Search

DIII-D research is addressing critical challenges in preparation for ITER and the next generation of fusion devices through focusing on plasma physics fundamentals that underpin key fusion goals, understanding the interaction of disparate core and boundary plasma physics, and developing integrated scenarios for achieving high performance fusion regimes. Fundamental investigations into fusion energy science find that anomalous dissipation of runaway electrons (RE) that arise following a disruption is likely due to interactions with RE-driven kinetic instabilities, some of which have been directly observed, opening a new avenue for RE energy dissipation using naturally excited waves. Dimensionless parameter scaling of intrinsic rotation and gyrokinetic simulations give a predicted ITER rotation profile with significant turbulence stabilization. Coherence imaging spectroscopy confirms near sonic flow throughout the divertor towards the target, which may account for the convection-dominated parallel heat flux. Core-boundary integration studies show that the small angle slot divertor achieves detachment at lower density and extends plasma cooling across the divertor target plate, which is essential for controlling heat flux and erosion. The Super H-mode regime has been extended to high plasma current (2.0 MA) and density to achieve very high pedestal pressures (~30 kPa) and stored energy (3.2 MJ) with H 98y2 ≈ 1.6–2.4. In scenario work, the ITER baseline Q = 10 scenario with zero injected torque is found to have a fusion gain metric independent of current between q 95 = 2.8–3.7, and a lower limit of pedestal rotation for RMP ELM suppression has been found. In the wide pedestal QH-mode regime that exhibits improved performance and no ELMs, the start-up counter torque has been eliminated so that the entire discharge uses ≈0 injected torque and the operating space is more ITER-relevant. Finally, the high- (⩽3.8) hybrid scenario has been extended to the high-density levels necessary for radiating divertor operation, achieving ~40% divertor heat flux reduction using either argon or neon with P tot up to 15 MW.

This paper describes joint ITPA studies of the I-mode regime, which features an edge thermal barrier together with L-mode-like particle and impurity transport and no edge localized modes (ELMs). The regime has been demonstrated on the Alcator C-Mod, ASDEX Upgrade and DIII-D tokamaks, over a wide range of device parameters and pedestal conditions. Dimensionless parameters at the pedestal show overlap across devices and extend to low collisionality. When they are matched, pedestal temperature profiles are also similar. Pedestals are stable to peeling–ballooning modes, consistent with lack of ELMs. Access to I-mode is independent of heating method (neutral beam injection, ion cyclotron and/or electron cyclotron resonance heating). Normalized energy confinement H 98,y2 ≥ 1 has been achieved for a range of 3 ≤ q 95 ≤ 4.9 and scales favourably with power. Changes in turbulence in the pedestal region accompany the transition from L-mode to I-mode. The L–I threshold increases with plasma density and current, and with device size, but has a weak dependence on toroidal magnetic field B T. The upper limit of power for I-modes, which is set by I–H transitions, increases with B T and the power range is largest on Alcator C-Mod at B > 5 T. Issues for extrapolation to ITER and other future fusion devices are discussed., United States. Department of Energy (DE-FC02-99ER54512-CMOD), United States. Department of Energy (DE-SC0012469), United States. Department of Energy (DE-FC02-04ER54698), United States. Department of Energy (DE-FG02-94ER54235), United States. Department of Energy (DE-AC52-07NA27344), United States. Department of Energy (DE-AC02-09CH11466), United States. Department of Energy (DE-FG02-89ER53296), United States. Department of Energy (DE-FG02-08ER54999), United States. Department of Energy (DE-FG02-08ER54984)

As shown on DIII-D edge localized modes (ELMs) can be either completely eliminated or mitigated with resonant magnetic perturbation (RMP) fields. Two infrared cameras, separated 105° toroidally, were used to make simultaneous measurements of ELM heat loads with high frame rates. Without the RMP fields ELMs display a variety of different heat load dynamics and a range of toroidal variability that is characteristic of their 3D structure. Comparing radial averages there is no asymmetry between two toroidal locations. With RMP-mitigated ELMs, the variability in the radially averaged power loads is significantly reduced and toroidal asymmetries in power loads are introduced. In addition to RMP ELM suppression scenarios an RMP scenario with only very small ELMs and very good confinement has been achieved.

Stochastic boundaries in fusion devices have been investigated in tokamaks, stellarators and reversed field pinch experiments for many years. However, since edge localized modes (ELMs) have been successfully eliminated in H-mode plasmas at the DIII-D tokamak [1,2] with small, edge resonant magnetic perturbations, they have become a widely investigated topic in tokamaks. In DIII-D stochastic boundaries are produced by coils external to the plasma. The magnetic field there consists of field lines with very different connection lengths, which produces a three dimensional, heterogenous structure of stochastic volume. The most obvious manifestation of the perturbed plasma edge is the strike line splitting observable in heat and particle fluxes, which changes with collisionality. The interaction of the magnetic perturbation and the magnetic equilibrium is of a resonant nature and the structure of the stochastic volume is a strong function of q(95). This is observed as a modulation of e.g. electron temperature as measured by ECE or Thomson scattering. In this work we summarize recent experimental findings on properties of the stochastic boundary in DIII-D. (C) 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim

Large Type-I edge localized modes (ELMs) were suppressed by n = 3 resonant magnetic perturbations (RMPs) from a set of internal coils in plasmas with an ITER similar shape at the ITER pedestal collisionality, ν e ∗ ∼ 0.1 and low edge safety factor (q95 ≈ 3.6), with either a single toroidal row of the internal RMP coils or two poloidally separated rows of coils. ELM suppression with a single row of internal coils was achieved at approximately the same q95 surface-averaged perturbation field as with two rows of coils, but required higher current per coil. Maintaining complete suppression of ELMs using n = 3 RMPs from a single toroidal row of internal coils was less robust to variations in input neutral beam injection torque than previous ELM suppression cases using both rows of internal coils. With either configuration of RMP coils, maximum ELM size is correlated with the width of the edge region having good overlap of the magnetic islands from vacuum field calculations.

