Edge Localized Modes Suppression Research Articles

Ion temperature and toroidal velocity edge transport barriers in KSTAR

The structure and evolution of the ion temperature () and toroidal rotation () profile have been investigated in neutral beam injection (NBI)-heated KSTAR H-mode plasmas, both without and with resonant magnetic pertubations (RMPs). A clear disparity between the width of the -pedestal and that of the -pedestal was observed. Also, it was found that there exists a close correlation and weak relative hysteresis between the pedestal and during both L → H and H → L transitions. During the L → H transition, the -pedestal is observed to form ahead of the -pedestal, and build inward from the separatrix. Linear gyrokinetic stability analysis of these KSTAR profiles was performed. The results indicate that parallel velocity shear is a relevant drive for pedestal turbulence and transport. This was largely ignored in previous studies of the pedestal micro-stability. Pedestal ion temperature and rotation profiles were also measured during edge localized mode (ELM) suppression experiments on KSTAR using an n = 1 RMPs. It was found that the top values of the ion temperature and toroidal rotation pedestal drop with RMPs when ELMs are suppressed.

Fast ion transport during applied 3D magnetic perturbations on DIII-D

Measurements show fast ion losses correlated with applied three-dimensional (3D) fields in a variety of plasmas ranging from L-mode to resonant magnetic perturbation (RMP) edge localized mode (ELM) suppressed H-mode discharges. In DIII-D L-mode discharges with a slowly rotating magnetic perturbation, scintillator detector loss signals synchronized with the applied fields are observed to decay within one poloidal transit time after beam turn-off indicating they arise predominantly from prompt loss orbits. Full orbit following using M3D-C1 calculations of the perturbed fields and kinetic profiles reproduce many features of the measured losses and points to the importance of the applied 3D field phase with respect to the beam injection location in determining the overall impact on prompt beam ion loss. Modeling of these results includes a self-consistent calculation of the 3D perturbed beam ion birth profiles and scrape-off-layer ionization, a factor found to be essential to reproducing the experimental measurements. Extension of the simulations to full slowing down timescales, including fueling and the effects of drag and pitch angle scattering, show the applied RMPs in ELM suppressed H-mode plasmas can induce a significant loss of energetic particles from the core. With the applied fields, up to 8.4% of the injected beam power is predicted to be lost, compared to 2.7% with axisymmetric fields only. These fast ions, originating from minor radii , are predicted to be primarily passing particles lost to the divertor region, consistent with wide field-of-view infrared periscope measurements of wall heating in RMP ELM suppressed plasmas. Edge fast ion (FIDA) measurements also confirm a large change in edge fast ion profile due to the fields, where the effect was isolated by using short 50 ms RMP-off periods during which ELM suppression was maintained yet the fast ion profile was allowed to recover. The role of resonances between fast ion drift motion and the applied 3D fields in the context of selectively targeting regions of fast ion phase space is also discussed.

3D toroidal physics: Testing the boundaries of symmetry breakinga)

Toroidal symmetry is an important concept for plasma confinement; it allows the existence of nested flux surface MHD equilibria and conserved invariants for particle motion. However, perfect symmetry is unachievable in realistic toroidal plasma devices. For example, tokamaks have toroidal ripple due to discrete field coils, optimized stellarators do not achieve exact quasi-symmetry, the plasma itself continually seeks lower energy states through helical 3D deformations, and reactors will likely have non-uniform distributions of ferritic steel near the plasma. Also, some level of designed-in 3D magnetic field structure is now anticipated for most concepts in order to provide the plasma control needed for a stable, steady-state fusion reactor. Such planned 3D field structures can take many forms, ranging from tokamaks with weak 3D edge localized mode suppression fields to stellarators with more dominant 3D field structures. This motivates the development of physics models that are applicable across the full range of 3D devices. Ultimately, the questions of how much symmetry breaking can be tolerated and how to optimize its design must be addressed for all fusion concepts. A closely coupled program of simulation, experimental validation, and design optimization is required to determine what forms and amplitudes of 3D shaping and symmetry breaking will be compatible with the requirements of future fusion reactors.

