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We report on the observation of spatially asymmetric turbulent structures with a long radial correlation length in the core of high-collisionality H -mode plasmas on DIII-D tokamak. These turbulent structures develop from shorter wavelength turbulence and have a radially elongated structure. The envelope of turbulence spans a broad radial range in the mid-radius region, leading to streamer-like transport events. The underlying turbulence is featured by intermittency, long-term memory effect, and the characteristic spectrum of self-organized criticality. The amplitude and the radial scale increase substantially when the shearing rate of the mean flow is reduced below the turbulent scattering rate. The enhanced long-radial-range-correlated (LRRC) transport events are accompanied by apparent degradation of normalized energy confinement time. The emergence of such LRRC transport events may serve as a candidate explanation for the degrading nature of H-mode core plasma confinement at high collisionality on DIII-D tokamak. [ABSTRACT FROM AUTHOR]

The multi-machine International Tokamak Physics Activity (ITPA) Global H-mode Confinement Database has been upgraded with new data from JET with the ITER-like wall and ASDEX Upgrade with the full tungsten wall. This paper describes the new database and presents results of regression analysis to estimate the global energy confinement scaling in H-mode plasmas using a standard power law. Various subsets of the database are considered, focusing on type of wall and divertor materials, confinement regime (all H-modes, ELMy H or ELM-free) and ITER-like constraints. Apart from ordinary least squares (OLS), two other, robust regression techniques are applied, which take into account uncertainty on all variables. Regression on data from individual devices shows that, generally, the confinement dependence on density and the power degradation are weakest in the fully metallic devices. Using the multi-machine scalings, predictions are made of the confinement time in a standard ELMy H-mode scenario in ITER. The uncertainty on the scaling parameters is discussed with a view to practically useful error bars on the parameters and predictions. One of the derived scalings for ELMy H-modes on an ITER-like subset is studied in particular and compared to the IPB98(y, 2) confinement scaling in engineering and dimensionless form. Transformation of this new scaling from engineering variables to dimensionless quantities is shown to result in large error bars on the dimensionless scaling. Regression analysis in the space of dimensionless variables is therefore proposed as an alternative, yielding acceptable estimates for the dimensionless scaling. The new scaling, which is dimensionally correct within the uncertainties, suggests that some dependencies of confinement in the multi-machine database can be reconciled with parameter scans in individual devices. This includes vanishingly small dependence of confinement on line-averaged density and normalized plasma pressure (β), as well as a noticeable, positive dependence on effective atomic mass and plasma triangularity. Extrapolation of this scaling to ITER yields a somewhat lower confinement time compared to the IPB98(y, 2) prediction, possibly related to the considerably weaker dependence on major radius in the new scaling (slightly above linear). Further studies are needed to compare more flexible regression models with the power law used here. In addition, data from more devices concerning possible 'hidden variables' could help to determine their influence on confinement, while adding data in sparsely populated areas of the parameter space may contribute to further disentangling some of the global confinement dependencies in tokamak plasmas. [ABSTRACT FROM AUTHOR]

The Multi-Mode Model (MMM7.1) for anomalous transport is tested in predictive modeling of temperature profiles of a high beta poloidal DIII-D discharge. This new H-mode plasma regime, with high beta poloidal and high bootstrap currents, has been studied in DIII-D tokamak discharges [A. Garofalo et al., Nucl. Fusion 55, 123025 (2015)]. The role of instabilities that can drive the anomalous transport described by MMM7.1 is investigated. The temperature profiles for a high beta poloidal DIII-D discharge are computed using the NCLASS model for the neoclassical transport and the Weiland and Electron Temperature Gradient (ETG) components of the MMM7.1 model for the anomalous transport. The neoclassical transport is found to be the main contributor to the ion thermal transport in the plasma core. The contributions from the ion temperature gradient driven modes are found to be important only outside of the internal transport barrier. The magnitudes of the predicted temperature profiles are found to be in a reasonable agreement with experimental profiles. The simulation results approximately reproduce the internal transport barrier in the ion temperature profile but not in the electron temperature profile due to a weak dependence of the ETG driven transport on the Shafranov shift in the ETG component of MMM7.1. Possible effects that can contribute to stabilization of these modes, for example, effects associated with the large beta poloidal such as the Shafranov shift stabilization in the MMM7.1 model, are discussed. It is demonstrated that the E × B flow shear has a relatively small effect in the formation of the internal transport barrier in the high beta poloidal DIII-D discharge 154406. The Shafranov shift (alpha stabilization) and small or reversed magnetic shear profiles are found to be the primary reasons for quenched anomalous transport in this discharge. [ABSTRACT FROM AUTHOR]

H-mode plasma operation of the Korea Superconducting Tokamak Advanced Research (KSTAR) device has been expanded to significantly surpass the ideal MHD no-wall beta limit. Plasmas with high normalized beta, βN, up to 4.3 have been achieved with reduced plasma internal inductance, li, to near 0.7, exceeding the computed n=1 ideal no-wall limit by a factor of 1.6. Pulse lengths at maximum βN were extended to longer pulses by new, more rapid control. The stability of the observed m/n=2/1 tearing mode that limited the achieved high βN is computed by the M3D-C¹ code, and the effect of sheared toroidal rotation to tearing stability is examined. As a method to affect the mode stability in high βN plasmas, the non-resonant alteration of the rotation profile by non-axisymmetric magnetic fields has been used, enabling a study of the underlying neoclassical toroidal viscosity (NTV) physics and stability dependence on rotation. Non-axisymmetric field spectra were applied using in-vessel control coils (IVCCs) with varied n=2 field configurations to alter the plasma toroidal rotation profile in high beta H-mode plasmas and to analyze their effects on the rotation. The rotation profile was significantly altered with rotation reduced by more than 60% without tearing activity or mode locking. To investigate the physical characteristics and scaling of the measured rotation braking by NTV, changes in the rotation profile are analytically examined in steady state. The expected NTV scaling with the square of the normalized applied field perturbation agrees with the measured profile change δB2.1-2.3. The NTV is also found to scale as Ti2.1-2.4, in general agreement with the low collisionality "1/ν" regime scaling of the NTV theory (TNTV-(1/v) ∝Ti 2.5). [ABSTRACT FROM AUTHOR]