Density fluctuation measurements covering a wavenumber range relevant to ITG, TEM and ETG type instabilities (k⊥ ρi ≈ 0.2→ 10) are compared during Ohmic and neutral beam injection (NBI). These short time duration NBI perturbations modified the background temperature as well as the density fluctuation behaviour but had little effect on the background density. It is found that both the measured and calculated responses to these perturbations varied with wavenumber supporting the need for broad wavenumber comparisons. The experimentally measured fluctuation changes and calculated growth rates (from the linear gyrokinetic code GKS) show qualitative similarities to each other at low and intermediate k while the comparison at high k is inconclusive.

Active feedback control for regulation of the safety factor (q) profile at the start of the high stored energy phase of an advanced tokamak discharge has been demonstrated in the DIII-D tokamak. The time evolution of the on-axis or minimum value of q is controlled during and just following the period of ramp-up of the plasma current using electron heating to modify the rate of relaxation of the current profile. In L-mode and H-mode discharges, feedback control of q is effective with the appropriate choice of either off-axis electron cyclotron heating or neutral beam heating as the actuator. The q profile is calculated in real time from a complete equilibrium reconstruction fitted to external magnetic field and flux measurements and internal poloidal field measurements from the motional Stark effect diagnostic.

Active control of plasma profiles is an essential requirement for operating within plasma stability limits, for steady-state operation and for optimization of the plasma performance. In DIII-D, plasma profiles have been actively controlled using various actuators in the following manner: (a) real time closed loop control of the q profile evolution using electron cyclotron heating and neutral beam injection as actuators; (b) active control of the density and pressure profiles in quiescent H-mode and quiescent double barrier plasmas using electron cyclotron current drive (ECCD) and pellet injection; (c) active control of the edge profiles to suppress edge localized modes using resonant magnetic perturbation with toroidal mode number n = 3, (d) real time control of the current density profile to suppress neoclassical tearing modes using localized deposition of co-ECCD.

Characteristics and scaling properties of the geodesic acoustic mode (GAM), a coherent, radially-sheared high frequency (~15?kHz) zonal flow oscillation, are studied systematically using time-delay-estimation techniques applied to localized, multi-point density fluctuation measurements obtained by beam emission spectroscopy on DIII-D. The GAM amplitude is shown to increase strongly with increasing safety factor, q95, and to likewise become undetectably small for q95 < 4.2, qualitatively consistent with theoretical predictions based on collisional damping as well as simulations. The radial structure of the GAM exhibits peak amplitude in the radial range 0.88 < r/a < 0.95 with a rapid amplitude reduction inside and outside this region. The measured frequency is close to the predicted frequency, though some deviation to higher frequency is observed at lower q. The GAM amplitude is also shown to increase with plasma elongation, ?, while its frequency decreases.

The quiescent H (QH) mode, an edge localized mode (ELM)-free and high-confinement mode, combines well with an internal transport barrier to form quiescent double barrier stationary state, high performance plasmas. The QH-mode edge pedestal pressure is similar to that seen in ELMing phases of the same discharge, with similar global energy confinement. The pedestal density in early ELMing phases of strongly pumped counter injection discharges drops, and a transition to QH-mode occurs, leading to lower calculated edge bootstrap current. The plasmas current ramp experiment and ELITE code modelling of edge stability suggest that QH-modes lie near an edge current stability boundary. At high triangularity, QH-mode discharges operate at higher pedestal density and pressure and have achieved ITER level values of βPED and ν*.

Using resonant magnetic perturbations with toroidal mode number n = 3, we have produced H-mode discharges without edge localized modes (ELMs) which run with constant density and radiated power for periods up to about 2550 ms (17 energy confinement times). These ELM suppression results are achieved at pedestal collisionalities close to those desired for next step burning plasma experiments such as ITER and provide a means of eliminating the rapid erosion of divertor components in such machines which could be caused by giant ELMs. The ELM suppression is due to an enhancement in the edge particle transport which reduces the edge pressure gradient and pedestal current density below the threshold for peeling-ballooning modes. These n = 3 magnetic perturbations provide a means of active control of edge plasma transport.

Toroidal plasma rotation of the order of a few per cent of the Alfven velocity can stabilize the resistive wall mode (RWM) and extend the operating regime of tokamaks from the conventional, ideal magnetohydrodynamic (MHD) no-wall limit up to the ideal MHD ideal-wall limit. The stabilizing effect has been measured in DIII-D passively by measuring the critical plasma rotation required for stability and actively by probing the plasma with externally applied resonant magnetic fields. The comparison of these measurements to predictions of rotational stabilization of the sound wave damping and of the kinetic damping model using the MARS-F code results in qualitative agreement, but also indicates the need for further refinement of the measurements and models.

An approach has been implemented on the DIII-D tokamak to allow the correct determination of the plasma poloidal velocity from charge exchange spectroscopy measurements. Unlike usual techniques, the need for detailed atomic physics calculations to properly interpret the results is alleviated. Instead, the needed atomic physics corrections are self-consistently determined directly from the measurements, by making use of specially chosen viewing chords. Modeling results are presented that were used to determine a set of views capable of measuring the correction terms. We present the analysis of a quiescent H-mode discharge, illustrating that significant modifications to the velocity profiles are required in these high ion temperature conditions. We also present preliminary measurements providing a direct comparison of the standard cross-section correction to the atomic physics calculations.

