Toroidal Rotation Research Articles

Understanding core heavy impurity transport in a hybrid discharge on EAST

The behavior of heavy/high-Z impurity tungsten (W) in the core of hybrid (high normalized beta β N plasmas) scenario on EAST with international thermonuclear experimental reactor-like divertor is analyzed. W accumulation is often observed and seriously degrades the plasma performance (Gao et al 2017 Nucl. Fusion 57 056021). The dynamics of the W accumulation process of a hybrid discharge are examined considering the concurrent evolution of the background plasma parameters. It is found that the toroidal rotation and density peaking of the bulk plasma are usually large in the central region, which is particularly prone to the W accumulation. A time slice during the W accumulation phase is modeled, accounting for both neoclassical and turbulent transport components of W, through NEO with poloidal asymmetry effects induced by toroidal rotation, and TGLF, respectively. This modeling reproduces the experimental observations of W accumulation and identifies the neoclassical inward convection/pinch velocity of W due to the large density peaking of the bulk plasma and toroidal rotation in the central region as one of the main reasons for the W accumulation. In addition, the NEO + TGLF + STRAHL modeling can not only predict the core W density profile but also closely reconstruct the radiated information mainly produced by W in the experiment.

Understanding core tungsten (W) transport and control in an improved high-performance fully non-inductive discharge on EAST

The behavior of heavy/high-Z impurity tungsten (W) in an improved high-performance fully non-inductive discharge on EAST with ITER-like divertor (ILD) is analyzed. It is found that W could be well controlled. The causes of no W accumulation are clarified by analyzing the background plasma parameters and modeling the W transport. It turns out that the electron temperature (T e) and its gradient are usually high while the toroidal rotation and density peaking of the bulk plasma are small. In this condition, the modeled W turbulent diffusion coefficient is big enough to offset the total turbulent and neoclassical pinch, so that W density profile for zero particle flux will not be very peaked. Combining NEO and TGLF for the W transport coefficient and the impurity transport code STRAHL, not only the core W density profile is predicted but also the radiated information mainly produced by W in the experiment can be closely reconstructed. At last, the physics of controlling W accumulation by electron cyclotron resonance heating (ECRH) is illustrated considering the effects of changed T e by ECRH on ionization balance and transport of W. It shows that the change of ionization and recombination balance by changed T e is not enough to explain the experimental observation of W behavior, which should be attributed to the changed W transport. By comparing the W transport coefficients in two kinds of plasmas with different T e profiles, it is shown that high T e and its gradient play a key role to generate large turbulent diffusion through increasing the growth rate of linear instability so that W accumulation is prevented.

Simulation study of fast ion losses associated with the rotating n = 1 resonant magnetic perturbations in KSTAR

This paper describes a simulation framework for testing the fast ion loss mechanism associated with the experimentally observed beam ion losses when an externally applied toroidally rotating perturbed magnetic field is used to control edge localized modes in the KSTAR tokamak. The simulations reproduce the key qualitative features of neutral beam injection (NBI) ion detection by a fast ion loss detector. The NBI ion losses in the simulation mainly occur for passing particles due to orbit stochastization, which is caused by orbit resonance with the 3D field perturbations. The relative toroidal angle of NBI ion deposition to the orbit island determines the radial path of the particles between confinement or loss. The fast ion loss quantity depends on the toroidal angle rotation of the 3D field with respect to the beam deposition position. The averaged transport of toroidal canonical angular momentum reveals that linear and nonlinear resonance of the NBI ions within the 3D field are the main factors determining fast ion transport and loss.