Extremely low intrinsic non-axisymmetric field in KSTAR and its implications

A surprisingly low level of intrinsic non-axisymmetric field (called ‘error field’) has been measured in KSTAR, suggesting at least an order of magnitude lower than in other major tokamaks. Specifically, the KSTAR was found to have an extremely low level of pitch resonant intrinsic error field at the m/n = 2/1 surface in the order of 10−5 of the magnetic field at the geometric centre, instead of 10−4 typically observed in other devices. Using a single array of in-vessel control coils (IVCCs) at the outboard midplane, the n = 1 intrinsic error field was diagnosed. Such a low level of intrinsic non-axisymmetric field as measured in KSTAR is less than or comparable to the Earth's magnetic field or a remanent field in the KSTAR plasma chamber. Considering that a very low level of n = 1 intrinsic error field (mostly associated with kink-resonance) helps the plasma to be less vulnerable to mode-locking, this might have allowed the n = 1 resonant magnetic perturbation (RMP) currents (configured to be dominantly pitch-resonant for edge localized mode (ELM) suppression) to increase without invoking a kink-resonant mode-locking, consistent with experimental observation and poloidal mode spectra calculations in KSTAR. Further clarification of the influence of the intrinsic error field in terms of a 3D structure is expected to provide a solid foundation to understand the n = 1 RMP-driven ELM suppression uniquely observed in KSTAR.

Pedestal bifurcation and resonant field penetration at the threshold of edge-localized mode suppression in the DIII-D Tokamak.

Rapid bifurcations in the plasma response to slowly varying n=2 magnetic fields are observed as the plasma transitions into and out of edge-localized mode (ELM) suppression. The rapid transition to ELM suppression is characterized by an increase in the toroidal rotation and a reduction in the electron pressure gradient at the top of the pedestal that reduces the perpendicular electron flow there to near zero. These events occur simultaneously with an increase in the inner-wall magnetic response. These observations are consistent with strong resonant field penetration of n=2 fields at the onset of ELM suppression, based on extended MHD simulations using measured plasma profiles. Spontaneous transitions into (and out of) ELM suppression with a static applied n=2 field indicate competing mechanisms of screening and penetration of resonant fields near threshold conditions. Magnetic measurements reveal evidence for the unlocking and rotation of tearinglike structures as the plasma transitions out of ELM suppression.

Observation of a multimode plasma response and its relationship to density pumpout and edge-localized mode suppression.

Density pumpout and edge-localized mode (ELM) suppression by applied n=2 magnetic fields in low-collisionality DIII-D plasmas are shown to be correlated with the magnitude of the plasma response driven on the high-field side (HFS) of the magnetic axis but not the low-field side (LFS) midplane. These distinct responses are a direct measurement of a multimodal magnetic plasma response, with each structure preferentially excited by a different n=2 applied spectrum and preferentially detected on the LFS or HFS. Ideal and resistive magneto-hydrodynamic (MHD) calculations find that the LFS measurement is primarily sensitive to the excitation of stable kink modes, while the HFS measurement is primarily sensitive to resonant currents (whether fully shielding or partially penetrated). The resonant currents are themselves strongly modified by kink excitation, with the optimal applied field pitch for pumpout and ELM suppression significantly differing from equilibrium field alignment.

A Flexible System for the Control of External Magnetic Perturbations in the JET Tokamak