Shear flow stabilization of edge turbulence in tokamaks has been the accepted paradigm for the improvement in confinement observed in high (H) confinement mode plasmas. Results on the generation of zonal flow and fields in finite β plasmas are presented. This theory yields a criterion for bifurcation from low to high (L–H) confinement mode, proportional to Te/Ln, where Te is the electron temperature and Ln is the density scale-length at the steepest part of the density gradient. When this parameter exceeds a critical value (mostly determined by the strength of the toroidal magnetic field), the transition occurs. The predicted threshold based on this parameter shows good agreement with edge measurements on discharges undergoing L–H transitions in DIII-D [J. L. Luxon, R. Anderson, F. Batty et al., in Proceedings of the 11th Conference on Plasma Physics and Controlled Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. The observed differences in the transitions with the reversal of the toroidal magnetic field are reconciled in terms of this critical parameter due to the differences in the density gradient scale-lengths in the edge. The theory also provides a possible explanation for lowered threshold power, pellet injection H modes in DIII-D, thereby providing a unified picture of the varied observations on the L–H transition.Shear flow stabilization of edge turbulence in tokamaks has been the accepted paradigm for the improvement in confinement observed in high (H) confinement mode plasmas. Results on the generation of zonal flow and fields in finite β plasmas are presented. This theory yields a criterion for bifurcation from low to high (L–H) confinement mode, proportional to Te/Ln, where Te is the electron temperature and Ln is the density scale-length at the steepest part of the density gradient. When this parameter exceeds a critical value (mostly determined by the strength of the toroidal magnetic field), the transition occurs. The predicted threshold based on this parameter shows good agreement with edge measurements on discharges undergoing L–H transitions in DIII-D [J. L. Luxon, R. Anderson, F. Batty et al., in Proceedings of the 11th Conference on Plasma Physics and Controlled Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. The observed differences in the transitions with th...

Advanced tokamak regimes obtained in ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TCV and Tore Supra experiments are assessed both in terms of their fusion performance and capability for ultimately reaching steady-state using data from the international internal transport barrier database. These advanced modes of tokamak operation are characterized by an improved core confinement and a modified current profile compared to the relaxed Ohmically driven one. The present results obtained in these experiments are studied in view of their prospect for achieving either long pulses ('hybrid' scenario with inductive and non-inductive current drive) or ultimately steady-state purely non-inductive current drive operation in next step devices such as ITER. A new operational diagram for advanced tokamak operation is proposed where the figure of merit characterizing the fusion performances and confinement, H × βN / q 295, is drawn versus the fraction of the plasma current driven by the bootstrap effect. In this diagram, present day advanced tokamak regimes have now reached an operational domain that is required in the non-inductive ITER current drive operation with typically 50% of the plasma current driven by the bootstrap effect (Green et al 2003 Plasma Phys. Control. Fusion 45 587). In addition, the existence domain of the advanced mode regimes is also mapped in terms of dimensionless plasmas physics quantities such as normalized Larmor radius, normalized collisionality, Mach number and ratio of ion to electron temperature. The gap between present day and future advanced tokamak experiments is quantitatively assessed in terms of these dimensionless parameters.

Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios.Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs.High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high βp are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and βp.The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U, DIII-D) in terms of their relevance for ITER. Successful plasma edge control, at high triangularity (~0.5) and high density (~0.7nGR), in ITB scenarios in JET is reported.Active control of ELMs by edge current, pellet injection, impurities and external magnetic perturbations creating an ergodic zone localized at the separatrix are discussed for present-day experiments and from the perspective of future reactors.

Results from recent experiments on the DIII-D tokamak have revealed many important details on transport barriers at the plasma edge and in the plasma core. These experiments include: (a) the formation of the H-mode edge barrier directly by pellet injection; (b) the formation of a quiescent H-mode edge barrier (QH-mode) which is free from edge localized modes, but which still exhibits good density and radiative power control; (c) the formation of multiple transport barriers, such as the quiescent double barrier (QDB) which combines an internal transport barrier with the quiescent H-mode edge barrier. Results from the pellet-induced H-mode experiments indicate that: (a) the edge temperature (electron or ion) does not need to attain a critical value for the formation of the H-mode barrier, (b) pellet injection leads to an increased gradient in the radial electric field, Er, at the plasma edge; (c) the experimentally determined edge parameters at barrier transition are well below the predictions of several theories on the formation of the H-mode barrier, (d) pellet injection can lower the threshold power required to form the H-mode barrier. The quiescent H-mode barrier exhibits good density control as the result of continuous magnetohydrodynamic activity at the plasma edge called the edge harmonic oscillation (EHO). The EHO enhances the edge particle transport whilst maintaining a good energy transport barrier. The ability to produce multiple barriers in the QDB regime has led to long duration, high-performance plasmas with βNH89 values of 7 for up to 10 times the confinement time. Density profile control in the plasma core of QDB plasmas has been demonstrated using on-axis electron cyclotron heating.

It is known that the toroidal angular momentum and the ion thermal energy are correlated in tokamak discharges heated by neutral beam injection. Here, data from ten years of measurements on DIII-D are considered, for representative discharges from all types and all conditions. The ratio of simple replacement times for momentum and energy is found to order this correlation indicating that these times are approximately equal, across the minor radius. Representative discharges of several types are discussed in more detail, as well as transport analysis results for the momentum and thermal ion diffusivities.

Details of the low-high (L-H) confinement transition are studied by a new technique which allows for an arbitrarily slow transition between the L- and H-modes on DIII-D. During the transition, the plasma is in an intermediate state (IM-mode) of temperature, density, confinement and edge shear flow. The IM-mode is characterized by periodic bursts of an edge instability, governed by relaxation oscillations, which evolve into type III edge localized modes (ELMs) as the neutral beam heating power is raised. An ELM-free H-mode is achieved when the edge pressure gradient is large enough to support shear flow sufficient to quell the IM-mode edge instability.