To obtain a quantitative understanding of the recent radiative divertor experiments and to give an instructive prediction for the future relative work performed on EAST, the 2D numerical tool SOLPS has been used to investigate the argon seeded EAST H-mode plasmas. The simulations were mainly based on a typical H-mode discharge with lower single null divertor configuration, in which the partially detached divertor plasma has been achieved due to the argon seeding. First the perpendicular particle and energy transport coefficients for particle density diffusivity D⊥, electron and ion conduction, χ⊥i,e were radially varied to determine the edge transport barrier in H-mode plasma by comparing the simulated upstream profiles of electron temperature (Te) and electron density (ne) with the edge Thomson scattering data. Then the reduction of the particle flux, static pressure, and the peak heat load onto the lower outer divertor target have been reproduced by radiative divertor simulations in agreement with the experimental measurements, which demonstrated that the argon seeding is effective to mitigate the heat load onto the divertor target and achieve partially detached plasma during the H-mode discharge. The simulations also revealed that the strongest radiation occurred near the X-point and in the divertor region for short pulsed argon seeding with a relatively low puffing rate in agreement with bolometer measurements. [ABSTRACT FROM AUTHOR]

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment of the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3–5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne , and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D2 -seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D2 -seeded SF divertor at P\text {SOL}=3-4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected P\text {SOL} \simeq 9 MW case. The radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q\text {peak} reduction factors (see standard divertor). [ABSTRACT FROM AUTHOR]

High confinement mode (H-mode) in tokamaks is characterized by the formation of a transport barrier at the edge of the plasma. The H-mode increases the plasma energy confinement time by around a factor of two compared to the Low confinement mode (L-mode). This result made a great step towards achieving the higher temperatures and pressures needed to create ignition conditions. The Grad-Shafranov Equation describes the magnetic flux distribution of plasma in an axisymmetric system like a tokamak. In this paper, we have calculated the magnetic flux surfaces in IR-T1 tokamak by two methods. An analytical solution of magnetic surfaces approximated by equilibrium calculation of GS equation based on first three (and then four) terms of expansion in flux function. Results of two methods are in good agreement with each other. [ABSTRACT FROM AUTHOR]

The influence of current density and pressure gradient profiles in the pedestal on the access to the regimes free from edge localized modes (ELMs) like quiescent H-mode in ITER is investigated. Using the simulator of MHD modes localized near plasma boundary based on the KINX code, calculations of the ELM stability were performed for the ITER plasma in scenarios 2 and 4 under variations of density and temperature profiles with the self-consistent bootstrap current in the pedestal. Low pressure gradient values at the separatrix, the same position of the density and temperature pedestals and high poloidal beta values facilitate reaching high current density in the pedestal and a potential transition into the regime with saturated large scale kink modes. New version of the localized MHD mode simulator allows one to compute the growth rates of ideal peeling-ballooning modes with different toroidal mode numbers and to determine the stability region taking into account diamagnetic stabilization. The edge stability diagrams computations and sensitivity studies of the stability limits to the value of diamagnetic frequency show that diamagnetic stabilization of the modes with high toroidal mode numbers can help to access the quiescent H-mode even with high plasma density but only with low pressure gradient values at the separatrix. The limiting pressure at the top of the pedestal increases for higher plasma density. With flat density profile the access to the quiescent H-mode is closed even with diamagnetic stabilization taken into account, while toroidal mode numbers of the most unstable peeling-ballooning mode decrease from n = 10−40 to n = 3−20. [ABSTRACT FROM AUTHOR]

The characteristics of high-confinement mode (H-mode) pedestal are examined on the Experimental Advanced Superconducting Tokamak. It is found that they are closely dependent on each other between electron pedestal characteristics and global parameters for all types of edge localized mode (ELM). The scaling of pedestal temperature based on thermal conduction and pedestal pressure width is carried out. Based on pedestal pressure gradient and pedestal density, six pedestal pressure width models are applied to predict the pedestal temperature height of type I ELMy H-mode. Compared to experimental results, the normalized poloidal beta model is more consistent than other models. [ABSTRACT FROM AUTHOR]

The JET tokamak is unique amongst present fusion devices in its capability to operate at high plasma current, providing the closest plasma parameters to ITER. The physics benefits of high current operation have to be balanced against the risks to the integrity of the machine due to high force disruptions. The installation of the ITER-Like Wall (ILW) has added risks due to the thermal characteristics of the metal Plasma Facing Components. This paper describes the operational aspects of the scientific development of high current H-mode plasmas with the ILW, focusing on disruption prediction, avoidance and amelioration. The development yielded baseline H-mode plasmas up to 4 MA/3.74 T, comparable to the maximum current achieved in JET in Carbon-Wall (CFC) conditions with similar divertor geometry. [ABSTRACT FROM AUTHOR]

The power fall-off width in the H-mode scrape-off layer (SOL) in tokamaks shows a strong inverse dependence on the plasma current, which was noticed by both previous multi-machine scaling work [T. Eich et al., Nucl. Fusion 53, 093031 (2013)] and more recent work [L. Wang et al., Nucl. Fusion 54, 114002 (2014)] on the Experimental Advanced Superconducting Tokamak. To understand the underlying physics, probe measurements of three H-mode discharges with different plasma currents have been studied in this work. The results suggest that a higher plasma current is accompanied by a stronger E × B shear and a shorter radial correlation length of turbulence in the SOL, thus resulting in a narrower power fall-off width. A simple model has also been applied to demonstrate the suppression effect of E × B shear on turbulence in the SOL and shows relatively good agreement with the experimental observations. [ABSTRACT FROM AUTHOR]

The general geometry of the experimental tokamak magnetic equilibrium is implemented in the global gyrokinetic simulation code GEM. Compared to the general geometry, the well used Miller parameterization of the magnetic equilibrium is a good approximation in the core region and up to the top of the pedestal. Linear simulations indicate that results with the two geometries agree for r/a≤0.9. However, in the edge region, the instabilities are sensitive to the magnetic equilibrium in both the L-mode and the H-mode plasmas. A small variation of the plasma shaping parameters leads to large changes to the edge instability. [ABSTRACT FROM AUTHOR]