High-βp scenario realized by the integration of internal and external transport barriers in the HL-2A tokamak

High-βp scenario addresses needs for an attractive tokamak fusion reactor design, where βp is the normalized plasma poloidal beta. High-βp experiments have been performed during recent years in the HL-2A tokamak. The high-performance region is realized by the integration of internal and external transport barriers which are dubbed as ITBs and ETBs, namely, double transport barriers (DTBs), with the high-power NBI heating. Generally, the ITB forms and becomes strong after the NBI injection on HL-2A. Subsequently, the edge ion temperature and toroidal rotation increase, as a result that the ETB or pedestal creates and L-H transition occurs. A high-βp (βp∼3) scenario is obtained with Ip∼110 kA and q95∼5, and another high-βp (βp∼2.7) scenario is also realized with Ip∼160 kA and q95∼4.2. The two high-βp scenarios are both characterized by a large-radius ITB and an ETB. However, the two scenarios are different. For the former there are strong large-radius ITB and weak ETB on radial ion temperature profiles, but for the later there are weak large-radius ITB and strong ETB. Although two high-βp scenarios on HL-2A have been accessed, however they are both transient. In addition, MHD instabilities in high-βp scenarios are also present for a high fusion gain and reactor relevant high fusion-density plasma.

Dynamics of ultralow-q plasmas in the RFX-mod device

The results are presented of an experimental activity performed in the RFX-mod device aimed at characterizing plasma dynamics in the so-called Ultralow-q (ULq) magnetic configuration, which corresponds to edge safety factor values below 1. The role of the edge safety factor in determining plasma dynamics is studied. In particular, a characterization of MHD activity is performed. The results of dedicated non-linear 3D visco-resistive MHD simulations are in good quantitative agreement with the experimental observations. In particular, the predicted tendency for ULq plasmas to be characterized by magnetic spectra dominated by a single mode (either a kink or a double resonant internal mode) is confirmed by experiment. Magnetic reconnection plays a relevant role in determining the dynamics of the magnetic topology. Both almost quiescent and largely fluctuating plasmas are observed with a strong sensitivity on the edge safety factor. The main MHD properties of the ULq are compared to those of RFP and tokamak discharges, also produced in the RFX-mod device. MHD modes exhibit toroidal rotation at a frequency depending on mode amplitude. Differently from what encountered in RFP plasmas at comparable current levels, no wall locking is detected.

Non-linear MHD modelling of edge localized modes suppression by resonant magnetic perturbations in ITER

Edge localized modes (ELMs) suppression by resonant magnetic perturbations (RMPs) was studied with the non-linear magneto-hydro-dynamic (MHD) code JOREK for the ITER H-mode scenarios at 15 MA, 12.5 MA, 10 MA/5.3 T. The main aim of this work was to demonstrate that ELMs can be suppressed by RMPs while the divertor 3D footprints of heat and particle fluxes remain within divertor material limits. The unstable peeling–ballooning modes responsible for ELMs without RMPs were modelled first for each scenario using numerically accessible parameters for ITER. Then the stabilization of ELMs by RMPs was modelled with the same parameters. RMP spectra, optimized by the linear MHD MARS-F code, with main toroidal harmonics N = 2, N = 3, N = 4 have been used as boundary conditions of the computational domain of JOREK, including realistic RMP coils, main plasma, scrape off layer (SOL) divertor and realistic first wall. The model includes all relevant plasma flows: toroidal rotation, two fluid diamagnetic effects and neoclassical poloidal friction. With RMPs, the main toroidal harmonic and the non-linearly coupled harmonics remain dominant at the plasma edge, producing saturated modes and a continuous MHD turbulent transport thereby avoiding ELM crashes in all scenarios considered here. The threshold for ELM suppression was found at a maximum RMP coils current of 45 kAt–60 kAt compared to the coils maximum capability of 90 kAt. In the high beta poloidal steady-state 10 MA/5.3 T scenario, a rotating QH-mode without ELMs was observed even without RMPs. In this scenario with RMPs N = 3, N = 4 at 20 kAt maximum current in RMP coils, similar QH-mode behaviour was observed however with dominant edge harmonic corresponding to the main toroidal number of RMPs. The present MHD modelling was limited in time by few tens of ms after RMPs were switched on until the magnetic energy of the modes saturates. As a consequence the thermal energy was still evolving on this time scale, far from the ITER confinement time scale and hence only the form of 3D footprints on the divertor targets can be indicated within this set-up. Also note, that the divertor physics was missing in this model, so realistic values of fluxes are out of reach in this modelling. However the stationary 3D divertor and particle fluxes could be simply extrapolated from these results to the stationary situation considering that a large power fraction should be radiated in the core and SOL and only about 50 MW power is going to the divertor, which is an arbitrary, but reasonable number used here. The 3D footprints with RMPs show the characteristic splitting with the main RMP toroidal symmetry. The maximum radial extension of the footprints typically was ∼20 cm in inner divertor and ∼40 cm in outer divertor with stationary heat fluxes decreasing further out from the initial strike point from ∼5 MW m−2 to ∼1 MW m−2 assuming a total power in the divertor and walls is 50 MW. The heat fluxes remain within the divertor target and baffle areas, however with rather small margin in the outer divertor which could be an issue for the first wall especially in transient regimes when part of the plasma thermal energy is released due to switching on the RMP coils. This fact should be considered when RMPs are applied with a more favorable application before or soon after the L–H transition, although optimization is required to avoid increasing the L–H power threshold with RMPs.