For the first time, data from eight different tokamaks have been combined in an international database for internal transport barriers (ITBs). An analysis of the data for the formation of an ITB with dominant ion heating shows a clear dependence of the threshold power on the minor radius and line-averaged electron density for the formation of ion ITBs. The dependence of ITB formation on the toroidal magnetic field is weak. For the formation of ITBs with dominant electron heating, the database is smaller, but for the threshold power a strong increase with plasma size and a weak toroidal field dependence could also be identified. Based on these results, an expression for the power required to form an ITB is given using global variables only. These results give a basis for the analysis of the database using local values (like magnetic shear) and a detailed comparison with theory-based models.

The formation of transport barriers in fusion plasmas is studied through examination of mechanisms which can stabilize plasma turbulence and, thereby, reduce turbulence-driven transport. These include the effects of E?B velocity shear, negative central magnetic shear and the Shafranov shift. Transport barriers formed at the plasma edge and in the plasma core are considered, as well as the formation of multiple barriers. The access conditions for barrier formation are examined, particularly by considering local conditions compared to global parameters. Processes that can destroy transport barriers, such as magnetohydrodynamic instabilities, are described. Plasmas with high confinement and good plasma stability then require effective control of plasma profiles and transport. Efforts towards real-time control of transport barriers are described.

Experiments in DIII-D, using non-axisymmetric magnetic perturbation fields in high-purity low toroidal rotation, He-4 plasmas have resulted in Type-I edge localized mode (ELM) suppression and mitigation. Suppression is obtained in plasmas with zero net input torque near the L-H power threshold using either electron cyclotron resonant heating (ECRH) or balanced co- and counter-I-p neutral beam injection (NBI) resulting in conditions equivalent to those expected in ITER's non-active operating phase. In low-power ECRH H-modes, periods with uncontrolled density and impurity radiation excursions are prevented by applying n = 3 non-axisymmetric magnetic perturbation fields. ELM suppression results from a reduction and an outward shift of the electron pressure gradient peak compared to that in the high-power ELMing phase. The change in the electron pressure gradient peak is primarily due to a drop in the pedestal temperature rather than the pedestal density.

Edge conditions in DIII-D are being quantified in order to provide insight into the physics of the H?mode regime. Several studies show that electron temperature is not the key parameter that controls the L-H transition. Gradients of edge temperature and pressure are much more promising candidates for elements of such parameters. They systematically increase during the L phases of discharges which make a transition to H?mode, and these increases are typically larger than the increases in the underlying quantities. The quality of H?mode confinement is strongly correlated with the height of the H?mode pedestal for the pressure. The gradient of the pressure is limited by MHD modes, in particular by ideal kink ballooning modes with finite mode number n. For a wide variety of discharges, the width of the barrier for electron pressure is well described by a relationship that is proportional to (?pedp)1/2. A new regime of confinement, called the quiescent H?mode, which provides steady state operation with no ELMs, low radiated power and normal H?mode confinement, has been discovered. A coherent edge MHD mode provides adequate particle transport to control the plasma density while permitting the pressure pedestal to remain almost identical to that observed in ELMing discharges.

A new sustained high-performance regime, combining discrete edge and core transport barriers, has been discovered in the DIII-D tokamak. Edge localized modes (ELMs) are replaced by a steady oscillation that increases edge particle transport, thereby allowing particle control with no ELM-induced pulsed divertor heat load. The core barrier resembles those usually seen with a low (L) mode edge, without the degradation often associated with ELMs. The barriers are separated by a narrow region of high transport associated with a zero crossing in the E x B shearing rate.

Charge exchange spectroscopy is one of the key ion diagnostics on the DIII-D tokamak. It allows measurement of impurity densities, toroidal and poloidal rotation speeds, ion temperatures, and the radial electric field. For the 2000 experimental campaign, we have replaced the intensified photodiode array detectors on the edge portion of the system with advanced charge-coupled device (CCD) detectors mounted on faster (f/4.7) Czerny–Turner spectrometers equipped with toroidal mirrors. The combination has improved the photoelectron signal level by about a factor of 20 and the signal to noise by a factor of 2–8, depending on the absolute signal level and readout mode. A major portion of the signal level improvement comes from the improved quantum efficiency of the back-illuminated, thinned CCD detector (70% to 85% quantum efficiency for the CCD versus 10% for the image intensifier) with the remainder coming from the faster spectrometer. The CCD camera also allows shorter minimum integration times: 0.33 ms whil...

The article reports results of experiments to further determine the underlying physics behind the formation and development of internal transport barriers (ITBs) in the DIII-D tokamak. The initial ITB formation occurs when the neutral beam heating power exceeds a threshold value during the early stages of the current ramp in low density discharges. This region of reduced transport, made accessible by suppression of long wavelength turbulence by sheared flows, is most evident in the ion temperature and impurity rotation profiles. In some cases, reduced transport is also observed in the electron temperature and density profiles. If the power is near the threshold, the barrier remains stationary and encloses only a small fraction of the plasma volume. If, however, the power is increased, the transport barrier expands to encompass a larger fraction of the plasma volume. The dynamic behaviour of the transport barrier during the growth phase exhibits rapid transport events that are associated with both broadening of the profiles and reductions in turbulence and associated transport. In some but not all cases, these events are correlated with the safety factor q passing through integer values. The final state following this evolution is a plasma exhibiting ion thermal transport at or below neoclassical levels. Typically the electron thermal transport remains anomalously high. Recent experimental results are reported in which RF electron heating was applied to plasmas with an ion ITB, thereby increasing both the electron and the ion transport. Although the results are partially in agreement with the usual E × B shear suppression hypothesis, the results still leave questions that must be addressed in future experiments.

