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Heating loads on the divertor has been the major issue for the tokamak operation. The innovative negative triangularity plasmas are considered to be of the favorable property for the power handling. Investigating the hot-plasma vertical displacement events of the negative triangularity plasma has an important significance for the future fusion reactors. In this paper, the hot-plasma vertical displacement event of the negative triangularity plasmas are investigated by the DINA code, based on the new medium-sized copper conducting tokamak HL-2M. Simulation results show that: (i) The initial vertical instability, of the negative triangularity configuration, develops much more faster than that of the positive triangularity configurations; (ii) During the hot VDE, larger absolute value of the negative triangularity causes speeder development of the initial vertical instability and severer mechanical loads; (iii) During the hot VDE of the negative triangularity plasma, the most non-uniform distribution of the electromagnetic force almost occurs in the outer part of the vacuum vessel. Therefore, an arc design of this part is recommended; (iv) The initial vertical instability, as well as the electromagnetic loads, can be mitigated, by further optimizing the magnetic structure around the divertor region.

An upgrade to the magnet power digital feedback system (DFS) for the real time control of the PF coil power supplies is currently being developed and commissioned in HL-2A. The system aims to perform higher control resolution, more stable control cycle, real time monitoring and protection. The DFS has been shown to have better performance than the existing system in the control of the PF coil power supplies. It can control effectively all the PF coil power supplies, including the ohmic heating power supply (OHPS), vertical field power supply (VFPS), radial field power supply (RFPS) and the divertor coil power supply (DCPS). Meanwhile, the high-speed data transmission with reflective memory is used to transfer the control parameters of the PF coil power supplies. Moreover, the DFS also provides the real time monitor and protection based on the experimental physics and industrial control system (EPICS).

In order to study the advanced divertor physics and support the high performance plasma operation, a new control algorithm has been developed for the advanced divertor configuration (snowflake and tripod) on HL-2M. The response of the PF currents due to the change of the X-points positions has been established in this algorithm. The two X-points of the snowflake or tripod configuration can be independently controlled both at the radial and the vertical direction by this control algorithm. With this control algorithm, the preliminary simulation results have been achieved. Furthermore, with this algorithm the exact snowflake configuration can be obtained from the tripod configuration by control the X-points positions.

The plasma breakdown in HL-2A is found to match the Paschen law similar to the other tokamaks. It is also found on HL-2A that the breakdown time is about 10 ms, which is a little short for its high toroidal electric field (E ∼ 1.0–2.0 Vm−1), and the breakdown happens when the criterion of E × L (connection length) is about 33 V relative to a theoretical prediction of 67 V. Flux consumption optimization is achieved by scanning the loop voltage with identical target plasma current and heating with LHCD in the ramp-up phase. Based on Ohmic Heating during ramp-up phase, the limiter configuration consumes more flux than the divertor configuration. Adding LHCD during ramp-up phase reduces overall flux consumption.

SOLPS modeling including full cross-filed drifts is used, for the first time, to reproduce the step-like detachment onset observed in the DIII-D tokamak previously simulated with UEDGE (Jaervinen et al 2018 Phys. Rev. Lett. 121 075001), and to further investigate the underlying physics processes. The SOLPS analysis reveals a strong interplay between the E × B drift-driven flows and carbon radiation in DIII-D. The modeling shows that the high confinement H-mode facilitates the formation of such a detachment cliff due to the narrow SOL width and hence steep radial gradients near the separatrix, which enhances the poloidal E × B drift. Furthermore, the SOLPS modeling indicates that carbon radiation also plays a key role in the formation of the detachment cliff in DIII-D without additional impurity seeding. These findings may have significant implications for the control and maintenance of divertor detachment for the steady-state operation of high power fusion devices.