Self-consistent simulations of 15 MA ITER H-mode DT scenarios, from ramp-up through flat-top, are carried out. Electron and ion temperatures, toroidal angular frequency, and currents are evolved, in simulations carried out using the predictive TRANSPort and integrated modeling code starting with initial profiles and equilibria obtained from tokamak simulation code studies. Studies are carried out examining the dependence and sensitivity of fusion power production on electron density, argon impurity concentration, choice of radio frequency heating, pedestal temperature without and with E × B flow shear effects included, and the degree of plasma rotation. The goal of these whole-device ITER simulations is to identify dependencies that might impact ITER fusion performance. [ABSTRACT FROM AUTHOR]

High-performance and long-pulse operation is a crucial goal of current magnetic fusion research. Here, we demonstrate a high-confinement plasma regime known as an H-mode with a record pulse length of over 30 s in the Experimental Advanced Superconducting Tokamak sustained by lower hybrid wave current drive (LHCD) with advanced lithium wall conditioning. We find that LHCD provides a flexible boundary control for a ubiquitous edge instability in H-mode plasmas known as an edge-localized mode, which leads to a marked reduction in the heat load on the vessel wall compared with standard edge-localized modes. LHCD also induces edge plasma ergodization that broadens the heat deposition footprint. The heat transport caused by this ergodization can be actively controlled by regulating the edge plasma conditions. This potentially offers a new means for heat-flux control, which is a key issue for next-step fusion development. [ABSTRACT FROM AUTHOR]

Plasma equilibrium is an important basis for tokamak plasma research. The equilibrium reconstructed from experimental diagnostics is a key element for experiments analysis and for theory study. The kinetic equilibrium has the profiles information (current or safety factor profile, kinetic pressure profile), which are key issues for some studies of physics. With the constraints of magnetic measurements, pressure profile and edge current profile, we achieved the first reconstructed kinetic equilibrium on EAST tokamak. The pressure and edge current profiles are based on the diagnostics and theoretical bootstrap current model. The kinetic equilibrium has the pedestal structure for H-mode plasma, which the magnetic reconstruction missed. This improved equilibrium is an important basis for some experimental analysis and theory studies on EAST. [ABSTRACT FROM AUTHOR]

The effect of gas desorption from the tokamak first wall on the pedestal recovery in the H-mode after an edge-localized-mode burst is considered. Results of FACE code simulations of hydrogen desorption from a beryllium wall are presented. It is found that the wall has a significant effect on plasma processes only at sufficiently low temperatures (of about 400 K), which agrees with qualitative estimates obtained earlier in the zero-dimensional approximation. [ABSTRACT FROM AUTHOR]

Recent research on the Alcator C-Mod tokamak has focused on a range of scientific issues with particular emphasis on ITER needs and on detailed comparisons between experimental measurements and predictive models. Research on ICRF (ion cyclotron range of frequencies) heating emphasized the origins and mitigation of metallic impurities while work on lower hybrid current drive experiments have focused on linear and nonlinear wave interactions that limit efficiency at high densities in regimes with low single pass absorption. Experiments in core turbulence and transport focused on quantitative, multi-field comparisons between nonlinear gyro-kinetics simulations and experimental measurements of profiles, fluxes and fluctuations. Experiments into self-generated rotation observed spontaneous flow reversal at a critical density identical to the transition density between linear ohmic confinement and saturated ohmic confinement regimes. H-mode studies have measured pedestal widths consistent with kinetic-ballooning-mode-like instabilities, while the pedestal heights quantitatively match the EPED code predictions. Experiments with I-mode have increased the operating window for this promising edge-localized-mode-free regime. Extrapolation of I-mode to ITER suggests that the fusion gain Q ∼ 10 could be possible in ITER. Investigations into the physics and scaling of the power exhaust channel width in attached enhanced D-alpha H-mode and L-mode plasma showed a direct connection between the midplane pressure-folding length and the outer divertor target footprint. The width was found to scale inversely with IP, while being independent of conducted power, BT or q95 and insensitive to the scrape-off layer connection length—a behaviour that suggests critical-gradient physics sets both pressure and heat-flux profiles. [ABSTRACT FROM AUTHOR]

Significant progress, on both the technological and physical fronts, towards high-performance, long-pulse plasma discharges has been made in EAST (the experimental advanced superconducting tokamak) since the last IAEA FEC. With enhanced operational capabilities, the following key results have been achieved with lithium wall conditioning: fully steady-state long-pulse diverted plasmas entirely driven by the lower hybrid current drive (LHCD) over 400 s and stationary H-mode discharges over 30 s with LHCD and ion cyclotron resonant heating (ICRF). H-modes with various types of edge localized modes (ELMs) have been achieved with HIPB98(y,2) ranging from 0.7 to over unity, providing great opportunities for the study of H-mode physics. New and exciting physics with dominant radio-frequency heating has emerged, including new findings of LHCD-induced 3D edge magnetic topology and its effect on ELM dynamics and particle and heat deposition; the role of zonal flows during the L–H transition and I-phase; and a new turbulence-flow cycle state at the H-mode pedestal. Various means for mitigating ELMs have also been demonstrated to facilitate long-pulse operation, including supersonic molecular beam injection, as well as innovative solid Li granule injection. A brief overview of these recent advances is presented. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Following the completion in May 2011 of the shutdown for the installation of the beryllium wall and the tungsten divertor, the first set of JET campaigns have addressed the investigation of the retention properties and the development of operational scenarios with the new plasma-facing materials. The large reduction in the carbon content (more than a factor ten) led to a much lower Zeff (1.2–1.4) during L- and H-mode plasmas, and radiation during the burn-through phase of the plasma initiation with the consequence that breakdown failures are almost absent. Gas balance experiments have shown that the fuel retention rate with the new wall is substantially reduced with respect to the C wall. The re-establishment of the baseline H-mode and hybrid scenarios compatible with the new wall has required an optimization of the control of metallic impurity sources and heat loads. Stable type-I ELMy H-mode regimes with H98,y2 close to 1 and βN ∼ 1.6 have been achieved using gas injection. ELM frequency is a key factor for the control of the metallic impurity accumulation. Pedestal temperatures tend to be lower with the new wall, leading to reduced confinement, but nitrogen seeding restores high pedestal temperatures and confinement. Compared with the carbon wall, major disruptions with the new wall show a lower radiated power and a slower current quench. The higher heat loads on Be wall plasma-facing components due to lower radiation made the routine use of massive gas injection for disruption mitigation essential. [ABSTRACT FROM AUTHOR]