External magnetic perturbations are typically utilized in tokamak devices with two operational or experimental purposes: 1) correction of intrinsic 3-D error fields and 2) mitigation or suppression of edge localized modes (ELMs). At Joint European Torus (JET), dedicated coils are used for the generation of these toroidally asymmetric perturbations. While error fields exist even in the absence of plasma, in ELM mitigation experiments, the external fields are meant to slightly ergodize the magnetic topology in the plasma periphery hence reducing the drive for the destabilization of these instabilities. The control of the magnetic field produced by these coils is achieved by controlling the current flowing in them. The real-time system responsible for this control recently underwent a number of functional improvements since its original implementation utilizing the present voltage-controlled voltage sources. This paper describes the overall system, built-in functionality, and control algorithms and presents preliminary experimental results along with performance assessment studies. In particular, the main improvements are: 1) the possibility of automatically reducing the current references in case the plasma amplifies the applied perturbation; 2) a real-time limitation of $dI/dt$ to reduce the electromotive force in machine protection diagnostic systems; 3) implementation of a model predictive controller as an alternative to the proportional integral derivative; and 4) the possibility of adapting the current references, in real time, using an external system. The result is a flexible control system contributing toward state-of-the-art physics research at JET’s international and dynamic scientific environment.

Advances in the physics understanding of ELM suppression using resonant magnetic perturbations in DIII-D

Recent experiments on DIII-D have increased confidence in the ability to suppress edge-localized modes (ELMs) using edge-resonant magnetic perturbations (RMPs) in ITER, including an improved physics basis for the edge response to RMPs as well as expansion of RMP ELM suppression to more ITER-like conditions. Complete ELM suppression has been achieved utilizing n = 3 RMPs in the ITER baseline scenario. In addition, RMP ELM suppression has been expanded to include plasmas with helium concentrations near 25% and the use of n = 2 RMPs. Analysis of the kinetic profile response suggests that ELM suppression is correlated with the co-alignment of the ω⊥e = 0 location, an n = 3 rational surface, and the top of the pedestal. Modelling predicts that such a co-alignment could potentially lead to island (or island chain) formation just inside the top of the pedestal, inhibiting the growth of the pedestal and thereby maintaining the ELM-free state. Detailed analysis of data obtained during toroidal phase variations of the applied n = 3 RMPs have provided further evidence of an island-like structure at the top of the pedestal. In addition, nearly matched discharges with co-neutral-beam-injection (co-NBI) and counter-NBI have demonstrated the importance of the presence of the ω⊥e = 0 location for ELM suppression. In the counter-NBI cases, the toroidal rotation profile is such that there is no ω⊥e = 0 location and ELMs are not suppressed in conditions in which ELM suppression is generally observed with co-NBI.

Long-pulse stability limits of the ITER baseline scenario

DIII-D has made significant progress in developing the techniques required to operate ITER, and in understanding their impact on performance when integrated into operational scenarios at ITER-relevant parameters. Long duration plasmas, stable to m/n = 2/1 tearing modes, with an ITER-similar shape and Ip/aBT, have been demonstrated in DIII-D, that evolve to stationary conditions. The operating region most likely to reach stable conditions has normalized pressure, βN ≈ 1.9–2.1 (compared to the ITER baseline design of 1.6–1.8), and a Greenwald normalized density fraction, fGW 0.42–0.70 (the ITER design is fGW ≈ 0.8). The evolution of the current profile, using internal inductance (li) as an indicator, is found to produce a smaller fraction of stable pulses when li is increased above ≈1.1 at the beginning of βN flattop. Stable discharges with co-neutral beam injection are generally accompanied with a benign n = 2 magnetohydrodynamic mode. However if this mode exceeds ≈10 G, the onset of a m/n = 2/1 tearing mode occurs with a loss of confinement. In addition, stable operation with low applied external torque, at or below the extrapolated value expected for ITER has also been demonstrated. With electron cyclotron injection, the operating region of stable discharges has been further extended at ITER equivalent levels of torque and to edge-localized mode (ELM) free discharges at higher torque but with the addition of an n = 3 magnetic perturbation from the DIII-D internal coil set. The characterization of the ITER baseline scenario evolution for long pulse duration, extension to more ITER-relevant values of torque and electron heating, and suppression of ELMs have significantly advanced the physics basis of this scenario, although significant effort remains in the simultaneous integration of all these requirements.

Mechanism of edge localized mode mitigation by resonant magnetic perturbations.