International audience; The SOLEDGE-EIRENE edge plasma code provides solutions for particle and energy transport in the plasma edge within complex and realistic 2D geometries (Bufferand et al 2015 Nucl. Fusion 55 053025). In this work, divertor detachment is simulated on HL-2M alternative magnetic configurations in pure deuterium plasma. Starting from a typical HL-2M low single-null configuration, the snowflake plus (SF+) and snowflake minus (SF-) configurations have then been investigated. Detachment of the outer target is studied in these configurations during plasma density ramps controlled by a fueling source, with constant input power and constant radial transport coefficients. Some typical characteristics of detachment, like threshold, depth and upstream window of detachment are investigated. In the three geometries, detachment onset and evolution with upstream plasma density is characterized by the gradual displacement of a radiation front from the outer target to the main X-point, as observed in experiments. It is found that, whatever the detachment in terms of particle, momentum or power dissipation, the detachment threshold is dominated primarily by the geometrical structure of the divertor plate and does not exhibit dependence on the magnetic configuration of the diverted plasma volume. In particular, the parallel connection length in the divertor is not found to affect the detachment threshold, in contrast with simple expectations from the two-point model, but in agreement with experimental findings.

HL-2M is a tokamak device that is under construction and will be put into operation in the near future. Based on the magnetic coil design of HL-2M, standard divertor, snowflake divertor and tripod divertor configurations have been designed. The potential properties of snowflake divertor configurations have been analyzed, such as the low poloidal field ( B p ) area around the X-point, the connection length, target plate and magnetic field shear. The linear peeling-ballooning (P-B) mode is studied by BOUT++ code for snowflake divertor configurations. According to the divertor configuration properties of HL-2M, asymmetric target plates have been concept designed to be compatible with the intended single null (SN) divertor configurations as well as double null (DN) divertor configurations. The SOLPS5.0 code is used to predict the details of the divertor plasma under the conditions of the divertor configurations noted above without impurities. This result shows that the peak heat load on an outer target plate of the advanced divertor is about 40% of that of the standard divertor. But more power will be transported to the inner target plate of advanced divertor, and this will cause a higher peak heat load on the inner target plate. The advanced divertor will also have to work under low plasma recycling conditions with high particle temperature and low density in an open divertor target geometry. When the SN configuration changes to a DN tripod divertor configuration, most of the power exhaust is handled by the outer divertor target plates and the peak heat load on these is about 4.1 MW/m2 (with a power exhaust of 20 MW). This range of optimized divertor configurations and target geometry will enable the study of advanced divertor physics and high performance plasmas in HL-2M tokamak.

The physical reason and picture of particle flux rollover (PFR) are investigated by using SOLPS with the effects of argon impurity and pumping in this paper. This modeling work uncovers the root/direct reason for PFR is the radiation power loss from impurity and recycling particles (D and D2). The recombination particle loss is the indirect reason for PFR. Besides, the radiation power loss also could play a key role in inducing momentum loss. This work could unify the four factors (recombination particle loss, impurity radiation power loss, power loss, momentum loss) from P. C. Stangeby and Krasheninnikov.

For the first time supersonic molecular beam injection (SMBI) and cluster jet injection (CJI) were applied to mitigate edge-localized modes (ELMs) in HL-2A successfully. The ELM frequency increased by a factor of 2–3 and the heat flux on the divertor target plates decreased by 50% on average after SMBI or CJI. Energetic particle induced modes were observed in different frequency ranges with high-power electron cyclotron resonance heating (ECRH). The high frequency (200–350 kHz) of the modes with a relatively small amplitude was close to the gap frequency of the toroidicity-induced Alfven eigenmode. The coexistent multi-mode magnetic structures in the high-temperature and low-collision plasma could affect the plasma transport dramatically. Long-lived saturated ideal magnetohydrodynamic instabilities during strong neutral beam injection heating could be suppressed by high-power ECRH. The absolute rate of nonlinear energy transfer between turbulence and zonal flows was measured and the secondary mode competition between low-frequency (LF) zonal flows (ZFs) and geodesic acoustic modes (GAMs) was identified, which demonstrated that ZFs played an important role in the L–H transition. The spontaneously generated E × B shear flow was identified to be responsible for the generation of a large-scale coherent structure (LSCS), which provided unambiguous experimental evidence for the LSCS generation mechanism. New meso-scale electric potential fluctuations (MSEFs) at frequency f ∼ 10.5 kHz with two components of n = 0 and m/n = 6/2 were also identified in the edge plasmas for the first time. The MSEFs coexisted and interacted with magnetic islands of m/n = 6/2, turbulence and LF ZFs.