Numerical study of the effect of kinetic damping on resistive wall modes with plasma toroidal rotation in CFETR

Herein, the effect of kinetic damping on the resistive wall mode (RWM) instability in the China nuclear fusion engineering test reactor (CFETR) is numerically studied using the magnetohydrodynamic resistive spectrum-kinetic (MARS-K) code. A CFETR hybrid equilibrium with a plasma current of Ip = 13 MA is adopted. The sum of the electron diamagnetic, plasma rotation, and mode frequencies increases at q = 2 rational surface with increasing plasma rotation frequency. When the sum of the frequencies is less than 0, kinetic damping has a stabilizing effect on RWM. In contrast, when the sum of the frequencies is greater than 0, the kinetic damping has a destabilizing effect on RWM, causing a second unstable branch to appear. Additionally, the effects of plasma pressure and wall distance on the two unstable branches are investigated. The mode frequency is excited with increasing plasma pressure and wall distance. Consequently, the first branch is stabilized and the second branch is destabilized.

Role of NBI fuelling in contributing to density peaking between the ICRH and NBI identity plasmas on JET

Density peaking has been studied between an ICRH and NBI identity plasma in JET. The comparison shows that 8 MW of NBI heating/fueling increases the density peaking by a factor of two, being R/L n = 0.45 for the ICRH pulse and R/L n = 0.93 for the NBI one averaged radially over ρ tor = 0.4, 0.8. The dimensionless profiles of q, ρ *, υ *, β n and T i/T e ≈ 1 were matched within 5% difference except in the central part of the plasma (ρ tor < 0.3). The difference in the curvature pinch (same q-profile) and thermo-pinch (T i = T e) between the ICRH and NBI discharges is virtually zero. Both the gyro-kinetic simulations and integrated modelling strongly support the experimental result where the NBI fuelling is the main contributor to the density peaking for this identity pair. It is to be noted here that the integrated modeling does not reproduce the measured electron density profiles, but approximately reproduces the difference in the density profiles between the ICRH and NBI discharge. Based on these modelling results and the analyses, the differences between the two pulses in impurities, fast ions (FIs), toroidal rotation and radiation do not cause any such changes in the background transport that would invalidate the experimental result where the NBI fuelling is the main contributor to the density peaking. This result of R/L n increasing by a factor of 2 per 8 MW of NBI power is valid for the ion temperature gradient dominated low power H-mode plasmas. However, some of the physics processes influencing particle transport, like rotation, turbulence and FI content scale with power, and therefore, the simple scaling on the role of the NBI fuelling in JET is not necessarily the same under higher power conditions or in larger devices.

Low-frequency zonal flow eigen-structures in tokamak plasmas

The nonlocal eigenmode analysis of low-frequency zonal flows (ZFs) in toroidally rotating tokamak plasmas is performed in the framework of the reduced one-fluid ideal magnetohydrodynamic model. It is shown that for typical profiles of plasma parameters toroidal plasma rotation results in the global ZF formation on the periphery of plasma column. For some types of equilibria these ZFs are aperiodically unstable that leads to the excitation of the differential plasma rotation at the tokamak plasma edge.