A possible mechanism of edge localized modes (ELMs) mitigation by resonant magnetic perturbations (RMPs) is proposed based on the results of nonlinear resistive magnetohydrodynamic modeling using the jorek code, realistic JET-like plasma parameters and an RMP spectrum of JET error-field correction coils (EFCC) with a main toroidal number n=2 were used in the simulations. Without RMPs, a large ELM relaxation is obtained mainly due to the most unstable medium-n ballooning mode. The externally imposed RMP drives nonlinearly the modes coupled to n=2 RMP which produce small multimode relaxations, mitigated ELMs. The modes driven by RMPs exhibit a tearinglike structure and produce additional islands. Mitigated ELMs deposit energy into the divertor mainly in the structures ("footprints") created by n=2 RMPs, however, slightly modulated by other nonlinearly driven even harmonics. The divertor power flux during a ELM phase mitigated by RMPs is reduced almost by a factor of 10. The mechanism of ELM mitigation by RMPs proposed here reproduces generic features of high collisionality RMP experiments, where large ELMs are replaced by small, much more frequent ELMs or magnetic turbulence. Total ELM suppression was also demonstrated in modeling at higher RMP amplitude.

Edge-localized-modes in tokamaks

Edge-localized-modes (ELMs) are a ubiquitous feature of H-mode in tokamaks. When gradients in the H-mode transport barrier grow to exceed the MHD stability limit the ELM instability grows explosively, rapidly transporting energy and particles onto open field lines and material surfaces. Though ELMs provide additional particle and impurity transport through the H-mode transport barrier, enabling steady operation, the resulting heat flux transients to plasma facing surfaces project to large amplitude in future low collisionality burning plasma tokamaks. Measurements of the ELM heat flux deposition onto material surfaces in the divertor and main chamber indicate significant broadening compared to inter-ELM heat flux, with a timescale for energy deposition that is consistent with sonic ion flow and numerical simulation. Comprehensive ELM simulation is highlighting the important physics processes of ELM transport including parallel transport due to magnetic reconnection and turbulence resulting from collapse of the H-mode transport barrier. Encouraging prospects for ELM control and/or suppression in future tokamaks include intrinsic modes of ELM free operation, ELM triggering with frequent small pellet injection and the application of 3D magnetic fields.

Rotation characteristics during the resonant magnetic perturbation induced edge localized mode suppression on the KSTAR.

Measuring rotation profiles with a reliable spatial resolution is one of the critical diagnostics in understanding the plasma behavior especially for the edge transport. In the KSTAR experiments, it has been consistently observed from the charge exchange spectroscopy measurements that the magnetic perturbations not only suppresses edge localized modes (ELMs) but also reduces toroidal rotations. In this paper, toroidal velocities of the carbon impurity and their profile evolutions during ELMy and ELM-suppressed phases are presented. The rotation profiles are shown to collapse immediately after an ELM burst and continue to build up until the next burst that accompanies another collapse. Toroidal rotations following the resonant magnetic perturbations applications are observed to be reduced along with the ELMs suppressed.

The radial electric field as a measure for field penetration of resonant magnetic perturbations

In this paper we introduce a new indirect method for identifying the radial extent of the stochastic layer due to applying resonant magnetic perturbations (RMPs) in H-mode plasmas by measuring the spin-up of the plasma near the separatrix. This spin-up is a predicted consequence of enhanced electron loss, due to magnetic stochastization (Kaveeva et al 2008 Nucl. Fusion 48 075003). We find that in DIII-D H-mode plasmas with n = 3 RMPs applied for edge localized mode suppression, the stochastic layer is limited to the outer 5% region in normalized magnetic flux, ΨN. This is in contrast to vacuum modelling predictions where this layer can penetrate up to 20% in ΨN. Theoretical predictions of a stochastic radial electric field, Er component exceed the experimental measurements by about a factor 3 close to the separatrix, suggesting that the outer region of the plasma is weakly stochastic. Linear response calculations with M3D-C1, a resistive two-fluid model, show that in this outer 5% region, plasma response often reduces the resonant magnetic field components by 67% or more in comparison with vacuum calculations. These results for DIII-D are in reasonable agreement with results from the MAST tokamak, where the magnetic field perturbation from vacuum field calculations needed to be reduced by 75% for agreement with experimental measurements of the x-point lobe structures (Kirk et al 2012 Phys. Rev. Lett. 108 255003).