Plasma shear effects on thermal ion transport in present advanced tokamak scenarios are analyzed. In order to overcome some limitations encountered from previous shear dependent models [e.g., V. Parail et al., Plasma Phys. Controlled Fusion 40, 805 (1998), and references therein], a semiempirical model combining the effects of the magnetic and E×B rotation shears is proposed. These shear dependences are based on simple theoretical arguments from turbulence studies. The dominant stabilizing term of our shear correction is the shear in toroidal rotation. The predictive modeling of the formation and evolution of thermal ion internal transport barriers in relevant advanced scenarios of Tokamak Fusion Test Reactor [E. J. Synakowski et al., Phys. Plasmas 4, 1736 (1997)], Doublet III-D (DIII-D) [B. W. Rice et al., Nucl. Fusion 36, 1271 (1996)] and Joint European Torus (JET) [F. X. Soldner and the JET Team, Plasma Phys. Controlled Fusion 39, B353 (1997)] with our shear correction is presented. A multimachine test...

Energy transport in discharges with off-axis neutral beam injection (NBI) has been compared with transport in similar discharges with on-axis NBI. A two fluid power balance analysis assuming purely diffusive transport indicates substantial improvements in the plasma centre in both the ion and the electron channels during off-axis NBI. The ion thermal diffusivity, χi, and the angular momentum diffusivity, χ, in the plasma centre improve by over 2 orders of magnitude, and χi decreases below the Chang-Hinton neoclassical level. A detailed error analysis of the energy transport has been carried out. The electron thermal diffusivity in the plasma centre improves by an order of magnitude during off-axis NBI. The predicted peak ion temperature, assuming the same ion diffusivity profile for off-axis NBI as with on-axis NBI, is nearly a factor of 2 lower than the measured ion temperature profile with off-axis NBI. In addition, assuming unchanged diffusivities, significant inward energy flow is required to explain the observed ion temperature profiles during off-axis NBI.

High spatially resolved measurements of the radial electric field, Er, have been made across the transition from L mode to H mode plasmas for many different plasma parameters and conditions. The evolution of the well-like structure of the Er profile formed at the L-H transition has been investigated. No distinct variation in the shape or width of the Er well at the L-H transition is observed as a function of the plasma parameters investigated, such as the plasma current, the toroidal magnetic field and the plasma density. The value of Er is negative just inside the last closed flux surface (LCFS) for all the plasmas studied. There is a variation in the depth of the Er well for different conditions. The experimental results have been compared with theoretical predictions for suppression of plasma turbulence by sheared E × B plasma flow.

Plasma discharges with negative central magnetic shear (NCS) and access to the ballooning second stability regime have been produced on DIII-D with edge plasma conditions similar to those in the standard L mode confinement regime. Confinement enhancement factors up to H identical to tau E/ tau ITER-89P~2.5 are obtained while maintaining the L mode edge. Compared with discharges with monotonic q profiles, highly peaked toroidal rotation (fphi (0) approximately=30-70 kHz), ion temperature (Ti(0) approximately=15-22 keV) and density (ne(0)/ne approximately=2.2) profiles are observed. This regime has yielded the highest DD neutron rates observed to date on DIII-D. Steep pressure gradients drive a large bootstrap current, with up to ~75% total non-inductive current drive. These features make this an attractive advanced tokamak operating regime for further study

Measurements of local electron density n{sub e}, electron temperature T{sub e}, and ion temperature T{sub i} have been made at the very edge of the plasma just prior to the transition into H-mode for four different single parameter scans in the DIII-D tokamak. The means and standard derivations of n{sub e}, T{sub e}, and T{sub i} under these conditions for a value of the normalized toroidal flux of 0.98 are respectively, 1.5 {+-} 0.7 x 10{sup 19} m{sup -3}, 0.051 {+-} 0.016 keV, and 0.14 {+-} 0.03 keV. The threshold condition for the transition is more sensitive to temperature than to density. The data indicate that the dependence is not as simple as a requirement for a fixed value of the ion collisionality.

Experiments have been carried out on the DIII-D tokamak to investigate whether off-axis NBI can: (a) drive significant perpendicular flow to lead to increased suppression of turbulence and improved confinement, and (b) be used to control the radial electric field profile. Measurements of both impurity ion poloidal and toroidal rotation profiles were made using charge exchange recombination spectroscopy. These experiments used a low current, low elongation (I{sub p} = 0.5 MA, {kappa} = 1.2) plasma whose magnetic axis was shifted 36 cm vertically upward from the vessel midplane and then shifted downward to be centered on the midplane later in the discharge. 10.7 MW of beam power was applied to maximize NBI effect while operating at low target densities and high temperature to minimize poloidal damping. Results from these experiments show a slight increase in impurity ion poloidal rotation velocity during the vertical shifted phase of off-axis NBI discharge. The toroidal rotation profile is more peaked during off-axis NBI. Both these effects lead to a change in the V x B contribution to the radial electric field during off-axis NBI.

Charge exchange spectroscopy is one of the key ion diagnostics on the DIII-D tokamak. It allows determination of ion temperature, poloidal and toroidal velocity, impurity density, and radial electric field Er throughout the plasma. For the 2003 experimental campaign, we replaced the intensified photodiode array detectors on the central portion of the DIII-D charge exchange spectroscopy system with advanced charge–coupled device (CCD) detectors mounted on faster (f/4.7) Czerny–Turner spectrometers equipped with toroidal mirrors. The CCD detectors are improved versions of the ones installed on our edge system in 1999. The combination improved the photoelectron signal level by about a factor of 20 and the signal to noise by a factor of 2–8, depending on the absolute signal level. The new cameras also allow shorter minimum integration times while archiving to PC memory: 0.552 ms for the slower, lower-read noise (15 e) readout mode and 0.274 ms in the faster, higher-read noise (30 e) mode.

