Recent Publications

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

Southwestern Institute of Physics, Xihua University

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.

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.

Southwestern Institute of Physics, IRFM, Dalian University of Technology

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.

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

University of South China, Southwestern Institute of Physics

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.

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.

Southwestern Institute of Physics, General Atomics, Sichuan Technology and Business University, Tsinghua University

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.

Density limit experiments and core-localized kinetic ballooning modes in HL-2A ohmic heating plasmas

Liwen Hu, Wei Chen, Peiwan Shi, Ting Long, Jianqiang Xu, R R Ma, Yonggao Li, Liming Yu, Xin Yu, Min Jiang, et al.

Southwestern Institute of Physics

The density limit is a mysterious barrier to magnetic confinement nuclear fusion, and is still an unresolved issue. In this paper, we will present the experimental results of the density limit and core-localized kinetic MHD instabilities on HL-2A. Firstly, the high density shots with ne/ne G > 1 have been achieved by the conventional gas-puff fuelling method in Ohmic heating plasmas, and the corresponding duration time is close to t ∼ 500 ms (∼ 30τ E ), where τ E is the global energy confinement time. Secondly, it is found for the first time that there are kinetic MHD instabilities in the core plasmas while ne/ne G ∼ 1. The analysis suggests that the core-localized MHD activities belong to kinetic ballooning modes (KBM), and firstly it is found on experiment that they trigger the minor or major disruption of bulk plasmas while the density profile is peaked. These new findings are of great importance to figure out and understand the origin of the density limit.

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

Southwestern Institute of Physics, Sun Yat-Sen University

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.

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.

Southwestern Institute of Physics, Dalian University of Technology

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.

Southwestern Institute of Physics, Dalian University of Technology

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.

Simulations of SMBI on High-Field and Low-Field Sides of the HL-2A Tokamak

GuanFeng Wu, Jingchun Li, Xueqiao Xu, kai xuan fan, Guoliang Xiao, Yiren Zhu

Shenzhen University, Southwestern Institute of Physics, Lawrence Livermore National Laboratory, Shanghai Jiao Tong University

Enhancing plasma injection efficiency and penetration depth is crucial for achieving high-performance steady-state fusion confinement in ITER, the next generation magnetic confinement fusion device. Based on the trans-neut module within the BOUT++ boundary plasma turbulence program framework, this study conducts two-dimensional simulations and comparative analyses of supersonic molecular beam injection (SMBI) on the high-field side (HFS) and the low-field side (LFS), using the actual divertor geometry of the HL-2A tokamak. The physical model encompasses plasma density, heat, and momentum transport equations, as well as neutral particle density and momentum transport equations. The results show that HFS SMBI produces deeper neutral penetration in normalized poloidal-flux space than LFS SMBI. The ion density inside the LCFS in the HFS case reaches about 1.5-2 times that in the LFS case, accompanied by a stronger reduction of the edge plasma temperature. The deeper HFS penetration is associated with the smaller local field-line slope on the HFS path, which causes a finite-width SMBI source to be mapped differently along the magnetic field compared with the LFS case. These results indicate that HFS SMBI can enhance both the penetration depth and the fueling efficiency in the present HL-2A geometry.

Structural integrity analysis of glovebox as the secondary confinement barrier in fusion reactor projects

Jie Liu, Zhe Chen, Fengchao Zhao, Haijie Dong, Xinghua Wu

China Fusion Energy Co., Ltd, Southwestern Institute of Physics

Effect of surface roughness on corrosion behavior of CLF-1 steel in magnetic field environments: increased surface roughness accelerates corrosion

Ye Li, Yi-Ming Lyu, Shufeng Zhang, Ning Liu, Jiejie Li, Yuxin Xiao, Guoping Yang, Shanliang Zheng

Hefei Comprehensive National Science Center (Anhui Energy Laboratory), Hefei Institutes of Physical Science, Chinese Academy of Sciences, University of Science and Technology of China, Southwestern Institute of Physics

Theoretical insights into the key properties of Ce1−xMxB6 (M = Ba, Sr, Ca) as promising plasma grid materials for N-NBI systems

Jun Hu, Zilin Cui, Yucai Li, Yuhong Xu, Wei Li, Xin Zhang, Guangjiu Lei, Sanqiu Liu, Xiaolong Li, Huaqing Zheng, et al.

Southwest Jiaotong University, Nanchang University, Sichuan University, Southwestern Institute of Physics

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.

Southwestern Institute of Physics, University of Electronic Science and Technology of China, Chongqing university

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.

Preliminary engineering analysis for CN HCCB TBM regarding ITER new baseline scenario

Xinghua Wu, Shen Qu, Ruyan Li, Hongxiang Zhang, Qixiang Cao, Fengchao Zhao, Long Zhang, Xiaoyu Wang

