Recent Publications

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.

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.

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.

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.

Machine learning aided neutron yield for dud detection based on JET and TFTR Deuterium-Tritium plasmas

Lidia Piron, Alessandro Pau, Nicolò Ferron, Eric Fredrickson, Olivier Sauter, Matteo Baruzzo, Clive D Challis, Remi Dumont, Dirk Van Eester, Michael Fitzgerald, et al.

as it indicates fusion performance. To optimize Tritium consumption and limit neutron activation, a support function included in the plasma control system called a dud detector will trigger an alarm if the plasma fails to achieve expected fusion performance. This function has been developed and routinely employed at JET during DT campaigns. This study presents machine-learning methods based on a surrogate model of the neutron rate, which can be used as an advanced dud detector. In preparation for DT operations in BEST, HL-3, ITER, and SPARC, we investigate the portability and inherent limitations of these ML methods by analysing similar DT experiments conducted at TFTR.

Linear and quasi-linear plasma response to resonant magnetic perturbations during ELM mitigation in HL-3

N. Zhang, Y. Q. Liu, G. Z. Hao, J. M. Yu, T. F. Sun, G. Q. Dong, Yi Liu, L. Wang, J. Huang, M. Y. He, et al.

Active mitigation of edge-localized modes (ELMs)with the n=1 (n is the toroidal mode number) resonant magnetic perturbation (RMP) has recently been achieved for the first time on the HL-3 tokamak. The linear and quasi-linear plasma responses to RMP fields are numerically investigated by utilizing the MARS-F [Liu et al., Phys. Plasmas 7, 3681 (2000)] and MARS-Q [Liu et al., Phys. Plasmas 20, 042503 (2013)] codes. The linear results show that RMP induces a strong edge-peeling response which facilitates the ELM mitigation. A −50° phase shift for the n=1 coil current between the upper and lower rows of the RMP coils presents the optimal coil phase. MARS-Q quasi-linear results show that: (i) without involving perturbation mode near the plasma edge, the applied RMP has minor side effects on both the toroidal momentum and radial particle transport in this HL-3 case; (ii) allowing weak perturbation mode together with RMP produces finite flow damping and density pump-out level comparable to experiments; and (iii) the modeled flow damping and density pump-out is not very sensitive to the assumed resistivity model (Spitzer vs uniform resistivity). We found that both the neoclassical toroidal viscosity and resonant electromagnetic torques play important roles in the plasma toroidal momentum transport in HL-3.

Modeling the effects of carbon impurity transport and E×B drift on detachment cliff in the HL-3 open X-divertor by SOLPS-ITER

Rui Wang, Hailong Du, Na Wu, Chaofeng Sang, Lei Xue, Jiaxian Li, Xiao Song, Guangzhou Hao, Dongmei FAN, Guoliang Xiao, et al.

This work employs the SOLPS-ITER code package to investigate the effects of carbon impurity transport and E×B drift on the detachment cliff in the HL-3 open X-divertor configuration (OXD), and elucidates the physical mechanism behind the experimental observation that divertor detachment in the HL-3 OXD requires a higher upstream separatrix density ( n e,sep ) to occur than in the standard single null divertor configuration (SN). In this work, we find that: (1) The E×B drift more strongly suppresses divertor detachment in the HL-3 OXD under low n e,sep than in the SN; (2) A detachment cliff is observed in the HL-3 OXD with/without E×B drift. The open structure and large poloidal magnetic flux expansion of the HL‑3 OXD lead to a detachment cliff at the outer strike point under low n e,sep conditions without drifts, resulting in an inhomogeneous divertor state (strike point detached, far SOL high-recycling). With drifts included, the inhomogeneous state drives poloidal and radial E×B flows that together transport carbon impurities away from the outer strike point, making the HL‑3 OXD harder to detach—consistent with experiments. Besides, we find that the fundamental cause of the detachment cliff in the HL‑3 OXD with/without drifts: as n e,sep rises above a certain threshold, the abrupt reversal of carbon impurity transport direction and the resulting positive‑feedback loop in the divertor region. The difference lies in what drives the carbon impurity transport reversal: without drifts, it is driven by competition between thermal and frictional forces on carbon ions; with drifts, by changes in the poloidal E×B drift flow near the separatrix. These findings indicate that the open divertor structure of the HL‑3 OXD is the primary reason for its difficulty in achieving detachment. Enhancing its closure can significantly improve its detachment capability, providing a reference for the application of the X‑divertor in future devices (ITER/CFETR).

Direction-resolved synthetic DD neutron spectra based on arbitrary fast-ion distributions for the HL-3 tokamak

Xiangfeng Wu, Zhanhong Lin, Guangzhi Ren, Jie Zhang, Baolong Hao, Guoliang Yuan, Guangzhou Hao, Guanming Yang, Lei Feng, Zhengji Li, et al.