Through a diverse research programme, the Tokamak à Configuration Variable (TCV) addresses physics issues and develops tools for ITER and for the longer term goals of nuclear fusion, relying especially on its extreme plasma shaping and electron cyclotron resonance heating (ECRH) launching flexibility and preparing for an ECRH and NBI power upgrade. Localized edge heating was unexpectedly found to decrease the period and relative energy loss of edge localized modes (ELMs). Successful ELM pacing has been demonstrated by following individual ELM detection with an ECRH power cut before turning the power back up to trigger the next ELM, the duration of the cut determining the ELM period. Negative triangularity was also seen to reduce the ELM energy release. H-mode studies have focused on the L–H threshold dependence on the main ion species and on the divertor leg length. Both L- and H-modes have been explored in the snowflake configuration with emphasis on edge measurements, revealing that the heat flux to the strike points on the secondary separatrix increases as the X-points approach each other, well before they coalesce. In L-mode, a systematic scan of the auxiliary power deposition profile, with no effect on confinement, has ruled it out as the cause of confinement degradation. An ECRH power absorption observer based on transmitted stray radiation was validated for eventual polarization control. A new profile control methodology was introduced, relying on real-time modelling to supplement diagnostic information; the RAPTOR current transport code in particular has been employed for joint control of the internal inductance and central temperature. An internal inductance controller using the ohmic transformer has also been demonstrated. Fundamental investigations of neoclassical tearing mode (NTM) seed island formation by sawtooth crashes and of NTM destabilization in the absence of a sawtooth trigger were carried out. Both stabilizing and destabilizing agents (electron cyclotron current drive on or inside the q = 1 surface, respectively) were used to pace sawtooth oscillations, permitting precise control of their period. Locking of the sawtooth period to a pre-defined ECRH modulation period was also demonstrated. Sawtooth control has permitted nearly failsafe NTM prevention when combined with backup NTM stabilization by ECRH. [ABSTRACT FROM AUTHOR]

Abstract: In the vast majority of tokamaks, the high confinement mode (H mode) has been systematically achieved and studied. The transition to the H-mode is a very complex phenomenon but it can be represented by the relation between a limited number of macroscopic quantities. The predicted power (P Thresh) for accessing the H-mode is generally estimated by statistical analysis of multi-machine experimental data and typically presented in terms of equations in power law monomial form. Even if the positive effects of the plasma shape on the confinement time have been clearly documented, the dependence of the P Thresh on elongation and triangularity is still an aspect no completely clarified. The objective of this paper is to assess whether the P Thresh dependency on the shape of the plasma can be derived by a statistical approach using a recent version of ITPA International Global Threshold Data Base (IGDBTHv6b). A statistical method to select relevant macroscopic quantities in the candidate models for the scaling of P Thresh is presented and a multi-machine scaling law, using non-linear regression techniques, is derived. [Copyright &y& Elsevier]

Global gyrokinetic simulations of DIII-D H-mode edge pedestal show two types of instabilities may exist approaching the onset of edge localized modes: an intermediate-n, high frequency mode which we identify as the 'kinetic peeling ballooning mode (KPBM),' and a high-n, low frequency mode. Our previous study [W. Wan et al., Phys. Rev. Lett. 109, 185004 (2012)] has shown that when the safety factor profile is flattened around the steep pressure gradient region, the high-n mode is clearly kinetic ballooning mode and becomes the dominant instability. Otherwise, the KPBM dominates. Here, the properties of the two instabilities are studied by varying the density and temperature profiles. It is found that the KPBM is destabilized by density and ion temperature gradient, and the high-n mode is mostly destabilized by electron temperature gradient. Nonlinear simulations with the KPBM saturate at high levels. The equilibrium radial electric field (Er) reduces the transport. The effect of the parallel equilibrium current is found to be weak. [ABSTRACT FROM AUTHOR]

The application of resonant magnetic perturbations (RMPs) with a toroidal mode number of n = 4 or n = 6 to lower single null plasmas in the MAST tokamak produces up to a factor of 5 increase in edge-localized mode (ELM) frequency and reduction in plasma energy loss associated with type-I ELMs. A threshold current for ELM mitigation is observed above which the ELM frequency increases approximately linearly with current in the coils. Despite a large scan of parameters, complete ELM suppression has not been achieved. The results have been compared with modelling performed using either the vacuum approximation or including the plasma response. During the ELM mitigated stage clear lobe structures are observed in visible-light imaging of the X-point region. The size of these lobes is correlated with the increase in ELM frequency observed. The characteristics of the mitigated ELMs are similar to those of the natural ELMs suggesting that they are type-I ELMs which are triggered at a lower pressure gradient. The application of the RMPs in the n = 4 and n = 6 configurations before the L-H transition has little effect on the power required to achieve H-mode while still allowing the first ELM to be mitigated. [ABSTRACT FROM AUTHOR]