Pedestal Structure without and with 3D Fields

AbstractAchieving high pedestal pressures in H(high)‐mode plasmas confined in tokamaks is critical for obtaining fusion burning plasmas in ITER. Recent characterizations of quasi‐equilibrium plasma parameter profiles in low collisionality H‐mode pedestals in the DIII‐D tokamak are briefly summarized. Critical plasma transport properties (large radial electron heat flow, density pinch) that establish the transport barrier structure of the pedestal profiles are identified. The paleoclassical transport model, which naturally includes a density pinch, is shown to provide the minimum electron heat and density transport in the pedestal. Microinstabilities can provide additional plasma transport within and especially at the top of pedestals. Macroscopic peeling‐ballooning (P‐B) instabilities cause periodic edge localized modes (ELMs) that limit the temporal and spatial growth of the pedestal initially and between ELMs. Externally imposed 3D resonant magnetic perturbations (RMPs) in the pedestal have been used to stabilize P‐B modes and suppress ELMs. A magnetic flutter model of plasma transport induced by the 3D RMPs has been developed for low collisionality DIII‐D pedestals. Comparisons of it with data on ELM suppression by RMPs indicate it can provide a “diffusivity hill” at the pedestal top that can impede pedestal growth and thereby stabilize P‐B modes and suppress ELMs. Finally, transport equations for plasma density, electron and ion pressures and, most importantly, the plasma toroidal rotation frequency (and hence, via radial force balance, the radial electric field) in the presence of plasma transport due to collisional, paleoclassical, microturbulence‐induced and 3D field effects are presented. (© 2014 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Edge-localized mode avoidance and pedestal structure in I-mode plasmas

I-mode is a high-performance tokamak regime characterized by the formation of a temperature pedestal and enhanced energy confinement, without an accompanying density pedestal or drop in particle and impurity transport. I-mode operation appears to have naturally occurring suppression of large Edge-Localized Modes (ELMs) in addition to its highly favorable scalings of pedestal structure and overall performance. Extensive study of the ELMy H-mode has led to the development of the EPED model, which utilizes calculations of coupled peeling-ballooning MHD modes and kinetic-ballooning mode (KBM) stability limits to predict the pedestal structure preceding an ELM crash. We apply similar tools to the structure and ELM stability of I-mode pedestals. Analysis of I-mode discharges prepared with high-resolution pedestal data from the most recent C-Mod campaign reveals favorable pedestal scalings for extrapolation to large machines—pedestal temperature scales strongly with power per particle Pnet/n¯e, and likewise pedestal pressure scales as the net heating power (consistent with weak degradation of confinement with heating power). Matched discharges in current, field, and shaping demonstrate the decoupling of energy and particle transport in I-mode, increasing fueling to span nearly a factor of two in density while maintaining matched temperature pedestals with consistent levels of Pnet/n¯e. This is consistent with targets for increased performance in I-mode, elevating pedestal βp and global performance with matched increases in density and heating power. MHD calculations using the ELITE code indicate that I-mode pedestals are strongly stable to edge peeling-ballooning instabilities. Likewise, numerical modeling of the KBM turbulence onset, as well as scalings of the pedestal width with poloidal beta, indicates that I-mode pedestals are not limited by KBM turbulence—both features identified with the trigger for large ELMs, consistent with the observed suppression of large ELMs in I-mode.

Progress on the application of ELM control schemes to ITER scenarios from the non-active phase to DT operation

Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of ∼2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix. Therefore, specifications for the rotation of the 3D perturbation applied for ELM control in order to avoid excessive localized erosion of the ITER divertor target are derived. It is shown that a rotation frequency in excess of 1 Hz for the whole toroidally asymmetric divertor power flux pattern is required (corresponding to n Hz frequency in the variation of currents in the coils, where n is the toroidal symmetry of the perturbation applied) in order to avoid unacceptable thermal cycling of the divertor target for the highest power fluxes and worst toroidal power flux asymmetries expected. The possible use of the in-vessel vertical stability coils for ELM control as a back-up to the main ELM control systems in ITER is described and the feasibility of its application to control ELMs in low plasma current H-modes, foreseen for initial ITER operation, is evaluated and found to be viable for plasma currents up to 5–10 MA depending on modelling assumptions.