Southwestern Institute of Physics, China Fusion Energy Co. Ltd

Among the various breeding blanket concepts proposed for DEMO reactor design by different countries, China has ultimately determined to develop and test the Helium-Cooled Ceramic Breeder Test Blanket Module (HCCB TBM) in ITER, in alignment with its national strategy for the development of magnetic confinement fusion energy. During the preliminary design phase, the CN HCCB TBM team implemented several design updates to improve its engineering performance and manufacturing feasibility, however, in response to certain engineering and technical challenges, a new ITER baseline has been under development since early February 2023. This new baseline proposed a revised ITER operation strategy aimed at initiating the nuclear phase as early as possible while reducing selected operational parameters in the initial stages. In accordance with the ITER 2024 new baseline scenario, further design optimization and engineering analysis have been conducted for the CN HCCB TBM. Thermal-hydraulic analysis results demonstrate that, the maximum operating temperatures of structural and functional materials were significantly reduced under the 2024 baseline scenario, especially for the tritium breeder Li 4 SiO 4 pebble bed, whose temperature was decreased by 20.6%. By installing electric heaters within the tritium breeder zone, the peak temperature of Li 4 SiO 4 can be increased to 821 °C, and the average temperature raised o 604 °C, which essentially satisfied the temperature requirements for tritium release. Based on the calculated temperature distributions under three operating conditions, a system-level transient tritium transport analysis was further performed. The results reveal that installing electric heaters exclusively in the tritium breeder region is sufficient to meet tritium balance requirements, eliminating the need for additional electric heaters in the neutron multiplier region. Preliminary structural analysis was also carried out, and the results indicated that the linearized stresses remain well below the allowable limits of the structural material, thereby ensuring the integrity of the overall structure.

Development of a Reduced Multi-Fluid Equilibrium Model and Its Application to Proton-Boron Spherical Tokamaks

Hua-Sheng Xie, XingYu Li, Jiaqi Dong, Zhiwei Ma, Yunfeng Liang, Yuejiang Shi, Wenjun Liu, Yueng-Kay Martin Peng, Lai WEI, Zheng-Xiong Wang, et al.

ENN Science and Technology Development Co., Ltd, Dalian University of Technology, Beijing VeloAlpha Technology Co., Ltd, Southwestern Institute of Physics, Zhejiang University

Proton-Boron (p-$^{11}$B) fusion represents a promising pathway toward aneutronic clean energy but requires extremely high ion temperatures and robust magnetic confinement. Spherical Tokamaks/Torus (ST) driven by high-power neutral beam injection are a primary candidate for this regime. In such devices, the combination of strong toroidal rotation and the significant mass disparity between protons and boron ions leads to complex multi-fluid effects---specifically centrifugal species separation and electrostatic polarization---which standard single-fluid magnetohydrodynamic (MHD) models fail to capture. Conversely, comprehensive multi-fluid models that include poloidal flows often suffer from numerical stiffness and excessive complexity, hindering their use in routine engineering analysis. To address these challenges, we have developed a \textit{reduced multi-fluid equilibrium model} designed to balance physical fidelity with computational robustness. By retaining the dominant toroidal rotation and self-consistent electrostatic potential while neglecting secondary effects such as poloidal flow inertia and pressure anisotropy, the model is formulated as a generalized Grad-Shafranov equation coupled with species-specific Bernoulli relations and a quasi-neutrality constraint. The model is applied to analyze the equilibrium configurations of two representative p-$^{11}$B ST devices designed by the ENN Group: the experimental EHL-2 and the reactor-scale EHL-3B. Simulation results demonstrate that the equilibrium modification is governed by the ion Mach number ($M$). In the low-rotation regime ($M < 0.5$), multi-fluid effects are weak, and the solution converges toward the single-fluid limit. However, in the high-rotation regime ($M > 2$), strong centrifugal forces drive significant boron accumulation at the low-field side (LFS) and generate an internal electrostatic potential on the order of 10 kV. These results confirm the necessity of multi-fluid modeling for accurate p-$^{11}$B reactor design within the assumptions quantified in this work.

Impact of biasing-induced magnetic perturbations on divertor heat loads in HL-2A

Dongmei FAN, Tengfei Sun, Jinming Gao, Guoliang Xiao, Hailong Du, Liang Liu, Zhongbing Shi, Wulyu Zhong

Southwestern Institute of Physics

Divertor biasing on HL-2A is observed to control edge localized modes (ELMs) and modify divertor heat load profiles by generating resonant magnetic perturbations (B-RMPs) via currents driven in the scrape-off layer (SOL). In H-mode plasmas, this leads to clear strike-point splitting and significant ELM mitigation. To interpret these results, we perform the first 3D edge transport simulations of such experiments using the EMC3-Eirene code. Toroidal modeling confirms that the magnetic perturbation generated by the divertor biasing current is capable of substantially modifying the edge magnetic topology, resulting in strike point splitting on the divertor target. The plasma transport simulations qualitatively capture the key feature of strike-point splitting observed in the experiment, showing reasonable qualitative agreement with experimental measurements.These combined experimental and modeling results support the physical feasibility of divertor biasing for controlling divertor heat loads and mitigating ELMs in tokamak plasmas, and establish a critical numerical benchmark for this technique.

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

Sichuan University, Southwestern Institute of Physics

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.

Southwestern Institute of Physics, CEA, UKAEA, Huazhong University of Science and Technology, National Institute for Fusion Science

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.

Research on power supply system for pulsed loads of fusion devices

Qiang Ye, Yingqiao Wang, Haibin Wang, Yuemeng Yin, Weibin Li

Southwestern Institute of Physics

Preliminary results of soft X-ray measurements on Thailand Tokamak-1 using silicon-based spectrometer

Sawarin Buakham, Siriyaporn Sangaroon, Kunihiro Ogawa, Arlee Tamman, Nopporn Poolyarat, Pasit Wonghabut, Kewalee Nilgumhang, Tossawat Autila, Thiti Rungseesumran, Poramate Chunpang, et al.

Thailand Institute of Nuclear Technology (Public Organization), Mahasarakham University, National Institute for Fusion Science, The Graduate University for Advanced Studies, SOKENDAI, Southwestern Institute of Physics