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For the first time, experiments on the DIII-D tokamak have demonstrated electron cyclotron current drive with more than double the conventional efficiency by tailoring the wave–particle interactions in velocity space using a novel ‘top launch’ geometry. Steering the EC waves to propagate nearly parallel to the resonance drives current more efficiently by (1) selective damping on electrons with higher parallel velocity v||, and (2) longer absorption path to compensate for inherently weaker absorption at higher v||. Experiments using a fixed-injection top launch system find an optimal velocity space interaction for maximum current drive efficiency at ρ ∼ 0.5 where the ease of drawing out a high v|| electron tail is balanced by sufficient absorption.

A program of electron cyclotron heating (ECH) with 4 MW at 140 GHz has been launched for developing scenarios of stable high performance through control of the pressure and current density profiles...

Electron cyclotron heating and current drive (ECH/ECCD) has become an essential tool for fusion plasma research in toroidal devices. In the Korea Superconducting Tokamak Advanced Research (KSTAR) t...

To move to a fusion DEMO power plant after ITER, a Fusion Nuclear Science Facility (FNSF) is needed in addition to ITER and research in operating tokamaks and those under construction. The FNSF wil...

Efficient off-axis current drive is crucial for economic, steady-state tokamak fusion power plants. “Top Launch” electron cyclotron current drive (ECCD) is one promising method for driving strong off-axis current to achieve the desired broad current profile for the Advanced Tokamak regime. New simulations show that by launching the electron cyclotron (EC) waves downwards (or upwards) nearly parallel to the resonance plane with a large toroidal steering, higher current drive efficiency can be obtained at large radii owing to the large Doppler shift, wave-particle interactions on HFS of the plasma, and the long absorption path. Implementations on CFETR and DIII-D are simulated identifying clear benefits and optimal configurations. The design of a prototype test being installed on DIII-D is presented, which will experimentally validate whether top launch ECCD can be an improved efficiency current drive technique for future fusion reactors.

The electron cyclotron heating and current drive complex on the DIII-D tokamak presently comprises six gyrotrons injecting rf power from the low field side at 110 GHz, the 2f ce resonance at the center of the vacuum chamber. Typical injected rf power is 600–650 kW per gyrotron. The launched rf can be directed over ±20° toroidally to create both co- and counter-current drive and scanned over 40° poloidally to permit the injected rf beams to intersect, and be absorbed at, the second harmonic resonance anywhere in the tokamak upper half plane. The elliptical polarization is controlled so that the desired extraordinary or ordinary modes are excited for any injection geometry. The maximum injected energy on a single plasma shot has been 16.6 MJ for six gyrotrons injecting a total of 3.4 MW for 5 seconds.

The response of plasma parameters and broad wavenumber turbulence (1–39 cm−1, kρs = 0.1–10, relevant to ion temperature gradient, trapped electron mode and electron temperature gradient mode turbulence, here ρs = ion gyroradius) to auxiliary electron cyclotron heating (ECH) is reported on. One fluid thermal fluxes and diffusivities increase appreciably with ECH. Significant changes to the density fluctuations over the full range of measured wavenumbers are observed, with an increase for lower wavenumbers and a more spatially complicated response at high k. Spatially resolved high k measurements (k = 39 cm−1, kρs = 4–10) show a varying response to ECH, with decreasing at r/a = 0.35 and increasing at r/a = 0.6 and 1. These variations were found to have a positive correlation with ∇Te evaluated at nearby locations, consistent with a ∇Te drive. Comparison of the changes in high k fluctuation levels with linear gyrokinetic growth rates show qualitative agreement at the innermost location, r/a = 0.35 and disagreement at r/a = 0.6.

The m = 2/n = 1 neoclassical tearing mode (NTM) has been observed to strongly degrade confinement and frequently lead to a disruption in high β discharges in DIII-D if allowed to grow to a large size. Stabilization of grown NTMs by the application of a highly localized electron cyclotron current drive (ECCD) at the island location has led to operation at an increased plasma pressure, up to the no-wall kink limit. After the NTM is stabilized by the ECCD, the correct location for the current drive is maintained using information from real-time equilibrium reconstructions which include measurements from the motional Stark effect diagnostic. This same process is used alternatively to prevent the mode from ever growing, leading to performance at the pressure limit in high performance hybrid discharges with β above 4%. Modelling using the modified Rutherford equation shows that the required power is in close agreement with the experimental threshold for the prevention of the 2/1 NTM.