The scaling of cross‐field heat transport with relative gyroradius ρ* was measured in low (L) and high (H) mode tokamak plasmas using the technique of dimensionally similar discharges. The relative gyroradius scalings of the electron and ion thermal diffusivities were determined separately using a two‐fluid transport analysis. For L‐mode plasmas, the electron diffusivity scaled as χe∝χBρ1.1±0.3* (gyro‐Bohm‐like) while the ion diffusivity scaled as χi∝χBρ−0.5±0.3* (worse than Bohm‐like). The results were independent of the method of auxiliary heating (radio frequency or neutral beam). Since the electron and ion fluids had different gyroradius scalings, the effective diffusivity and global confinement time scalings were found to vary from gyro‐Bohm‐like to Bohm‐like depending upon whether the electron or ion channel dominated the heat flux. This last property can explain the previously disparate results with dimensionally similar discharges on different fusion experiments that have been published. Experimen...

A better understanding of helium transport in the plasma core and edge in enhanced confinement regimes is now emerging from recent experimental studies on DIII-D. Overall, the results are encouraging. Significant helium exhaust ({tau}*{sub He}/{tau}{sub E} {approximately} 11) has been obtained in a diverted, ELMing H-mode plasma simultaneous with a central source of helium. Detailed analysis of the helium profile evolution indicates that the exhaust rate is limited by the exhaust efficiency of the pump ({approximately}5%) and not by the intrinsic helium transport properties of the plasma. Perturbative helium transport studies using gas puffing have shown that D{sub He}/X{sub eff}{approximately}1 in all confinement regimes studied to date (including H-mode and VH-mode). Furthermore, there is no evidence of preferential accumulation of helium in any of these regimes. However, measurements in the core and pumping plenum show a significant dilution of helium as it flows from the plasma core to the pumping plenum. Such dilution could be the limiting factor in the overall removal rate of helium in a reactor system.

The scaling of heat diffusion with relative gyroradius ${\ensuremath{\rho}}_{*}$ is measured on the DIII-D tokamak using dimensionally similar discharges. For the first time, a two-fluid transport analysis allows the ${\ensuremath{\rho}}_{*}$ scaling of the electron and ion fluids to be determined separately. The electron diffusivity is found to scale like ${\ensuremath{\chi}}_{e}\ensuremath{\propto}{\ensuremath{\chi}}_{B}{\ensuremath{\rho}}_{*}^{1.1\ifmmode\pm\else\textpm\fi{}0.3}$ (gyro-Bohm-like) while the ion diffusivity is found to scale like ${\ensuremath{\chi}}_{i}\ensuremath{\propto}{\ensuremath{\chi}}_{B}{\ensuremath{\rho}}_{*}^{\ensuremath{-}0.5\ifmmode\pm\else\textpm\fi{}0.3}$ (Goldston-like). The scaling of the global confinement time and the effective diffusivity can vary from gyro-Bohm-like to Bohm-like depending upon whether the electrons or ions dominate the heat transport.

A regime of improved H mode energy confinement, the VH mode, was obtained following boronization of the DIII-D Tokamak vacuum vessel. The gradual confinement improvement in VH mode is associated with expansion of the H mode pedestal and reduction of thermal diffusivity in a region just inside the pedestal. Disappearance of density fluctuation bursts is associated with a confinement improvement in the latter region. The VH mode is terminated in a rapid global energy loss which is initiated by a mode with toroidal mode number, n approximately 5, consistent with calculated instability to an edge localized ideal kink. The improved confinement in VH mode is consistent with the extension of the region of the E*B velocity shear turbulence suppression zone further in from the plasma boundary. Expansion of the ideal ballooning second stability region and of the region with drift reversal stabilization of trapped particle modes are also considered as possible explanations for the confinement improvement

Detailed information on the changes in the edge plasma at the L-H transition has been obtained by using charge exchange recombination spectroscopy with high spatial resolution (6 mm) and high time resolution (0.5 ms). Measurements of the radial electric field, Er, show that Er decreases abruptly at the transition resulting in the formation of a negative well like radial profile just within the separatrix. There is a well developed region of gradient in Er which extends inwards 1 to 2 cm into the plasma from the magnetic separatrix. This region of Er' exhibits the greatest increase in impurity ion poloidal rotation at the L-H transition and the greatest reduction in plasma fluctuations. A transport barrier is formed in this same region as is signified by large increases in the observed gradients of ion temperature, carbon density, electron temperature and electron density. The development of the region of significant Er', the increase in impurity ion poloidal rotation, the reduction of plasma turbulence and the formation of the transport barrier all occur simultaneously with the drop in the Dalpha signal which marks the transition. The well like structure of Er is located on and inside the separatrix as is clearly defined by the variation of the signal amplitudes of the spatial chords spanning the separatrix. The well like structure of Er moves with the separatrix as the separatrix location changes and is maintained throughout the H mode

A regime of improved H-mode energy confinement, VH-mode, is obtained in the DIII-D tokamak with adequate vessel conditioning. The improved confinement in VH-mode is consistent with the extension of the region of high E*B velocity shear turbulence suppression zone further in from the plasma boundary. The energy confinement enhancement in VH-mode can be limited by ELMs, localized momentum transfer events, or operation at high heating power or low q. Energy confinement enhancement improves with increasing triangularity of the plasma cross section and is independent of elongation. The termination of VH-mode is associated with an edge localized kink mode which is destabilized by both the large edge pressure gradient and edge current density.