Understanding the effect resonant magnetic perturbations have on ELMs

All current estimations of the energy released by type I edge-localized modes (ELMs) indicate that, in order to ensure an adequate lifetime of the divertor targets on ITER, a mechanism is required to decrease the amount of energy released by an ELM, or to eliminate ELMs altogether. One such amelioration mechanism relies on perturbing the magnetic field in the edge plasma region, either leading to more frequent, smaller ELMs (ELM mitigation) or ELM suppression. This technique of resonant magnetic perturbations (RMPs) has been employed to suppress type I ELMs at high collisionality/density on DIII-D, ASDEX Upgrade, KSTAR and JET and at low collisionality on DIII-D. At ITER-like collisionality the RMPs enhance the transport of particles or energy and keep the edge pressure gradient below the 2D linear ideal magnetohydrodynamic critical value that would trigger an ELM, whereas at high collisionality/density the type I ELMs are replaced by small type II ELMs. Although ELM suppression only occurs within limited operational ranges, ELM mitigation is much more easily achieved. The exact parameters that determine the onset of ELM suppression are unknown but in all cases the magnetic perturbations produce 3D distortions to the plasma and enhanced particle transport. The incorporation of these 3D effects in codes will be essential in order to make quantitative predictions for future devices.

Surface chemistry analysis of lithium conditioned NSTX graphite tiles correlated to plasma performance

Lithium wall conditioning in NSTX has resulted in reduced divertor recycling, improved energy confinement, and reduced frequency of edge-localized modes (ELMs), up to the point of complete ELM suppression. NSTX tiles were removed from the vessel following the 2008 campaign and subsequently analyzed using X-ray photoelectron spectroscopy as well as nuclear reaction ion beam analysis. In this paper we relate surface chemistry to deuterium retention/recycling, develop methods for cleaning of passivated NSTX tiles, and explore a method to effectively extract bound deuterium from lithiated graphite. Li–O–D and Li–C–D complexes characteristic of deuterium retention that form during NSTX operations are revealed by sputter cleaning and heating. Heating to ∼850°C desorbed all deuterium complexes observed in the O 1s and C 1s photoelectron energy ranges. Tile locations within approximately ±2.5cm of the lower vertical/horizontal divertor corner appear to have unused LiO bonds that are not saturated with deuterium, whereas locations immediately outboard of this region indicate high deuterium recycling. X-ray photo electron spectra of a specific NSTX tile with wide ranging lithium coverage indicate that a minimum lithium dose, 100–500nm equivalent thickness, is required for effective deuterium retention. This threshold is suspected to be highly sensitive to surface morphology. The present analysis may explain why plasma discharges in NSTX continue to benefit from lithium coating thickness beyond the divertor deuterium ion implantation depth, which is nominally <10nm.

