamics with the magnetic field topology reveal previously unexplored mechanisms governing reconnection physics.