Non-rotating (`locked') magnetic islands often lead to complete losses of confinement in tokamak plasmas, called major disruptions. Here locked islands were suppressed for the first time, by a combination of applied three-dimensional magnetic fields and injected millimetre waves. The applied fields were used to control the phase of locking and so align the island O-point with the region where the injected waves generated non-inductive currents. This resulted in stabilization of the locked island, disruption avoidance, recovery of high confinement and high pressure, in accordance with the expected dependencies upon wave power and relative phase between O-point and driven current., Comment: 4 figures

A new mechanism for driving current off-axis in high beta tokamaks using fast electromagnetic waves, called Helicons, will be experimentally tested for the first time in the DIII-D tokamak. This method is calculated to be more efficient than current drive using electron cyclotron waves or neutral beam injection, and it may be well suited to reactor-like configurations [1]. DIII-D can provide the conditions and measurement capabilities for a quantitative evaluation. At around 500 MHz, the optimum for DIII-D, an injected power of 1 MW would be adequate for these measurements and provide the equivalent current drive of a 2.5 MW neutral beam source. A “combline” antenna, which consists of many inductively coupled, electrostatically shielded, modular resonators, will be used to couple to the fast wave.

The first suppression of the important and deleterious m = 2/n = 1 neoclassical tearing mode (NTM) is reported using electron cyclotron current drive (ECCD) to replace the 'missing' bootstrap current in the island O-point. Experiments on the DIII-D tokamak verify that the maximum shrinkage of the m = 2/n = 1 island occurs when the ECCD location coincides with the q = 2 surface. The DIII-D plasma control system is put into a 'search and suppress' mode to make small changes in the toroidal field to find and lock onto the optimum position, based on real time measurements of dB?/dt, for complete m = 2/n = 1 NTM suppression by ECCD. The requirements of the ECCD for complete island suppression are well modelled by the modified Rutherford equation for the DIII-D plasma conditions.

Electron cyclotron current drive (ECCD) experiments on the DIII-D tokamak are solidifying the physics basis for localized, off-axis current drive, the goal being to validate a predictive model for ECCD. The ECCD profiles are determined from the magnetic field pitch angles measured by motional Stark effect polarimetry. The measured ECCD switches from the co- to the counter-direction as the toroidal injection angle is varied with a profile width that is in accordance with ray tracing calculations. Tests of electron trapping in low beta plasmas show that the ECCD efficiency decreases rapidly as the deposition is moved off-axis and towards the outboard side of the plasma, but the detrimental effects of electron trapping on the current drive are greatly reduced in high beta plasmas. Overall, the measured ECCD is in good agreement with theoretical calculations using a quasilinear Fokker–Planck code over a wide range of injection angles and plasma parameters.

Substantial progress has been made towards both understanding and control of internal transport barriers (ITBs) on DIII-D, resulting in the discovery of a new sustained high performance operating mode termed the quiescent double barrier (QDB) regime. The QDB regime combines core transport barriers with a quiescent ELM-free H mode edge (termed QH mode), giving rise to separate (double) core and edge transport barriers. The core and edge barriers are mutually compatible and do not merge, resulting in broad core profiles with an edge pedestal. The QH mode edge is characterized by ELM-free behaviour with continuous multiharmonic MHD activity in the pedestal region and has provided density and radiated power control for longer than 3.5 s (25τE) with divertor pumping. QDB plasmas are long pulse high performance candidates, having maintained a βN H89 product of 7 for five energy confinement times (Ti ≤ 16 keV, βN ≤ 2.9, H89 ≤ 2.4 τE ≤ 150 ms, DD neutron rate Sn ≤ 4 × 1015 s-1). The QDB regime has only been obtained in counter-NBI discharges (injection antiparallel to the plasma current) with divertor pumping. Other results include successful expansion of the ITB radius using (separately) both impurity injection and counter-NBI, and the formation of ITBs in the electron thermal channel using both ECH and strong negative central shear (NCS) at high power. These results are interpreted within a theoretical framework in which turbulence suppression is the key to ITB formation and control, and a decrease in core turbulence is observed in all cases of ITB formation.

Current driven by the fast Alfven wave is measured at the fourth and eighth harmonics of the deuterium ion cyclotron frequency in identical plasmas on the DIII-D tokamak. In non-sawtoothing discharges with neutral beam injection heating, the radial profile of the fast wave current drive (FWCD) is determined by the response of the loop voltage profile to co- and counter-antenna phasings. The dimensionless current drive efficiency is a factor of two greater for the eighth harmonic case compared to the fourth harmonic case. Modelling of the fast wave absorption using a ray tracing code shows that the decrease in FWCD efficiency for the latter situation can be explained by high harmonic damping of the fast waves on energetic ions.