Poloidal and toroidal rotation of the fuel ions (He2+) and the impurity ions (C6+ and B5+) in H-mode helium plasmas have been investigated in the DIII-D tokamak by means of charge exchange recombination spectroscopy, resulting in the discovery that the fuel ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction. The radial electric field obtained from radial force balance analysis of the measured pressure gradients and rotation velocities is shown to be the same regardless of which ion species is used and therefore is a more fundamental parameter than the rotation flows in studying H-mode phenomena. It is shown that the three contributions to the radial electric field (diamagnetic, poloidal rotation, and toroidal rotation terms) are comparable and consequently the poloidal flow does not solely represent the E*B flow. In the high-shear edge region, the density scale length is comparable to the ion poloidal gyroradius, and thus neoclassical theory is not valid there. In view of this new discovery that the fuel and impurity ions rotate in opposite sense, L-H transition theories based on the poloidal rotation may require improvement.

Poloidal and toroidal rotation of the main ions (${\mathrm{He}}^{2+}$) and the impurity ions (${\mathrm{C}}^{6+}$ and ${\mathrm{B}}^{5+}$) in H-mode helium plasmas have been measured via charge exchange recombination spectroscopy in the DIII-D tokamak. It was discovered that the main ion poloidal rotation is in the ion diamagnetic drift direction while the impurity ion rotation is in the electron diamagnetic drift direction, in qualitative agreement with the neoclassical theory. The deduced radial electric field in the edge is of the same negative-well shape regardless of which ion species is used, validating the fundamental nature of the electric field in L-H transition phenomenology.

Modification of the two existing DIII-D neutral beamlines is planned to allow vertical steering to provide off-axis neutral beam current drive (NBCD) peaked as far off-axis as half the plasma minor radius. New calculations for a downward-steered beam indicate strong current drive with good localization off-axis so long as the toroidal magnetic field, B T, and the plasma current, I p, point in the same direction. This is due to good alignment of neutral beam injection (NBI) with the local pitch of the magnetic field lines. This model has been tested experimentally on DIII-D by injecting equatorially mounted NBs into reduced size plasmas that are vertically displaced with respect to the vessel midplane. The existence of off-axis NBCD is evident in the changes seen in sawtooth behaviour in the internal inductance. By shifting the plasma upwards or downwards, or by changing the sign of the toroidal field, off-axis NBCD profiles measured with motional Stark effect data and internal loop voltage show a difference in amplitude (40–45%) consistent with differences predicted by the changed NBI alignment with respect to the helicity of the magnetic field lines. The effects of NBI direction relative to field line helicity can be large even in ITER: off-axis NBCD can be increased by more than 30% if the B T direction is reversed. Modification of the DIII-D NB system will strongly support scenario development for ITER and future tokamaks as well as provide flexible scientific tools for understanding transport, energetic particles and heating and current drive.

The L to H transition in the DIII‐D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (IAEA, Vienna, 1987), Vol. I, p. 159] is associated with two clear signatures: edge density fluctuations are abruptly suppressed (in ≊100 μsec), while the edge poloidal rotation velocity vθ increases, implying that the radial electric field Er becomes more negative. Detailed new spectroscopic profile measurements show that the changes in vθ and Er generate a region of sheared electric field and poloidal flow of width ≊3–5 cm. This region develops simultaneously with, and has the same spatial extent as, the edge fluctuation suppression zone as measured using a reflectometer system. Furthermore, the radial extent of the shear and fluctuation suppression zones encompass the location of the H‐mode edge transport barrier. These observations are consistent with recent theoretical models of the L–H transition, and a comparison with these theories is presented. Data are also presented on the evolution of edge parameters and density fluctuations after the transition: the shear and fluctuation suppression layers are maintained for the duration of the quiescent H‐mode phase, while relative density fluctuation levels decrease and interior plasma confinement gradually improves. Precursors to several different types of edge localized mode (ELMs) are also discussed.

A study of three-dimensional (3D) perturbed magnetic field structures and transport for edge localized mode control experiments with resonant magnetic perturbations at DIII-D is presented. We focus on ITER-Similar Shape plasmas at ITER relevant electron pedestal collisionalities . This study is performed in comparison with results from TEXTOR-Dynamic Ergodic Divertor circular limiter plasmas. For both experiments the magnetic field structure is analyzed in the vacuum paradigm—superimposing the external RMP field on the unperturbed equilibrium. For TEXTOR L-mode plasmas this description holds for normalized poloidal flux ΨN > 0.7 without tearing modes driven by the RMP field. For DIII-D H-mode plasmas the validity of this approach still needs to be established. In this paper a method is discussed to diagnose the degree of edge stochastization based on a comparison between modeled magnetic footprints on the divertor targets and experimental data. Clear evidence is presented for the existence of a generic separatrix perturbation causing striation of target particle fluxes. However, heat fluxes into these striations are small. This observation can be explained by accounting for the different heat and particle source locations and the 3D trajectories of the open, perturbed field lines toward the divertor target. Analysis of the transport characteristics filling the perturbed separatrix lobes based on initial EMC3/EIRENE modeling suggests the existence of open field lines connecting the stochastic edge to the target pattern. However, the width and inward most extent of the actual stochastic layer cannot yet be quantified.