This paper provides an investigation of the production/transport properties of W impurity in the edge of experimental advanced superconducting tokamak (EAST) H-mode discharges with upper-single-null configurations by using DIVIMP Monte-Carlo code. The background plasmas are provided by SOLPS5.0 calculations. Firstly, to address the detailed dependence of W impurity behaviors on plasma conditions in the scraped-off layer (SOL)/divertor region of EAST, two pre-edge localized mode (ELM) cases with divertor plasmas in high-recycling and partially-detached regimes, respectively, are considered and compared. It is found that, due to the competitions between the thermal and friction forces (FFs), in the high-recycling case, large quantities of W impurities are transported to the upstream; while in the partially-detached case, most of the W impurities are located near the inner and outer divertors. Furthermore, the W core contamination in type-I ELMy H-mode plasmas of EAST has been estimated by the DIVIMP-SOLPS5.0 simulations, without the consideration of the release of W impurity from the core due to ELMs. During the ELM, simulations exhibit the low-recycling regime of divertor plasma, and high (several hundred eV) target plasma temperatures together with low SOL collisionality ( ∼ 1–3), lead to an order of magnitude increase in the W core contamination rate. However, in the ELM-recovering phase, with the decrease (increase) of divertor plasma temperature (density), significant increase of divertor retention is obtained due to the remarkable increase of FFs on W ions. Besides, intra-ELM phase approximately contributes to more than 50% of the total W core contamination per an ELM cycle. This work represents a step toward a deeper understanding of the W impurity production/transport in EAST H-mode discharges. [ABSTRACT FROM AUTHOR]

Using a high-speed and broadband radio frequency (RF) (0.1–1 GHz) spectrum analyzer developed on the KSTAR tokamak, it is found that several distinct stages of RF emission appear at the pedestal collapse in high confinement discharges. Comparison with 2D electron cyclotron emission (ECE) images has revealed that each stage is related to the instantaneous condition at the outboard mid-plane edge. First, high-harmonic ion cyclotron emissions (ICE) are intensified with the appearance of a non-modal filamentary perturbation in the edge within several tens of microseconds before the collapse. Then, the RF emission becomes broad toward high-frequency range (<500 MHz) at the burst onset of the non-modal filament. During the pedestal collapse initiated by the filament burst, rapid chirping (1–3 μs) appear with additional filament bursts. The strong correlation between the RF spectra and the perturbation structure provides important clues on the stability of edge-localized modes and on the ion dynamics in the plasma boundary. [ABSTRACT FROM AUTHOR]

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power –5 MW and a range of plasma currents MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10–15% of total outer divertor power. Measurements of electron pressure and poloidal beta support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4–4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard divertor was found. The results complement the initial SF divertor studies conducted in high-power H-mode discharges in the NSTX and DIII-D tokamaks, and, along with snowflake divertor results from TCV and other tokamaks, contribute to the physics basis of the SF divertor as a power exhaust concept for future high power density tokamaks. [ABSTRACT FROM AUTHOR]

We show for JET and ASDEX Upgrade, based on Thomson-scattering measurements, a clear correlation of the density limit of the tokamak H-mode high-confinement regime with the approach to the ideal ballooning instability threshold at the periphery of the plasma. It is shown that the MHD ballooning parameter at the separatrix position increases about linearly with the separatrix density normalized to Greenwald density, for a wide range of discharge parameters in both devices. The observed operational space is found to reach at maximum –0.5 at values for –2.5, in the range of theoretical predictions for ballooning instability. This work supports the hypothesis that the H-mode density limit may be set by ballooning stability at the separatrix. [ABSTRACT FROM AUTHOR]

Experiments using nitrogen were conducted in H-mode plasmas at ASDEX Upgrade that has a full-W wall. The edge region of H-mode plasmas is modulated by the edge-localized modes (ELMs) which lead to a loss of energy and particles from the confined plasma. In order to gain a better understanding of the complex physical mechanisms which govern the behaviour of radiation and impurities in the presence of ELMs, the evolution of impurities and radiation has been modelled in a time-dependent way. The simulations have been carried out with the ASTRA-STRAHL package featuring the self-consistent interplay between impurity transport, radiation, heat and particle transport of the background plasma, and the effects of ELMs. ELMs are modelled based on the two different assumptions of a diffusive and a convective transport, respectively. The experimental discharge behaviour was reproduced providing only transport coefficients, heat, and particle source. The results underlie the importance of non-coronal effects through the ELM-induced transport which lead to a strong enhancement of the nitrogen radiation. Taking these effects into account is crucial in order to not underestimate the radiation. The radiation properties of high-Z impurities such as tungsten are instead very weakly influenced by non-coronal effects due to its faster equilibration time. While the nitrogen density does not change significantly decreasing the ELM frequency, tungsten density and consequently the radiation increase strongly. The degree to which W is flushed out depends on whether the ELM transport is diffusive or convective. Simulations show that for the N seeded cases considered here, the diffusive model reproduces more accurately the experimental observations. The different behaviour of N and W can be explained in terms of profile peaking which increases with Z (neoclassical pinch). The strong increase in W radiation when the ELM frequency is decreased is not only due to the lack of a sufficiently strong flush out of this impurity but also to the fact that the long time between two crashes gives the impurities time to penetrate further into the plasma escaping the region where they can be flushed out. This is in agreement with the experimental observations and highlights the importance of maintaining high ELM frequencies for the stability and performance of the discharges. [ABSTRACT FROM AUTHOR]

Reference scenarios for the JT-60SA tokamak have been simulated with one-dimensional transport codes to assess the stationary state of the flat-top phase and provide a profile database for further physics studies (e.g. MHD stability, gyrokinetic analysis) and diagnostics design. The types of scenario considered vary from pulsed standard H-mode to advanced non-inductive steady-state plasmas. In this paper we present the results obtained with the ASTRA, CRONOS, JINTRAC and TOPICS codes equipped with the Bohm/gyro-Bohm, CDBM and GLF23 transport models. The scenarios analysed here are: a standard ELMy H-mode, a hybrid scenario and a non-inductive steady state plasma, with operational parameters from the JT-60SA research plan. Several simulations of the scenarios under consideration have been performed with the above mentioned codes and transport models. The results from the different codes are in broad agreement and the main plasma parameters generally agree well with the zero dimensional estimates reported previously. The sensitivity of the results to different transport models and, in some cases, to the ELM/pedestal model has been investigated. [ABSTRACT FROM AUTHOR]