Deconvolving the roles of E × B shear and pedestal structure in the energy confinement quality of super H-mode experiments

Analysis of ‘super H-mode’ experiments on DIII-D has put forward that high plasma toroidal rotation, not high pedestal, plays the essential role in achieving energy confinement quality H 98y2 ≫ 1 (Ding et al 2020 Nucl. Fusion 60 034001). Recently, super H-mode experiments with variable input torque have confirmed that high rotation shear discharges have very high levels of H 98y2 (>1.5), independent of the pedestal height, and that high pedestal discharges with low rotation shear have levels of H 98y2 only slightly above 1 (⩽1.2). Although some increase in stored energy with higher pedestal occurs, the energy confinement quality mainly depends on the toroidal rotation shear, which varies according to different levels of injected neutral beam torque per particle. Quasi-linear gyrofluid modeling achieves a good match of the experiment when including the E × B shear; without including plasma rotation, the modeling predicts a confinement quality consistent with the empirical observation of H 98y2 ∼ 1.2 at low rotation. Nonlinear gyrokinetic transport modeling shows that the effect of E × B turbulence stabilization is far larger than other mechanisms, such as the so-called hot-ion stabilization (T i/T e) effect. Consistent with these experimental and modeling results are previous simulations of the ITER baseline scenario using a super H-mode pedestal solution (Solomon et al 2016 Phys. Plasmas 23 056105), which showed the potential to exceed the Q = 10 target if the pedestal density could be increased above the Greenwald limit. A close look at these simulations reveals that the predicted energy confinement quality is below 1 even at the highest pedestal pressure. The improvement in Q at higher pedestal density is due to the improved fusion power generation at the higher core density associated with higher pedestal density, not to an improved energy confinement quality.

Nonlinear excitation of geodesic acoustic mode by drift waves in anisotropic tokamak plasmas with toroidal rotation

Nonlinear excitation of geodesic acoustic mode (GAM) by ions drift waves is investigated in anisotropic toroidally rotating tokamak plasmas by using an ideal magnetohydrodynamic model and standard double adiabatic equations. It is found that the radial wave number kx of the effectively excited GAM and the GAM frequency increase with the ratio of the perpendicular pressure to the parallel pressure χ as well as the toroidal rotation Mach number. The equilibrium anisotropy and the ratio of the radial scale length of the equilibrium density to the tokamak major radius are shown to play a significant role in the growth rate and frequency shift of GAM. The growth rate increases with the normalized GAM frequency first and then decreases to zero as the GAM frequency increases. When the GAM frequency increases again, the growth rate continues to increase but with a relatively smaller magnitude.

Linear simulation of magnetohydrodynamic plasma response to three-dimensional magnetic perturbations in high-β P plasmas

We report the numerical analyses of the linear magnetohydrodynamics (MHD) plasma response to applied three-dimensional magnetic perturbations (MPs) in a joint DIII-D/EAST collaboration on high-β P (poloidal beta) plasmas, utilizing the extended-MHD code M3D-C1, with the purpose of gaining a better understanding of the existing experiment in which n = 3 MPs were applied to such high-β P plasmas attempting to control large-amplitude type-I edge-localized modes (ELMs). These high-β P plasmas obtained at the DIII-D tokamak feature an upper-biased double-null configuration, a high edge safety factor q 95 ∼ 6.4, and a stable internal transport barrier (ITB), leading to relatively high core pressures. Single-fluid simulations show that the plasma response to n = 3 MPs, including both non-resonant/kinking and resonant components, is significantly weaker than that to n = 1 or 2 MPs. To survey the impact of q 95 on the plasma response to applied MPs, the self-consistent equilibrium-generating workflow for analysis module, developed in the OMFIT integrated modeling framework, is employed to generate a series of equilibria with a wide range of q 95, while other key parameters, including the normalized beta, electron density at the pedestal top, and plasma shape, are kept fixed. Compared to the vacuum response, single-fluid M3D-C1 simulations predict a much more significant decrease in resonant plasma response to the applied n = 3 MPs at the maximum penetration radii as q 95 increases. In contrast to single-fluid simulation results, showing that resonant penetration occurs only near the pedestal top where the E × B toroidal rotation frequency is zero, two-fluid simulations show two comparable resonant penetrations located near the pedestal top and the ITB foot, where the perpendicular electron rotation frequency is zero. Such resonant field penetration near the ITB foot may be responsible for the observed formation of a staircase structure in both the electron density and temperature profiles, and thereby a considerable deterioration in the global plasma performance, when MPs are applied in high-β P plasmas. Motivated by this numerical work, we provide some ideas for future research, with the purpose of realizing effective ELM control in such high-β P plasmas in the DIII-D and EAST devices.

