By analyzing large quantities of discharges in the unfavorable ion B ×∇B drift direction, the I-mode operation has been confirmed in EAST tokamak. During the L-mode to I-mode transition, the energy confinement has a prominent improvement by the formation of a high-temperature edge pedestal, while the particle confinement remains almost identical to that in the L-mode. Similar with the I-mode observation on other devices, the E r profiles obtained by the eight-channel Doppler backscattering system (DBS8)[1] show a deeper edge E r well in the I-mode than that in the L-mode. And a weak coherent mode (WCM) with the frequency range of 40-150 kHz is observed at the edge plasma with the radial extend of about 2-3 cm. WCM could be observed in both density fluctuation and radial electric field fluctuation, and the bicoherence analyses showed significant couplings between WCM and high frequency turbulence, implying that the E r fluctuation and the caused flow shear from WCM should play an important role during I-mode. In addition, a low-frequency oscillation with a frequency range of 5-10 kHz is always accompanied with WCM, where GAM intensity is decreased or disappeared. Many evidences show that the a low-frequency oscillation may be a arXiv:1902.04750v3 [physics.plasm-ph]

We report an observation of robust suppression of edge-localized modes (ELMs) in the Experimental Advanced Superconducting Tokamak (EAST), enabled by continuous boron (B) powder injection. Edge harmonic oscillations appear during B powder injection, providing sufficient particle transport to maintain constant density and avoid impurity accumulation in ELM-stable plasmas. Quasi-steady ELM suppression discharges are demonstrated with modest energy confinement improvement and over a wide range of conditions: heating power and technique variation, electron density range over a factor ∼3.5, deuterium or helium ion species, and with either direction of the toroidal magnetic field. ELM suppression is observed above a threshold edge B intensity and ceases within 0.5 s of termination of the B injection. In contrast to ELM suppression accompanied by recycling reduction during Li powder injection in NSTX and EAST (Maingi et al 2018 Nucl. Fusion 58 024003), reduced recycling due to hydrogenic species retention is unnecessary for the ELM suppression with B powder injection, paving the way for its consideration as an ELM control tool for future fusion devices.

Full suppression of type-I edge localized modes (ELMs) using n = 4 resonant magnetic perturbations (RMPs) as planned for ITER has been demonstrated for the first time (n is the toroidal mode number of the applied RMP). This is achieved in EAST plasmas with low input torque and tungsten divertor, and the target plasma for these experiments in EAST is chosen to be relevant to the ITER Q = 10 operational scenario, thus also addressing significant scenario issues for ITER. In these experiments the lowest neutral beam injection (NBI) input torque is around T NBI ∼ 0.44 Nm, which extrapolates to around 14 Nm in ITER (compared to a total torque input of 35 Nm when 33 MW of NBI are used for heating). The q 95 is around 3.6 and normalized plasma beta β N ∼ 1.5–1.8, similar to that in the ITER Q = 10 scenario. Suppression windows in both q 95 and plasma density are observed; in addition, lower plasma rotation is found to be favourabe to access ELM suppression. ELM suppression is maintained with line averaged density up to 60%n GW (Greenwald density limit) by feedforward gas fuelling after suppression is achieved. It is interesting to note that in addition to an upper density, a low density threshold for ELM suppression of 40%n GW is also observed. In these conditions energy confinement does not significantly drop (<10%) during ELM suppression when compared to the ELMy H-mode conditions, which is much better than previous results using low n (n = 1 and 2) RMPs in higher q 95 regimes. In addition, the core plasma tungsten concentration is clearly reduced during ELM suppression demonstrating an effective impurity exhaust. MHD response modelling using the MARS-F code shows that edge magnetic field stochasticity has a peak at q 95 ∼ 3.65 for the odd parity configuration, which is consistent to the observed suppression window around 3.6–3.75. These results expand the physical understanding of ELM suppression and demonstrate the effectiveness of n = 4 RMPs for reliable control ELMs in future ITER high Q plasma scenarios with minimum detrimental effects on plasma confinement.

A reproducible stationary improved confinement mode (I-mode) has been achieved recently in the Experimental Advanced Superconducting Tokamak (EAST), featuring good confinement without particle transport barrier. The microscopic mechanism of sustaining stationary I-mode, based on the coupling between turbulence transitions and the edge temperature oscillation, has been discovered for the first time. A radially localized edge temperature ring oscillation (ETRO) with azimuthally symmetric structure (n = 0, m = 0) has been identified and it is accompanied by alternating turbulence transitions between an electron diamagnetic drift turbulence (ET) and an ion diamagnetic drift turbulence (IT). The transition is controlled by local electron temperature gradient and strong non-linear couplings between weak coherent mode (WCM) and ET could be identified near the pedestal top, suggesting the unique status of the pedestal top region in sustaining the stationary I-mode confinement on EAST.