We perform a gyrokinetic study of 3D resonant magnetic perturbation (RMP) field effect on zonal flows in tokamak plasmas. The result indicates that energy-dependent secular radial drift of particles due to the tangential component () of the RMP field induces a further long term algebraic decay of the Rosenbluth–Hinton (1998 Phys. Rev. Lett. 80 724) residual flow expected in an axisymmetric tokamak. This mechanism could be related to an increase of H-mode transition power threshold in the presence of RMP observed in many tokamak experiments. [ABSTRACT FROM AUTHOR]

One of key objectives of tokamak experiments is the exploration of enhanced confinement regimes, and the access of the internal transport barrier (ITB) formation is dealt with an important physics issue in the most of major tokamaks. Also, the advanced tokamak scenario with ITB is expected to lead to a continuous reactor with high fusion power density. From that point of view, the formation of the ITB in KSTAR which is designed for long pulse operation capability is very important although its heating and current drive systems are not fully equipped yet. We have therefore assumed that an early injection of the full NBI power (∼5.5 MW) during the current ramp-up would give a chance to form an internal barrier if the plasma could stay in the L-mode. To avoid the H-mode transition, we have produced inboard limited plasmas with detaching from the both upper and lower divertors. Using this approach, an ITB formation during L-mode has been observed which shows improved core confinement. Ion and electron temperature profiles show the barrier clearly in the temperature, and it was sustained for about 7 s in the dedicated experiment. This is the first stationary ITB observed in a full superconducting tokamak. This operation scenario with the ITB could be an alternative way to achieve a high performance regime in KSTAR, and the length of the ITB discharge could be extended even longer. In this paper, we present the formation of the ITB using measured and simulated characteristic profiles. [ABSTRACT FROM AUTHOR]

The first high fidelity gyrokinetic simulations of the energy losses in the transport barriers of large tokamaks in pursuit of fusion gain are presented. These simulations calculate the turbulent energy losses with an extensive treatment of relevant physical effects—fully kinetic, non-linear, electromagnetic—inclusive of all major plasma species, and in equilibria with relevant shape and local bootstrap current for fusion-relevant cases. We find that large plasmas with a small normalized gyroradius lie in an unexpected regime of enhanced losses that can prevent the projected energy gain. Our simulations are qualitatively consistent with recent experiments on JET with an ITER-like wall. Interestingly and very importantly, the simulations predict parameter regimes of reduced transport that are quite fusion-favorable. [ABSTRACT FROM AUTHOR]

We report the latest results on fast measurements of the electron temperature and parallel heat flux in the COMPASS tokamak scrape-off layer (SOL) and divertor region during ELMy H-mode plasmas. The system of ball-pen and Langmuir probes installed on the divertor target, the horizontal reciprocating manipulator and the fast data-acquisition system with sampling frequency rate f  =  5 MSa s−1 allow us to measure the electron temperature and parallel heat flux during inter-ELM and ELM periods with high temporal resolution. The filamentary structure of the electron temperature and parallel heat flux was observed during ELMs in the SOL as well as in the divertor region. The position of the filaments within ELMs is not regular and therefore the resulting conditionally averaged ELM neglects the peak values of the electron temperature and parallel heat flux. We have found a substantial difference between the value of the radial power decay length in the inter-ELM period λq,inter  =  2.5 mm and the decay length of the peak ELM heat flux λq,ELM  =  13.1 mm. The decay length of the ELM energy density was found to be λE,ELM  =  5.4 mm. [ABSTRACT FROM AUTHOR]

Among various edge localized mode (ELM) crash control methods, only non-axisymmetric magnetic perturbations (NAMPs) yield complete suppression of ELM crashes beyond their mitigation, and thus attract more attention than others. No other devices except KSTAR, DIII-D, and recently EAST have successfully achieved complete suppression with NAMPs. The underlying physics mechanisms of these successful ELM crash suppressions in a non-axisymmetric field environment, however, still remain uncertain. In this work, we investigate the ELM crash suppression characteristics of the KSTAR ELMy H-mode discharges in a controlled multi-spectral field environment, created by both middle reference and top/bottom proxy in-vessel control coils. Interestingly, the attempts have produced a set of contradictory findings, one expected (ELM crash suppression enhancement with the addition of n  =  1 to the n  =  2 field at relatively low heating discharges) and another unexpected (ELM crash suppression degradation at relatively high heating discharges) from the earlier findings in DIII-D. This contradiction indicates the dependence of the ELM crash suppression characteristics on the heating level and the associated kink-like plasma responses. Preliminary linear resistive MHD plasma response simulation shows the unexpected suppression performance degradation to be likely caused by the dominance of kink-like plasma responses over the island gap-filling effects. [ABSTRACT FROM AUTHOR]

Experiments have been performed under the coordination of the International Tokamak Physics Activity (ITPA) on several tokamaks, including ASDEX Upgrade (AUG), JET and DIII-D, to characterize the increased Ion cyclotron range of frequency (ICRF) antenna loading achieved by optimizing the position of gas injection relative to the RF antennas. On DIII-D, AUG and JET (with the ITER-Like Wall) a 50% increase in the antenna loading was observed when injecting deuterium in ELMy H-mode plasmas using mid-plane inlets close to the powered antennas instead of divertor injection and, with smaller improvement when using gas inlets located at the top of the machine. The gas injection rate required for such improvements (~0.7  ×  1022 el s−1 in AUG, ~1.0  ×  1022 el s−1 in JET) is compatible with the use of this technique to optimize ICRF heating during the development of plasma scenarios and no degradation of confinement was observed when using the mid-plane or top inlets compared with divertor valves. An increase in the scrape-off layer (SOL) density was measured when switching gas injection from divertor to outer mid-plane or top. On JET and DIII-D, the measured SOL density increase when using main chamber puffing is consistent with the antenna coupling resistance increase provided that the distance between the measurement lines of sight and the injection location is taken into account. Optimized gas injection was also found to be beneficial for reducing tungsten (W) sputtering at the AUG antenna limiters, and also to reduce slightly the W and nickel (Ni) content in JET plasmas. Modeling the specific effects of divertor/top/mid-plane injection on the outer mid-plane density was carried out using both the EDGE2D-EIRENE and EMC3-EIRENE plasma boundary code packages; simulations indeed indicate that outer mid-plane gas injection maximizes the density in the mid-plane close to the injection point with qualitative agreement with the AUG SOL density measurements for EMC3-EIRENE. Field line tracing for ITER in the 15 MA QDT  =  10 reference scenario indicates that the planned gas injection system could be used to tailor the density in front the antennas. Benchmarking of EMC3-EIRENE against AUG and JET data is planned as a first step towards the ITER SOL modelling required to quantify the effect of gas injection on the SOL density in front of the antennas. [ABSTRACT FROM AUTHOR]