Overview of recent experimental results from the ADITYA-U tokamak

Since the 2018 IAEA-FEC conference, in addition to expanding the parameter horizons of the ADITYA-U machine, emphasis has been given to dedicated experiments on inductively driven particle injection (IPI) for disruption studies, runaway electron (RE) dynamics and mitigation, plasma rotation reversal, radiative-improved modes using Ne and Ar injection, modulation of magneto–hydrodynamic modes, edge turbulence using periodic gas puffs and electrode biasing (E-B). Plasma parameters close to the design parameters of circular plasmas with H2 and D2 as fuel have been realized, and the shaped plasma operation has also been initiated. Consistent plasma discharges having I P ∼ 100–210 kA, t ∼ 300–400 ms, n e ∼ 3–6 × 1019 m−3, core T e ∼ 300–500 eV were achieved with a maximum B T of ∼1.5 T. The enhanced plasma parameters are the outcome of repeated cycles of baking (135 °C), followed by extensive wall conditioning, which includes pulsed glow discharge cleaning in H, He and Ar–H mixture, and lithiumization. A higher confinement time has been observed in D2 compared to H2 plasmas. Furthermore, shaped plasmas are attempted for the first time in ADITYA-U. A first of its kind inductively driven particle injection for disruption mitigation studies has been developed and operated. The injection of solid particles into the plasma core leads to a fast current quench. Two pulses of electron cyclotron resonance wave at 42 GHz are launched in a single discharge: one pulse is used for pre-ionization and the second for heating. In a novel approach, a positively biased electrode is used to confine REs after discharge termination. E-B is also used for controlling the rotation of drift-tearing modes by changing the plasma rotation. Cold pulse propagation and signatures of detachment are observed during the injection of short gas puffs. A correlation between the plasma toroidal rotation and the total radiated power has been observed with neon gas injection-induced improved confinement modes.

Physics studies of ADITYA & ADITYA-U tokamak plasmas using spectroscopic diagnostics

Several spectroscopic diagnostics encompassing the spectral emission range from the x-ray to near infrared (NIR) have been developed, installed and operated for diagnosing and physics studies in the ADITYA and ADITYA-U tokamaks. The recycling and impurity influxes and plasma Z eff after lithium (Li) coating have been studied using a PMT (photomultiplier tube)-filter based system by capturing H α , O1+, C2+, and visible continuum emissions. Significant reduction in the Z eff values has been observed in the discharge with the Li coated walls. The measured radial profile of H α emission using a filter-PMT array, has been modelled using a neutral transport code. The results show substantial contributions from the molecular hydrogen and molecular hydrogen ion dissociation (∼56%) and charge-exchange (∼30%) processes in the measured H α emission. Furthermore, a high-resolution, 1 m spectrometer with charge coupled device detector capable of multi-track measurements, has been used to study impurity transport, neutral and ion temperature and intrinsic plasma rotation. By modelling the measured radial profile of O4+ spectral line emission using an impurity transport code, substantial contribution of edge fluctuations on the oxygen transport has been observed. The toroidal ( ∼ 20 km s−1 in core) and poloidal ( ∼ 4.5 km s−1 at edge) rotation velocities are measured using C5+ (529 nm) and C2+ (464.7 nm) passive line emissions respectively. The measurement of radial profile of toroidal plasma rotation revealed a reversal of rotation direction depending on the electron density content of the ADITYA-U plasmas. The neutral temperature measurements showed a poloidal asymmetry indicating a presence of asymmetrical source of neutral heating. Moreover, the modelling of measured Fe14+ and Fe15+ vacuum ultraviolet spectral lines has revealed the neo-classical nature of iron transport in ADITYA core. Fast visible camera images captured the formation of filament structures triggered by interchange instabilities during plasma disruptions.