Overview of ASDEX Upgrade results

The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron resonance heating (ECRH) power, by installing 2 × 8 internal magnetic perturbation coils, and by improving the ion cyclotron range of frequency compatibility with the tungsten wall. With the perturbation coils, reliable suppression of large type-I edge localized modes (ELMs) could be demonstrated in a wide operational window, which opens up above a critical plasma pedestal density. The pellet fuelling efficiency was observed to increase which gives access to H-mode discharges with peaked density profiles at line densities clearly exceeding the empirical Greenwald limit. Owing to the increased ECRH power of 4 MW, H-mode discharges could be studied in regimes with dominant electron heating and low plasma rotation velocities, i.e. under conditions particularly relevant for ITER. The ion-pressure gradient and the neoclassical radial electric field emerge as key parameters for the transition. Using the total simultaneously available heating power of 23 MW, high performance discharges have been carried out where feed-back controlled radiative cooling in the core and the divertor allowed the divertor peak power loads to be maintained below 5 MW m−2. Under attached divertor conditions, a multi-device scaling expression for the power-decay length was obtained which is independent of major radius and decreases with magnetic field resulting in a decay length of 1 mm for ITER. At higher densities and under partially detached conditions, however, a broadening of the decay length is observed. In discharges with density ramps up to the density limit, the divertor plasma shows a complex behaviour with a localized high-density region in the inner divertor before the outer divertor detaches. Turbulent transport is studied in the core and the scrape-off layer (SOL). Discharges over a wide parameter range exhibit a close link between core momentum and density transport. Consistent with gyro-kinetic calculations, the density gradient at half plasma radius determines the momentum transport through residual stress and thus the central toroidal rotation. In the SOL a close comparison of probe data with a gyro-fluid code showed excellent agreement and points to the dominance of drift waves. Intermittent structures from ELMs and from turbulence are shown to have high ion temperatures even at large distances outside the separatrix.

An overview of KSTAR results

Since the first H-mode discharges in 2010, the duration of the H-mode state has been extended and a significantly wider operational window of plasma parameters has been attained. Using a second neutral beam (NB) source and improved tuning of equilibrium configuration with real-time plasma control, a stored energy of Wtot ∼ 450 kJ has been achieved with a corresponding energy confinement time of τE ∼ 163 ms. Recent discharges, produced in the fall of 2012, have reached plasma βN up to 2.9 and surpassed the n = 1 ideal no-wall stability limit computed for H-mode pressure profiles, which is one of the key threshold parameters defining advanced tokamak operation. Typical H-mode discharges were operated with a plasma current of 600 kA at a toroidal magnetic field BT = 2 T. L–H transitions were obtained with 0.8–3.0 MW of NB injection power in both single- and double-null configurations, with H-mode durations up to ∼15 s at 600 kA of plasma current. The measured power threshold as a function of line-averaged density showed a roll-over with a minimum value of ∼0.8 MW at . Several edge-localized mode (ELM) control techniques during H-mode were examined with successful results including resonant magnetic perturbation, supersonic molecular beam injection (SMBI), vertical jogging and electron cyclotron current drive injection into the pedestal region. We observed various ELM responses, i.e. suppression or mitigation, depending on the relative phase of in-vessel control coil currents. In particular, with the 90° phase of the n = 1 RMP as the most resonant configuration, a complete suppression of type-I ELMs was demonstrated. In addition, fast vertical jogging of the plasma column was also observed to be effective in ELM pace-making. SMBI-mitigated ELMs, a state of mitigated ELMs, were sustained for a few tens of ELM periods. A simple cellular automata (‘sand-pile’) model predicted that shallow deposition near the pedestal foot induced small-sized high-frequency ELMs, leading to the mitigation of large ELMs. In addition to the ELM control experiments, various physics topics were explored focusing on ITER-relevant physics issues such as the alteration of toroidal rotation caused by both electron cyclotron resonance heating (ECRH) and externally applied 3D fields, and the observed rotation drop by ECRH in NB-heated plasmas was investigated in terms of either a reversal of the turbulence-driven residual stress due to the transition of ion temperature gradient to trapped electron mode turbulence or neoclassical toroidal viscosity (NTV) torque by the internal kink mode. The suppression of runaway electrons using massive gas injection of deuterium showed that runaway electrons were avoided only below 3 T in KSTAR. Operation in 2013 is expected to routinely exceed the n = 1 ideal MHD no-wall stability boundary in the long-pulse H-mode (⩾10 s) by applying real-time shaping control, enabling n = 1 resistive wall mode active control studies. In addition, intensive works for ELM mitigation, ELM dynamics, toroidal rotation changes by both ECRH and NTV variations, have begun in the present campaign, and will be investigated in more detail with profile measurements of different physical quantities by techniques such as electron cyclotron emission imaging, charge exchange spectroscopy, Thomson scattering and beam emission spectroscopy diagnostics.