It is crucial to increase the total stored energy by realizing the transition from a low confinement (L-mode) state to a high confinement (H-mode) state in magnetic confinement fusion. The L–H transition process is simulated by using the predictive transport code based on Weiland’s fluid model. Based on the equilibrium parameters obtained from equilibrium fitting (EFIT) in the experiment, the electron density ne, electron temperature Te, ion temperatures Ti, ion poloidal Vp, and toroidal momenta Vt are simulated self-consistently. The L–H transition dynamic behaviors with the formation of the transport barriers of ion and electron temperatures, the electron density, and the ion toroidal momenta are analyzed. During the L–H transition, the strong poloidal flow shear in the edge transport barrier region is observed. The crashes of the electron and ion temperature pedestals are also observed during the L–H transition. The effects of the heating and particle sources on the L–H transition process are studied systematically, and the critical power threshold of the L–H transition is also found. [ABSTRACT FROM AUTHOR]

This paper describes the range of variations in edge and SOL turbulence observed using a gas puff imaging (GPI) diagnostic in NSTX discharges. The database consists of 140 shots including Ohmic, L-mode, and H-mode plasmas measured during steady-state conditions (e.g. without ELMs). Turbulence quantities were evaluated using both cross-correlation analysis and blob tracking. Relative fluctuation levels varied from –1.0, correlation times were –40 μs, correlation lengths were –10 cm, and turbulence velocities were km s−1 and km s−1 outwards. These variations were evaluated with respect to both the global and local edge plasma parameters, and compared with simplified theoretical models. [ABSTRACT FROM AUTHOR]

The I-mode regime, routinely observed on the Alcator C-Mod tokamak, is characterized by an edge energy transport barrier without an accompanying particle barrier and with broadband instabilities, known as weakly coherent modes (WCM), believed to regulate particle transport at the edge.Recent experiments on the DIII-D tokamak exhibit I-mode characteristics in various physical quantities.These DIII-D plasmas evolve over long periods, lasting several energy confinement times, during which the edge electron temperature slowly evolves towards an H-mode-like profile, while maintaining a typical L-mode edge density profile. During these periods, referred to as I-mode phases, the radial electric field at the edge also gradually reaches values typically observed in H-mode.Density fluctuations measured with the phase contrast imaging diagnostic during I-mode phases exhibit three features typically observed in H-mode on DIII-D, although they develop progressively with time and without a sharp transition: the intensity of the fluctuations is reduced; the frequency spectrum is broadened and becomes non-monotonic; two dimensional space-time spectra appear to approach those in H-mode, showing phase velocities of density fluctuations at the edge increasing to about 10 km s−1. However, in DIII-D there is no clear evidence of the WCM.Preliminary linear gyro-kinetic simulations are performed in the pedestal region with the GS2 code and its recently upgraded model collision operator that conserves particles, energy and momentum. The increased bootstrap current and flow shear generated by the temperature pedestal are shown to decrease growth rates, thus possibly generating a feedback mechanism that progressively stabilizes fluctuations. [ABSTRACT FROM AUTHOR]

Gyrokinetic simulations have been used to identify electromagnetic microinstabilities in the H-mode pedestal region of DIII-D shot 131 997 using global gyrokinetic code GTC. It was found that dominant instability at the top of the pedestal is the ion temperature gradient mode (ITG). In the maximum gradient location the most unstable mode is the kinetic ballooning mode (KBM) for the dominant poloidal wavenumber cm−1. For shorter wavelengths the dominant instability is the trapped-electron mode (TEM). We have demonstrated the ITG–KBM transition at the pedestal top, and TEM–KBM transition in the steep pressure gradient region as plasma pressure increases while gradients remain unchanged. [ABSTRACT FROM AUTHOR]

A new ‘Super H-mode’ regime is predicted, which enables pedestal height and predicted fusion performance substantially higher than for H-mode operation. This new regime is predicted to exist by the EPED pedestal model, which calculates criticality constraints for peeling–ballooning and kinetic ballooning modes, and combines them to predict the pedestal height and width. EPED usually predicts a single (‘H-mode’) pedestal solution for each set of input parameters, however, in strongly shaped plasmas above a critical density, multiple pedestal solutions are found, including the standard ‘H-mode’ solution, and a ‘Super H-Mode’ solution at substantially larger pedestal height and width. The Super H-mode regime is predicted to be accessible by controlling the trajectory of the density, and to increase fusion performance for ITER, as well as for DEMO designs with strong shaping. A set of experiments on DIII-D has identified the predicted Super H-mode regime, and finds pedestal height and width, and their variation with density, in good agreement with theoretical predictions from the EPED model. The very high pedestal enables operation at high global beta and high confinement, including the highest normalized beta achieved on DIII-D with a quiescent edge. [ABSTRACT FROM AUTHOR]

The solitary radial electric field in the edge of toroidal plasma is studied based on the electric field bifurcation model. Results are applied to tokamak and helical plasmas, and the dependence of the electric field structure on the plasma parameters and geometrical factors is analyzed. The order of magnitude estimate for tokamak plasma is not far from experimental observations. It is shown that, in helical plasmas, the height of electric field structure is reduced substantially owing to the ripple particle transport, while the width is influenced less. The implications of the results for the limit of achievable gradient in the H-mode pedestal are also discussed. [ABSTRACT FROM AUTHOR]