Dynamic measurement of impurity ion transport in runaway electron plateaus in DIII-D

The first dynamic (time-dependent) measurements of impurity ion radial (cross field) and parallel (along-field) diffusion coefficients for post-disruption runaway electron plateaus are presented. Small (∼1 mm diameter) carbon or silicon pellets are fired into the edge of steady-state runaway electron (RE) plateaus, and the resulting radial and toroidal transport of singly charged impurity ions (C+ or Si+) is monitored with spatially distributed visible spectrometers. Consistent with previous steady-state particle balance estimates of Ar+ radial transport, radial (cross field) diffusion coefficients D⊥≈2–5 m2/s are obtained, about 10× larger than expected from neo-classical theory. Parallel diffusion coefficients D∥≈30–80 m2/s are estimated, also much (≈50×) larger than classical. It is speculated at present that these large diffusion coefficients may be due to turbulent transport. Indications of fairly significant (almost 2×) toroidal variation in electron density are seen in the RE plateaus, and this appears to cause some toroidal variation in impurity radial diffusion rates. Indications of slow (≈1 Hz) toroidal rotation in the impurity ions are observed, although the uncertainty in this measurement is large.

Neoclassical transport analysis of high rotational trace limit tungsten impurities in KSTAR tokamak

The accumulation of tungsten impurities measured in a KSTAR experiment was analyzed theoretically using a drift-kinetic code, NEO, to determine the contribution of neoclassical transport. According to the NEO simulation results, there is a certain value of impurity toroidal rotation speed maximizing the neoclassical inward convection. The inward convection decreases or the outward convection increases as the rotation increases only beyond the speed value. The non-monotonic dependency of the neoclassical convection on the rotation is analyzed by the several coefficients for many profile effects, including ion and electron profiles. The dependency of the coefficients for the main ion density gradient on the rotation is different from that for the temperature gradient, so it results in the amplification of the temperature screening beyond the certain value of the rotation. In the KSTAR case with high toroidal rotation of the tungsten (around Mach number 4.5), only in the mid-radius does the rotation reduce the inward impurity particle convection or change the inward convection to the outward convection. Thus, the rotation is a useful tool to control the impurity accumulation conditionally. The favorable condition occurs only for high rotation, which significantly depends on the radius and the collisionality due to the complicated non-monotonic dependency of the convection on the rotation speed.

First Results From Simulations of Rapid Shutdown With Neon Deposition in J-TEXT Rotating Plasmas

Encapsulated payload pellet injection (EPPI) has been demonstrated to be an effective method of disruption mitigation system (DMS) through a number of experiments and simulations in DIII-D and JET. This article investigates the effects of plasma rotation on impurity spreading under neon EPPI. Our analysis of simulations shows that the magnetic surface stochastization has a decisive influence on the core-oriented spread of impurity. The MHD instabilities increase with impurity ionization, especially of the <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$n = 1$ </tex-math></inline-formula> component, leading to the enhancement of the local transport of impurities along the stochastic field lines. In the rotating plasma, the toroidal rotation can suppress the magnetic perturbation, and then, the boundary-oriented spreading of impurities is reduced. When the pellet source is toroidally localized at the magnetic axis, the rotation can quickly flatten the toroidal distribution of impurities, resulting in a smaller maximum radiated power and a lower toroidal peek factor (TPF).