HL-2M is a new medium-sized copper-conductor tokamak currently under final construction. Taking into account potential uncertainties in the engineering construction and to resolve design anomalies associated with new machine startup, high loop voltages (>10 V) without ECRH-assisted startup will be one of the important scenarios for the first plasmas in HL-2M. A basic field null configuration is obtained, that satisfies the initial discharge breakdown and the discharge evolution specifications, utilizing the CS, PF6 and PF8 coils of HL-2M. High loop voltage induces large stray magnetic field due to wall eddy currents in the vacuum vessel. Two schemes for compensating the stray field are investigated, using the PF8 coils. One scheme assumes the compensating current IPF8 being proportional to the wall eddy current, and the other scheme assumes fixed IPF8. With two schemes, self-consistent discharge waveforms are designed for the limiter configuration, by solving circuit equations for the plasma-conductor-power system. With the compensating current IPF8 fixed at a value between −150 A and −200 A, a good field null configuration is obtained. Evolution of the equilibrium configuration is further evaluated by the EFIT code, confirming self-consistency of the discharge waveform.

Author(s): Xu, M; Duan, XR; Liu, Y; Song, XM; Liu, DQ; Wang, YQ; Lu, B; Yang, QW; Zheng, GY; Ding, XT; Dong, JQ; Hao, GZ; Zhong, WL; Shi, ZB; Yan, LW; Zou, XL; Liu, YQ; Chen, W; Xiao, GL; Zhang, YP; Jiang, M; Hou, YM; Cheng, J; Liang, AS; Bai, XY; Cao, JY; Cao, Z; Chen, HT; Cui, CH; Cui, ZY; Delpech, L; Diamond, PH; Ekedahl, A; Feng, BB; Giruzzi, G; Gong, SB; He, HM; Hoang, T; Huang, M; Huang, Y; Ida, K; Inagaki, S; Isobe, M; Itoh, K; Itoh, SI; Ji, XQ; Ke, R; Kobayashi, M; Lei, GJ; Li, B; Li, GS; Li, HJ; Li, JQ; Li, Q; Li, Q; Li, XD; Li, YG; Li, Y; Mao, WC; Mazon, D; McKee, GR; Morita, S; Nie, L; Peng, JF; Peysson, Y; Rao, J; Ren, Y; Tynan, GR; Wang, HX; Wang, QM; Wei, HL; Xia, F; Xiao, XP; Xuan, WM; Yan, Z; Yao, K; Yu, LM; Yu, Y; Yu, DL; Yuan, BD; Yuan, BS; Zhang, JH; Zhou, Y; Liu, Y | Abstract: The mission of HL-2A is to explore the key physical topics relevant to ITER and to the advanced tokamak operation (e.g. the operation of future HL-2M), such as the access of H-mode, energetic particle physics, edge-localized mode (ELM) mitigation/suppression and disruption mitigation. Since the 2016 Fusion Energy Conference, the HL-2A team has focused on the investigations on the following areas: (i) pedestal dynamics and L-H transition, (ii) techniques of ELM control, (iii) turbulence and transport, (iv) energetic particle physics. The HL-2A results demonstrated that the increase of mean shear flow plays a key role in triggering L-I and I-H transitions, while the change of flow is mainly induced by the ion pressure gradient. Both mitigation and suppression of ELMs were realized by laser blow-off seeded impurity (Al, Fe, W). The 30% Ne mixture supersonic molecular beam injection seeding also robustly induced ELM mitigation. The ELMs were mitigated by low-hybrid current drive. The stabilization of m/n = 1/1 ion fishbone activities by electron cyclotron resonance heating was found on the HL-2A. A new m/n = 2/1 ion fishbone activity was observed recently, and the modelling indicated that passing fast ions dominantly contribute to the driving of 2/1 fishbone. The non-linear coupling between the toroidal Alfven eigenmode and tearing mode (TM) led to the generation of a high frequency mode with the toroidal mode number n = 0. The turbulence was modulated by TM when the island width exceeds a threshold and the modulation is localized merely in the inner area of the islands. Meanwhile, turbulence radial spread took place across the island region.