This article presents a radial electric field measurement by a heavy ion beam probe in the JFT-2M tokamak, during the L-H transition. An abrupt increase (time scale of O(100 µs)) of the strong edge radial electric field (localized in the radius with FWHM ∼7 mm) results in the increase of density gradient and turbulence reduction. Rapid inward propagation of the turbulence suppression front is observed at the transition. After the transition, the electric field structure in the tiny edge localized modes (ELMs) is analyzed. Transport self-regulation events observed in the vicinity of the L-H transition, i.e. the limit cycle oscillation (LCO) in the L-mode, the tiny ELM in the H-mode, as well as the L-H transition itself, are summarized in a single Lissajous diagram in the electric field-density gradient space, which provides a comprehensive explanation of the transition dynamics. [ABSTRACT FROM AUTHOR]

Two enhanced-recycling H-mode regimes, named low-enhanced-recycling (LER) and high-enhanced-recycling (HER) H-mode regimes, with edge coherent modes, have been achieved by lower hybrid current drive and ion cyclotron resonance heating with lithium-wall conditioning in the EAST superconducting tokamak. In the LER H-mode regime, the density and radiation increase during the ELM-free phase until the onset of edge-localized modes (ELMs), while in the HER H-mode regime, the density and radiation are well controlled without the presence of ELMs. Both LER and HER H-modes exhibit a low-frequency (frequency <100 kHz) edge quasi-coherent mode (ECM) with an initial frequency chirping down phase, following the L–H transition. In addition, an electromagnetic high-frequency coherent mode (HFM) with frequency >170 kHz appears shortly (<1 ms) after the transition during HER H-modes. Both ECM and HFM propagate in the electron diamagnetic drift direction in the lab frame with a low poloidal wavelength and may be responsible for enhanced recycling during the ELM-free phase. These two enhanced-recycling H-mode regimes may have significant implications for long-pulse high-performance operations in future fusion experiments. [ABSTRACT FROM AUTHOR]

The core micro-instability characteristics of hybrid and baseline plasmas in a selected set of JET plasmas with carbon wall are investigated through local linear and non-linear and global linear gyro-kinetic simulations with the GYRO code (Candy and Belli 2011 General Atomics Report GA-A26818). In particular, we study the role of plasma pressure on the micro-instabilities, and scan the parameter space for the important plasma parameters responsible for the onset and stabilization of the modes under experimental conditions. We find that a good core confinement due to strong stabilization of the micro-turbulence driven transport can be expected in the hybrid plasmas due to the stabilizing effect of the fast ion pressure that is more effective at the low magnetic shear of the hybrid discharges. While parallel velocity gradient destabilization is important for the inner core, at outer radii the hybrid plasmas may benefit from a strong quench of the turbulence transport by E × B rotation shear. [ABSTRACT FROM AUTHOR]

Blob properties are studied in the scrape-off layer of the tokamak ASDEX Upgrade with a fast camera. The gas-puff imaging technique is used to investigate the detection rate as well as the blob size and velocity scaling. The experiments were performed in L- and H-mode phases of the same discharges to study the change in blob properties after the L-H transition. In both regimes the detection rate is of the order of a few thousand blobs per second, which is compatible with the picture of blob generation by edge micro instabilities. The blob size increases in H-mode, while the radial velocity decreases slightly. The changes are, however, not indicating a drastic change in the blob dynamics in both phases. The experimentally found blob properties were compared to predictions from a novel blob model including effects due to a finite ion temperature, which should be more appropriate for the conditions in the SOL of fusion plasmas. [ABSTRACT FROM AUTHOR]

Broadband swept (i.e. frequency modulation of the continuous wave; FM-CW) and fixed frequency reflectometry (FFR) were used for the first time to study plasma filamentary activity; experiments were performed in ELMy H-mode plasmas at the ASDEX Upgrade tokamak. Electron density profiles were studied with FM-CW providing a first insight into filamentary activity and enabling us to localize the density layers probed with FFR. A novel filament detection technique was developed using a threshold criterion on the phase derivative signals from FFR. This technique was applied together with conditional averaging in measurements performed in the vicinity of the separatrix. Results showed good agreement with data from Langmuir probes and it was found that the majority of filaments propagate with dominant poloidal velocity in both periods of in-between edge localized modes (ELMs) (Vθ≈ 575 m s−1) and at the ELM onset (Vθ≈ 1180 m s−1). A time delay between the maximum filament activity at the outer mid-plane and the ELM peak at the inner divertor currents (≈ −461  ±  50 μs) agrees with expected time scales for the ELM lifetime. In inter-ELM periods we were able to estimate typical poloidal and toroidal sizes of filaments (Sθ≈ [5.75–11.50] cm and Sϕ≈ [33–66] cm) and a magnetohydrodynamic mode structure emerged from the measurements with poloidal and toroidal mode numbers (m ≈ [8–12] and n ≈ [2, 3]) in the range of possible peeling-ballooning modes. [ABSTRACT FROM AUTHOR]

We evaluate the characteristics of global energy confinement in KSTAR (τE,KSTAR) quantitatively in three ways; firstly by comparing it with multi-machine scalings, secondly by deriving multiple regression equations for the L- and the H-mode plasmas, respectively, and lastly by comparing confinement enhancement of the H-mode phase with respect to the L-mode phase in each discharge defined as Hexp. The KSTAR database exhibits τE,KSTAR of ∼0.04 to ∼0.16 s and of ∼0.06 to ∼0.19 s in L-mode and in H-mode plasmas, respectively. The multiple regression equations derived by statistical analysis present the similar dependency on PL and higher dependency on Ip compared with the multi-machine scalings, however the dependency on κ in both L- and H-mode plasmas draw the negative power dependency of κ−0.68 and κ−0.76 for H-mode and for L-mode database, respectively on the contrary to the positive dependency in all multi-machine empirical scalings. It is found that the energy confinement of both L-mode and H-mode of the discharges with Hexp > 1.5 can be well-predicted by multi-machine scalings, τE,89L and τE,92H. Apart from this, the H-mode confinement with 1.5 < Hexp < 2.0 is well-predicted by using the multi-machine empirical L-mode scaling τE,89L. [ABSTRACT FROM AUTHOR]