Integrated control of individual plasma scalars with simultaneous neoclassical tearing-mode suppression

A novel integrated-control architecture has been tested in nonlinear, one-dimensional simulations using the control-oriented transport simulator (COTSIM©) and in DIII-D experiments. Integrated architectures that can perform continuous-mission control while also handling off-normal events will be vital in future reactor-grade tokamaks. Continuous-mission controllers for individual magnetic and kinetic scalars (thermal stored-energy (W), volume-average toroidal rotation (Ω ϕ ), and safety factor profile (q) at different spatial locations) have been integrated in this work with event-triggered neoclassical tearing-mode (NTM) suppression controllers by combining them into an architecture augmented by a supervisory and exception handling (S&EH) system and an actuator management (AM) system. The AM system, which enables the integration of competing controllers, solves in real time a nonlinear optimization problem that takes into account the high-level control priorities dictated by the S&EH system. The resulting architecture offers a high level of integration and some of the functionalities that will be required to fulfill the advanced-control requirements anticipated for ITER. Initial simulations using COTSIM suggest that the plasma performance and its MHD stability may be improved under integrated feedback control. In addition, the integrated-control architecture has been implemented in the DIII-D plasma control system and tested experimentally for the first time ever in DIII-D in a high-q min scenario, which is a candidate for steady-state operation in ITER.

Influence of toroidal rotation on plasma response to external RMP fields in tokamak

The type-I edge localized mode (ELM) is a critical event associated with magneto-hydrodynamic(MHD) instabilities occurring in tokamak high-confinement (H-mode) discharges, that leads to huge heat loads on the plasma phasing components (PFC) and may result in material damages. It is important to effectively control large ELMs, in order to ensure safe operation of the future reactor-scale devices such as ITER and DEMO. Resonant magnetic perturbation (RMP) has been experimentally demonstrated to be a mature and robust technique for controlling ELMs. A set of parameters, such as the edge safety factor, the plasma flow, the RMP coil geometry and the spectrum of the applied external field, have been found to play important roles in controlling ELMs by RMP. Furthermore, the plasma pressure is known to affect the plasma response to the RMP field, in particular near the no-wall beta limit. This is because high plasma pressure drives the resonant field amplification of the external field by the plasma response. The ITER 10 MA steady state scenario will be operated near the no-wall stability limit. The new tokamak device HL-2M will also operate in the relatively high-beta regimes. On the other hand, more investigations are still needed to understand the influence of toroidal flow on the high-beta plasma response. This work employs a single fluid toroidal model to compute the plasma RMP response in HL-2M, emphasizing on the roles of two key physical quantities: the plasma resistivity and the toroidal rotation. The former allows penetration of the external RMP field into the plasma, while the latter mainly provides screening effect on the resonant field component. More specifically, the MARS-F code is utilized to study the plasma response to the externally applied &lt;i&gt;n&lt;/i&gt; =1 ( n is the toroidal mode number) RMP field for high-beta HL-2M discharges, while varying the plasma toroidal rotation profile. The plasma response is found to (i) substantially modify the poloidal spectrum of the applied vacuum RMP field, (ii) change the amplitude of the resonant radial field amplitude near the plasma edge, and (iii) affect optimal current phasing between the two rows of RMP coils on HL-2M. A sufficiently slow toroidal flow near the plasma edge amplifies the radial field at rational surfaces associated with the perturbation. Since the latter serves as a reliable indicator for controlling the type-I edge localized mode (Type-I ELM) by RMP, varying rotation profile near the plasma edge offers a promising approach to optimize ELM control.

Neoclassical toroidal viscosity torque prediction via deep learning

GPECnet is a densely connected neural network that has been trained on GPEC data, to predict the plasma stability, neoclassical toroidal viscosity (NTV) torque, and optimized 3D coil current distributions for desired NTV torque profiles. Using NTV torque, driven by non-axisymmetric field perturbations in a tokamak, can be vital in optimizing pedestal performance by controlling the rotation profile in both the core, to ensure tearing stability, and the edge, to avoid edge localized modes (ELMs). The generalized perturbed equilibrium code (GPEC) software package can be used to calculate the plasma stability to 3D perturbations and the NTV torque profile generated by applied 3D magnetic fields. These calculations, however, involve complex integrations over space and energy distributions, which takes time to compute. Initially, GPECnet has been trained solely on data representative of the quiescent H-mode (QH) scenario, in which neutral beams are often balanced and toroidal rotation is low across the plasma profile. This work provides the foundation for active control of the rotation shear using a combination of beams and 3D fields for robust and high performance QH mode operation.