Significant progress in obtaining high performance discharges lasting many energy confinement times in the DIII-D tokamak has been realized in recent experimental campaigns. Normalized performance ∼10 has been sustained for more than 5>E with qmin > 1.5. (The normalized performance is measured by the product βNH89, indicating the proximity to the conventional β limits and energy confinement quality, respectively.) These H mode discharges have an ELMing edge and β < 5%. The limit to increasing β is a resistive wall mode, rather than the tearing modes as previously observed. Confinement remains good despite qmin > 1. The global parameters were chosen to optimize the potential for fully non-inductive current sustainment at high performance, which is a key program goal for the DIII-D facility. Measurement of the current density and loop voltage profiles indicate that ≈75% of the current in the present discharges is sustained non-inductively. The remaining ohmic current is localized near the half-radius. The electron cyclotron heating system is being upgraded to replace this remaining current with ECCD. Density and β control, which are essential for operating advanced tokamak discharges, were demonstrated in ELMing H mode discharges with βNH89 ≈ 7 for up to 6.3 s or ≈34τE. These discharges appear to have stationary current profiles with qmin ≈ 1.05 in agreement with the current profile relaxation time ≈1.8 s.

Experimental evidence is reported of an internal kink instability driven by a new mechanism: barely trapped suprathermal electrons produced by off-axis electron cyclotron heating on the DIII-D tokamak. It occurs in plasmas with an evolving safety factor profile $q(r)$ when ${q}_{\mathrm{min}}$ approaches 1. This instability is most active when ECCD is applied on the high field side of the flux surface. It has a bursting behavior with poloidal/toroidal mode $\mathrm{number}\phantom{\rule{0ex}{0ex}}=\phantom{\rule{0ex}{0ex}}m/n\phantom{\rule{0ex}{0ex}}=\phantom{\rule{0ex}{0ex}}1/1$. In positive magnetic shear plasmas, this mode becomes the fishbone instability. This observation can be qualitatively explained by the drift reversal of the barely trapped suprathermal electrons.

Localized non-inductive currents due to electron cyclotron wave absorption have been measured on the DIII-D tokamak. Clear evidence of non-inductive currents is seen on the internal magnetic field measurements by motional Stark effect spectroscopy. The magnitude and location of the non-inductive current is evaluated by comparing the total and ohmic current profiles of discharges with and without electron cyclotron wave power. The measured current agrees with Fokker-Planck calculations near the magnetic axis, but exceeds the predicted value as the location of the current drive is moved to the half-radius.

Localized currents due to electron cyclotron current drive have been measured for the first time in experiments on the DIII-D tokamak. The location of driven current in the plasma has been varied from near the center of the tokamak out to half of the minor radius. The measured current drive efficiency agrees with quasilinear Fokker-Planck calculations near the center and exceeds the predicted value with increasing minor radius. Reduction of the trapped electron fraction due to finite collisionality is a leading candidate to explain the discrepancy. (c) 1999 The American Physical Society.

The expansion and upgrading of the electron cyclotron heating and current drive (ECH/ECCD) gyrotron complex on the DIII-D tokamak are continuing with the addition of the first of a series of depressed collector tubes in the 1 MW class. The ultimate goal is a 10 gyrotron system with rapid steering of the rf beams and full integration into the DIII-D Plasma Control System using the real time EFIT equilibrium calculation to determine the ECH/ECCD deposition locations to guide requirements for both steering and injected power.

The Electron Cyclotron Heating (ECH) system on DIII-D consists of six 110 GHz gyrotrons with 6 MW installed power for pulses limited administratively to 5 s in length. The transmission coefficient is better than -1.1 dB for four of the transmission lines, close to the theoretical value. A new depressed collector gyrotron was recently installed and is injecting up to 720 kW of power into DIII-D during 2013 tokamak operations. Three of the four dual waveguide launchers, which can steer the rf beams ±20 degrees both poloidally and toroidally, were used for real-time neoclassical tearing mode control and suppression with increased poloidal scanning speed up to 60 deg/s and positioning accuracy of the beams of ±2 mm at the plasma center. The ECH capabilities on DIII-D are being steadily updated, leading to increased experimental flexibility and high reliability of the system. In the past year the ECH system reliability reached 87%, for 2352 successful individual gyrotron shots into DIII-D. Planning is under way for the addition of two new depressed collector gyrotrons, one at 110 GHz, 1.2 MW and another at 117.5 GHz, 1.5 MW generated power, both of which are being manufactured at Communications and Power Industries (CPI).