A spatially resolved and direction-resolved framework for computing fusion reaction rates and synthetic neutron energy spectra is implemented in the Particle Orbit Tracing Code (PTC). The model accepts arbitrary ion velocity distributions and explicitly separates thermal, beam--thermal, and beam--beam contributions, enabling calculations of local fusion sources and line-integrated neutron spectra for specified viewing geometries. The approach is applied to HL-3 tokamak plasmas with neutral beam injection (NBI). For core ion temperatures $T_{i0} \le 20\,\mathrm{keV}$, the volume-integrated DD neutron production rate is dominated by beam--thermal reactions, accounting for approximately 70--80\% of the total. Increasing the beam injection energy enhances the beam--thermal contribution to the neutron production rate and leads to higher-energy and broader neutron spectra. The neutron spectral shape depends strongly on the viewing geometry: a mid-plane radial line-of-sight (LOS) exhibits a clear double-peaked beam--thermal spectrum, whereas a mid-plane chord LOS yields a smoother, single-peaked spectrum. The volume-integrated DD neutron production rate predicted by PTC is slightly higher than the TRANSP result, with a relative difference of \(\sim 7\%\) when normalised to the PTC value. Compared with unfolded HL-3 DD neutron spectra from shots~\#6002 and \#12260, the LOS-integrated synthetic spectra capture the main DD-peak structure and the spectral shape on the high-energy side at the source-spectrum level. The remaining low-energy-side differences are mainly associated with diagnostic-chain effects and residual source-modelling limitations.

Coupled kinetic-fluid simulations on sputtering and transport of intrinsic carbon impurity in HL-3

Z. Liang, Y.L. Liu, Y. Feng, S.Y. Dai, M. Hoelzl, A. Cathey, D. Hu, S.Q. Korving, Y.C. Liang, M. Szucs, et al.

Study on the pumping performance of the condensation pump for neutral beam injection on HL-3

Xianfu Yang, Huiling Wei, Peixuan Yu, Hongxia Zhou

Development and operational progress of ion source for the second neutral beam injector on HL-3 tokamak

Huaiyu Luo, Huiling Wei, Peixuan Yu, Bowen Zhou, Yinxiang Wan, Peiqin Tang, Chunzhe Zhang, Haoyuan Zhu, Feng Zhang

HL-3 research towards high-performance plasma and power exhaust solution

Wulyu Zhong, Xiao Quan Ji, Wei Chen, Xingyu Bai, Alain Becoulet, Jerome B Bucalossi, Zhe Chen, Shaoyong Chen, Zhipeng Chen, Yihang Chen, et al.

The HL-3 tokamak program addresses critical challenges in developing integrated high-performance scenarios compatible with power exhaust demands for ITER and future reactors. Through systematic facility enhancements including auxiliary heating upgrades to 19.5 MW and AI-enabled control systems achieving 95.5% disruption avoidance, HL-3 finished a new round of exploration and validation for high-performance operation and power exhaust solution. Experiments at mega-ampere plasma currents demonstrated a reactor-relevant hot-ion regime with core ion temperatures exceeding 10 keV and a fusion triple product of . High normalized beta scenarios featuring internal and double transport barriers were successfully established. To address the compatibility of the plasma boundary, various small- or no-edge localized mode (ELM) regimes were achieved, including the enhanced D-alpha (EDA) H-mode, quiescent H-mode (QH-mode), and the quasi continuous exhaust (QCE) regime. Investigations into power exhaust integration highlighted the efficacy of advanced divertor configurations, specifically snowflake and tripod geometries, in significantly reducing peak heat fluxes. Active ELM control was demonstrated via resonant magnetic perturbations (RMPs), lower hybrid waves (LHW), and impurity supersonic molecular beam injection (SMBI), complemented by real-time feedback control of divertor detachment. These developments establish the fundamental physics basis and technical foundations necessary for demonstrating and extrapolating high-performance operations to ITER and next-step devices.

Laser-induced ablation spectroscopy on a linear plasma device: Magnetic field effects on signal and application to fuel retention diagnosis in HL-3 wall samples

Yang Liu, Yiqin Wang, Tao Huang, Feng Li, Qiuyue Nie, Qingmei Xiao

In fusion devices, boundary plasma particle and heat fluxes incident on plasma-facing materials (PFMs) lead to plasma-wall interactions (PWI). Real-time, in-situ quantitative diagnosis of wall material composition and fuel retention is critical for the safe and efficient operation of fusion reactors. Linear plasma devices offer a cost-effective platform for PWI studies, featuring extensive diagnostic access, fusion-relevant plasma conditions, and the ability to controllably vary plasma and surface parameters. In this work, a laser-induced ablation spectroscopy (LIAS) diagnostic system was developed on the "The Plasma-Surface Interaction Research Platform at Harbin Institute of Technology" (HIT-PSI) linear plasma device and applied to analyze graphite samples as a representative PFM. The influence of magnetic field conditions on LIAS signals was systematically investigated. Furthermore, the system has been successfully applied to diagnose first-wall samples from the HL-3 fusion device, enabling the acquisition of retention signals for deuterium and hydrogen. These results demonstrate the significant potential of LIAS in diagnosing first-wall composition and assessing fuel retention during fusion device operation.