Effects of toroidal plasma flow, magnetic drift kinetic damping as well as feedback control, on the resistive wall mode instability in HL-2M tokamak are numerically investigated, using the linear stability codes MARS-F/K (Liu et al 2000 Phys. Plasmas 7 3681, Liu et al 2008 Phys. Plasmas 15 112503). It is found that the precession drift resonance damping due to trapped thermal particles ensures a robust passive stabilization of the n = 1 (n is the toroidal mode number) RWM in the 2 MA double-null advanced plasma scenario designed for HL-2M, provided that the toroidal flow speed is not too fast: . With two rows of magnetic control coils designed for HL-2M, the optimal poloidal location for the RWM stabilization is found to be . Toroidal modeling also shows that the plasma flow damping, drift kinetic damping and magnetic feedback can be arranged to synergistically stabilize the RWM in HL-2M, by tuning the feedback gain phase and/or including derivative actions in the control loop. The numerical results obtained by MARS-F/K are qualitatively well re-produced by an analytic single-pole model.

The China Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). CFETR will be operated in two phases. Steady-state operation and self-sufficiency will be the two key issues for Phase I with a modest fusion power of up to 200 MW. Phase II aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7 T, high frequency electron cyclotron resonance heating and lower hybrid current drive together with off-axis negative-ion neutral beam injection will be developed for achieving steady-state advanced operation. The recent detailed design, research and development (R&D) activities including integrated modeling of operation scenarios, high field magnet, material, tritium plant, remote handling and future plans are introduced in this paper.

Study of the hot-plasma vertical displacement event (VDE) in advanced divertor configurations is of significant importance for ITER and for future fusion reactors. The newly designed, medium-sized copper-conductor machine HL-2M has the capability of generating the second X-point for various advanced divertor configurations. In this paper, effects of the second X-point on the hot VDE in HL-2M are numerically investigated by utilizing the non-linear time-dependent DINA code. The simulation results show that the existence of the second X-point at certain special locations appears to have a better stability in the vertical direction, compared to the standard configuration with the same main plasma parameters. Meanwhile, the peak halo current during the current quench tends to increase as the second X-point changes in the horizontal direction. The same quantity decreases as the second X-point changes in the vertical direction away from the dominant X-point. From the view point of minimizing the halo current, the tripod is better than the standard configuration, followed by the snowflake-plus and the exact snowflake (SF) configuration. The SF-minus is the worst scenario. On the other hand, the tripod configuration, as well as the SF minus configurations, results in relatively higher peak electromagnetic force acting on the vacuum vessel, when compared to other aforementioned configurations.

The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with I p = 0.5 MA at the first stage of HL-2M and the high parameters with I p = 2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with d x = 30 cm present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, d x , to achieve a better performance in terms of the parameters of discharges.

During the tokamak operation, variation of the stored energy can cause internal perturbations of the plasma. These perturbations may develop into large-scale vertical movement of the whole column for the vertically elongated tokamak, eventually generating the hot vertical displacement event (VDE). It will cause considerable damage to the machine. In this work, the hot VDE process due to stored energy perturbations is investigated by a mature non-linear time-evolution code DINA. The influence on the vertical instability, the displacement direction and the electromagnetic loads on in-vessel components during the hot VDE are analyzed. It is shown that a larger perturbation leads to faster development of the vertical instability. Meanwhile the variation of the Shafranov shift, due to the energy change, is related to the VDE direction. The vertical electromagnetic force on the vacuum vessel and the halo current flowing in the divertor baffle become larger in the case of VDE moving towards the X point.

Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.