Large Type-I edge localized modes (ELMs) are completely eliminated with small n = 3 resonant magnetic perturbations (RMP) in low average triangularity, , plasmas and in ITER similar shaped (ISS) plasmas, , with ITER relevant collisionalities . Significant differences in the RMP requirements and in the properties of the ELM suppressed plasmas are found when comparing the two triangularities. In ISS plasmas, the current required to suppress ELMs is approximately 25% higher than in low average triangularity plasmas. It is also found that the width of the resonant q95 window required for ELM suppression is smaller in ISS plasmas than in low average triangularity plasmas. An analysis of the positions and widths of resonant magnetic islands across the pedestal region, in the absence of resonant field screening or a self-consistent plasma response, indicates that differences in the shape of the q profile may explain the need for higher RMP coil currents during ELM suppression in ISS plasmas. Changes in the pedestal profiles are compared for each plasma shape as well as with changes in the injected neutral beam power and the RMP amplitude. Implications of these results are discussed in terms of requirements for optimal ELM control coil designs and for establishing the physics basis needed in order to scale this approach to future burning plasma devices such as ITER.

Recent DIII-D [J. L. Luxon , Nucl. Fusion 43, 1813 (2003)] experiments show a correlation between the extent of overlap of magnetic islands induced in the edge plasma by perturbation coils and complete suppression of Type-I edge localized modes (ELMs) in plasmas with ITER-like electron pedestal collisionality nu(e)*similar to 0.1, flux surface shape and low edge safety factor (q(95)approximate to 3.6). With fixed amplitude n=3 resonant magnetic perturbation (RMP), ELM suppression is obtained only in a finite window in the edge safety factor (q(95)) consistent with maximizing the resonant component of the applied helical field. ELM suppression is obtained over an increasing range of q(95) by either increasing the n=3 RMP strength, or by adding n=1 perturbations to "fill in" gaps between islands across the edge plasma. The suppression of Type-I ELMs correlates with a minimum width of the edge region having magnetic islands with Chirikov parameter >1.0, based on vacuum calculations of RMP mode components excluding the plasma response or rotational shielding. The fraction of vacuum magnetic field lines that are lost from the plasma, with connection length to the divertor targets comparable to an electron-ion collisional mean free path, increases throughout the island overlap region in the ELM suppressed case compared with the ELMing case. (C) 2008 American Institute of Physics.

Doppler‐shift spectroscopy has shown that finite values of poloidal rotation velocity vθ and of radial electric field Er exist at the edge of a tokamak plasma and that dramatic increases occur in these parameters at an L‐H transition. Er is negative in the L mode and becomes more negative in the H mode; vθ increases in magnitude at the transition. In addition, the radial gradients (shear) of vθ and Er are large and these gradients also increase at the L‐H transition. These results are based on measurements of Doppler shifts of light emitted by He II ions, located in a region about 1–3 cm inside the separatrix. These observations have been made with horizontally viewing and vertically viewing spectrometer systems on the DIII‐D tokamak. The nearly orthogonal views of these systems are used to determine the plasma’s flow velocity in terms of the orthogonal sets vθ and vφ or of v⊥ and v∥. Knowledge of v⊥ is used to calculate Er from the force balance equation for a single ion species. The existing results imp...

The L to H transition in the DIII‐D tokamak [Plasma Physics and Controlled Nuclear Fusion Research 1986 (IAEA, Vienna, 1987), Vol. I, p. 159] is associated with a decrease in the edge density and magnetic fluctuations. In addition, in single‐null divertor plasmas, a reduction in the heat flux asymmetry between the inner and outer divertor hit spots occurs. These observations indicate that the L to H transition is associated with the reduction in anomalous, fluctuation‐connected transport across the outer midplane of the plasma. Magnetic fluctuations are measured with a poloidally distributed set of Mirnov loops while density fluctuations are detected with multiple fixed‐frequency microwave reflectometers. Spectroscopic observations of edge poloidal and toroidal rotation have allowed the inference that the radial electric field just inside the separatrix is negative in the L mode and becomes more negative at the L to H transition. These changes in fluctuations and in the edge electric field occur in plasma...

Recent DIII-D experiments have significantly improved the understanding of internal transport barriers (ITBs) that are triggered close to the time when an integer value of the minimum in q is crossed. While this phenomenon has been observed on many tokamaks, the extensive transport and fluctuation diagnostics on DIII-D have permitted a detailed study of the generation mechanisms of q-triggered ITBs as pertaining to turbulence suppression dynamics, shear flows, and energetic particle modes. In these discharges, the evolution of the q profile is measured using motional Stark effect polarimetry and the integer qmincrossings are further pinpointed in time by the observation of Alfvén cascades. High time resolution measurements of the ion and electron temperatures and the toroidal rotation show that the start of improved confinement is simultaneous in all three channels, and that this event precedes the traversal of integer qminby 5-20 ms. There is no significant low-frequency magnetohydrodynamic activity prior to or just after the crossing of the integer qminand hence magnetic reconnection is determined not to be the precipitant of the confinement change. Instead, results from the GYRO code point to the effects of zonal flows near low order rational q values as playing a role in ITB triggering. A reduction in local turbulent fluctuations is observed at the start of the temperature rise and, concurrently, an increase in turbulence poloidal flow velocity and flow shear is measured with the beam emission spectroscopy diagnostic. For the case of a transition to an enduring internal barrier the fluctuation level remains at a reduced amplitude. The timing and nature of the temperature, rotation, and fluctuation changes leading to internal barriers suggests transport improvement due to increased shear flow arising from the zonal flow structures. © 2005 American Institute of Physics.

Shear flow stabilization of edge turbulence leads to self-organized high (H) confinement modes in tokamak plasmas. Thus understanding the mechanisms for generation of shear/zonal flow and fields in finite β plasmas is an important area of research. A brief review of various mechanisms for shear flow generation and discussion of our recent theory which yields a criterion for bifurcation from low to high (L-H) confinement mode is presented. The predicted threshold based on this parameter shows good agreement with edge measurements on discharges undergoing L-H transitions in DIII-D with ▽B both towards and away from the X-point, as well as for pellet induced H-modes.