Thermal analysis of ICRH antenna in HL-3 tokamak

Chenggonghang Zhou, Lingfeng Lu, Baolong Hao, Xiangfeng Wu, Jiabin Wan, Mei Huang, Xingyu Bai, Zhi Li, Li Cao, Churui Zhang, et al.

Ion cyclotron resonance heating (ICRH) is a vital auxiliary heating method for tokamak devices. As the core component of the ICRH system, the antenna directly faces the plasma and is subjected to complex thermal loads from multiple sources. In this study, thermal loads on the ICRH antenna under two representative heating schemes (6 MW ICRH + 5 MW NBI, 5 s pulse duration in Deuterium-Tritium and Hydrogen-Deuterium-Helium3 plasmas) are systematically analyzed. The heat flux on the Faraday screen surface caused by plasma radiation, fast ion loss, and the antenna’s inherent RF loss are evaluated. Simulation adopts the plasma radiation as the primary heat source on the faraday screen, contributing up to 60% of the total thermal load, while fast ion loss and RF loss account for 25% and 15%, respectively. Transient thermal analysis reveals that antenna components reach peak temperatures of approximately 300 °C during the first heating cycle. Following passive cooling, the second heating cycle elevates the temperature to about 450 °C, with each heating cycle producing a cumulative temperature rise of approximately 150 °C. These results demonstrate that real-time temperature monitoring and an active cooling system are critical for safe operation.

First reduced model for integrated computations of helicon wave heating and current drive in magnetic fusion plasmas

Zi-Chen Kan, Lei Chang, Zhen-Yu Wang, Hua-Sheng Xie, Ping-Wei Zheng, Lai Wei, Qi-Bin Luan, Xue-Mei Zhai, Zhao-Qing Hu, Zheng-Xiong Wang, et al.

Fast predictive modelling of radio-frequency heating and current drive is important for integrated tokamak scenario design, yet kinetic calculations of helicon-wave absorption remain too computationally expensive for large-scale parameter scans. We present a reduced model for helicon-wave heating and current drive that retains the dominant parallel electron Landau-damping channel. The wave response is evaluated on the cold-plasma dispersion root, and a single-Landau-pole correction is introduced to obtain compact expressions for the local damping rate and current-drive efficiency. The model is benchmarked against the Chiu-Chan heating model using approximately 1.6 million samples covering representative conditions of EAST, HL-3, DIII-D and KSTAR. The reduction error is found to be governed primarily by the electron Landau parameter and electron beta. Within an identified sub-lower-hybrid-frequency validity window, results from different devices collapse onto a common error curve, which enables an empirical correction that is further tested using ITER-like and BEST-like extrapolation cases. Near and above the lower-hybrid frequency, the agreement deteriorates rapidly owing to changes in the cold-dispersion root structure and the breakdown of the single-branch WKB description. When coupled to a reduced current-drive source, the corrected heating model gives a median deviation of 10.8 percent from the Landau-channel Ehst-Karney reference and reproduces published CFETR current-density profiles. The resulting model provides a computationally efficient reduced closure for helicon-wave heating and current-drive calculations, together with physically interpretable limits on its range of validity.

ICRF system for the HL-3 tokamak

Ling-Feng Lu, Jun Liang, Ya-Li Chen, Cheng-Gong-Hang Zhou, Chu-Rui Zhang, Zhi Li, Yue-Xin Ma, Li Cao, Jie-Qiong Wang, Xing-Yu Bai, et al.

The HL-3 tokamak is a new medium-sized copper conductor tokamak at the Southwestern Institute of Physics (SWIPs). In order to provide central ion heating as well as generate energetic particles whose energy levels are equivalent to the fusion born alpha particles, a 6 MW ion cyclotron range of frequencies heating system is under construction at SWIP and will be available at HL-3 in September 2026. Radio frequencies (RFs) in the range of 25–50 MHz with pulses up to 5 s is considered. For a deuterium plasma, the hydrogen minority and the 2nd harmonic of deuterium are the main ion heating schemes with f = 33 MHz at B 0 = 2.2 T. For a deuterium–tritium plasma, the fundamental He-3 and the 2nd harmonic of tritium with f = 25 MHz at B 0 = 2.5 T is considered. The RF generator consists of 4×1.5 MW transmitters. The transmission line incorporates 3 dB hybrid couplers so as to divert the reflected power away from the transmitters. The matching unit is provided by a 1/4 λ stub tuner and a 1/2 λ phase shifter, with an additional pre-matching stub for each of the four transmission lines. Two 2-strap antennas with parallel wavenumber k // ∼ 6.5 m −1 are designed since the impurity issue is supposed to be moderate under a carbon wall.

Forward Monte Carlo modeling of HL-3 D-D neutron spectra with fitted differential cross sections

Y.F. Xie, X.J. Chen, J.J. Qian, Y.W. Su, X.L. Chen, J. Zhang