Demonstration of DEMO relevant fusion power (P fus) level and tritium self-sufficiency are two important goals of the China fusion engineering testing reactor (CFETR). In this work the integrated modeling including self-consistent core–pedestal coupling are used to design the hybrid scenario plasmas at flat-top phase for these goals. Such plasmas have been taken as the reference plasma for studying the compatibility of the hybrid scenario with CFETR engineering design in the past two years. The physics justification for the selection of plasma density, Z eff, safety factor profile, and in particular the choice of auxiliary heating and current drive is presented. According to a scan of plasma density and Z eff, the target of P fus ≈ 1 GW and finite ohmic flux consumption ∆Φohm (4 h) ⩽ 250 Vs can be met with Z eff = 1.9–2.2 and the density at the pedestal top set at 90% of the Greenwald limit. Turbulent transport analysis using the gyro-Landau-fluid model TGLF shows that the electromagnetic effects can enhance the energy confinement but reduce the particle confinement and thus P fus. A baseline hybrid scenario case matching the target in the concept design is built using a combination of neutral beams (NB) and electron cyclotron (EC) waves to flatten the safety factor profile in the deep core region (with the normalized plasma radius ρ ⩽ 0.4). Such profile can yield better particle and energy confinement than that with either higher magnetic shear in the deep core region or higher q value in outer core region (e.g., due to the addition of lower hybrid current drive). Switching a part of auxiliary heating from electron to ions, e.g., replacing a part of EC waves by waves in the ion cyclotron range of frequencies, reduces the particle confinement and thus P fus. Since high harmonic fast waves (HHFW) can drive current at the same location as ECCD with higher current drive efficiency than ECCD and yield more electron heating than NB, the case using HHFW to replace a part of EC waves and NB can yield higher P fus and lower ∆Φohm than the baseline case. A discussion is given on future simulations to explore the improvement in plasma performance and the broadening of the feasible design space.

Both a fully noninductive steady state operation scenario and a hybrid scenario with fusion power ∼ 1 GW and fusion gain &gt;10 are being considered to fulfill the mission of a Chinese fusion engineering testing reactor. Compared to the hybrid scenario, plasma current is generally lower in steady state operation, so that better confinement and stabilization of MHD instability introduced by higher normalized beta (possibly beyond the ideal MHD limit without a wall) are required to achieve the same fusion performance. Integrated modeling is used to find candidate scenarios to match both these requirements at the same time. By creating a localized strong reversed magnetic shear using radio frequency wave driven current, a strong off-axis internal transport barrier is formed, so that the target fusion power and fusion gain are achieved for Chinese fusion engineering testing reactor steady state operation. Further optimizing the location of the reversed magnetic shear by modifying radio frequency wave launch parameters can keep the normalized beta below the ideal MHD no-wall limit while the fusion power remains beyond 1 GW. Based on this finding, several combinations of heating and current drives are proposed with fusion gain close to 12.5.

The geodesic acoustic mode (GAM) has been observed for the first time in EAST H-mode operation using Doppler backscattering (DBS) systems. The poloidal and toroidal symmetries of the radial electric field (Er) fluctuation (m  =  0 and n  =  0) have also been demonstrated. Using the multichannel DBS system, the GAM frequencies at different radial locations have been investigated, with observations showing that H-mode GAM acts as a radial coherent eigenmode, rather than a continuum of frequencies in L-mode. The intermittency of GAM is observed in both L-mode and H-mode. The interplay between the background density fluctuations and the GAM intensity is revealed. The bicoherence analysis is also used to verify the presence of the nonlinear interaction between the GAM and background turbulence.

An important task of the China Fusion Engineering Test Reactor physics design is to develop operation scenarios with high fusion power (1 GW), high bootstrap current fraction for steady-state and a plasma edge compatible with heat and particle exhaust. To achieve these goals, triangularity (δ) effects on the fusion performance of two candidate scenarios, with or without reversed magnetic shear (RS), namely conventional H-mode and RS H-mode, are evaluated using core-edge coupled integrated modeling in this paper. For fixed pedestal density, it is shown that higher δ is favorable for higher fusion performance in the conventional H-mode scenario while the fusion performance decreases with increasing δ in the RS H-mode scenario. In conventional H-mode, the higher fusion performance at high δ mainly comes from a higher pedestal temperature as predicted by EPED in combination with stiff core kinetic profiles. In the RS H-mode scenario with a local reversed shear region, the profiles are non-stiff and a strong internal transport barrier (ITB) exists at low δ. This results in higher density and temperature inside the ITB for low δ, leading to higher fusion power. If the pedestal temperature is kept fixed, in both scenarios the significant increase in pedestal density, which extends into the core, dominates at high δ and leads to much higher fusion power. For conventional H-mode, destabilization from increasing δ is partially balanced by stabilization due to increasing ν*. Since the normalized heat sources are quite similar, it results in minimal changes in the temperature profiles except for the lowest density case. For RS H-mode, destabilization from increasing δ is approximately balanced by stabilization due to increasing ν* in foot region, but a strong temperature ITB is still evident for low δ. The ability to take advantage of the high pedestal density in conventional H-mode and reversed shear scenario depends on its compatibility with edge density requirements from efficient heat and particle exhaust. Transport analysis is presented to elucidate the roles of δ, collisionality and magnetic shear in altering the profiles and the ITB, which contribute to the different behavior in the two scenarios.