Several new protective systems are being installed on the DIII D tokamak to increase the safety margins for plasma operations with injected ECH power at densities approaching cutoff. Inadvertent overdense operation has previously resulted in reflection of an rf beam back into a launcher causing extensive arcing and melt damage on one waveguide line. Damage to microwave diagnostics, which are located on the same side of the tokamak as the ECH launchers, also has occurred. Developing a reliable microwave based interlock to protect the many vulnerable systems in DIII-D has proved to be difficult. Therefore, multiple protective steps have been taken to reduce the risk of damage in the future. Among these is a density interlock generated by the plasma control system, with setpoint determined by the ECH operators based on rf beam trajectories and plasma parameters. Also installed are enhanced video monitoring of the launchers, and an ambient light monitor on each of the waveguide systems, along with a Langmuir probe at the mouth of each launcher. Versatile rf monitors, measuring forward and reflected power in addition to the mode content of the rf beams, have been installed as the last miter bends in each waveguide line. As these systems are characterized, they are being incorporated in the interlock chains, which enable the ECH injection permits. The diagnostics most susceptible to damage from the ECH waves have also been fitted with a variety of protective devices including stripline filters, thin resonant notch filters tuned to the 110 GHz injected microwave frequency, blazed grating filters and shutters. Calculations of rf beam trajectories in the plasmas are performed using the TORAY ray tracing code with input from kinetic profile diagnostics. Using these calculations, strike points for refracted beams on the vacuum vessel are calculated, which allows evaluation of the risk of damage to sensitive diagnostics and hardware.

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.

A correlation electron cyclotron emission (CECE) diagnostic has been used to measure local, turbulent fluctuations of the electron temperature in the core of DIII-D plasmas. This paper describes the hardware and testing of the CECE diagnostic and highlights the importance of measurements of multifield fluctuation profiles for the testing and validation of nonlinear gyrokinetic codes. The process of testing and validating such codes is critical for extrapolation to next-step fusion devices. For the first time, the radial profiles of electron temperature and density fluctuations are compared to nonlinear gyrokinetic simulations. The CECE diagnostic at DIII-D uses correlation radiometry to measure the rms amplitude and spectrum of the electron temperature fluctuations. Gaussian optics are used to produce a poloidal spot size with w(o) approximately 1.75 cm in the plasma. The intermediate frequency filters and the natural linewidth of the EC emission determine the radial resolution of the CECE diagnostic, which can be less than 1 cm. Wavenumbers resolved by the CECE diagnostic are k(theta)or = 1.8 cm(-1) and k(r)or = 4 cm(-1), relevant for studies of long-wavelength turbulence associated with the trapped electron mode and the ion temperature gradient mode. In neutral beam heated L-mode plasmas, core electron temperature fluctuations in the region 0.5r/a0.9, increase with radius from approximately 0.5% to approximately 2%, similar to density fluctuations that are measured simultaneously with beam emission spectroscopy. After incorporating "synthetic diagnostics" to effectively filter the code output, the simulations reproduce the characteristics of the turbulence and transport at one radial location r/a = 0.5, but not at a second location, r/a = 0.75. These results illustrate that measurements of the profiles of multiple fluctuating fields can provide a significant constraint on the turbulence models employed by the code.

Several experiments on different machines and in different fast wave (FW) heating regimes, such as hydrogen minority heating and high harmonic fast waves (HHFW), have found strong interaction between radio-frequency (RF) waves and the scrape-off layer (SOL) region. This paper examines the propagation and the power loss in the SOL by using the full wave code AORSA, in which the edge plasma beyond the last closed flux surface (LCFS) is included in the solution domain and a collisional damping parameter is used as a proxy to represent the real, and most likely nonlinear, damping processes. 2D and 3D AORSA results for the National Spherical Torus eXperiment (NSTX) have shown a strong transition to higher SOL power losses (driven by the RF field) when the FW cut-off is removed from in front of the antenna by increasing the edge density. Here, full wave simulations have been extended for ‘conventional’ tokamaks with higher aspect ratios, such as the DIII-D, Alcator C-Mod, and EAST devices. DIII-D results in HHFW regime show similar behavior found in NSTX and NSTX-U, consistent with previous DIII-D experimental observations. In contrast, a different behavior has been found for C-Mod and EAST, which operate in the minority heating regime.

One modeling framework for integrated tasks (OMFIT) is a comprehensive integrated modeling framework which has been developed to enable physics codes to interact in complicated workflows, and support scientists at all stages of the modeling cycle. The OMFIT development follows a unique bottom-up approach, where the framework design and capabilities organically evolve to support progressive integration of the components that are required to accomplish physics goals of increasing complexity. OMFIT provides a workflow for easily generating full kinetic equilibrium reconstructions that are constrained by magnetic and motional Stark effect measurements, and kinetic profile information that includes fast-ion pressure modeled by a transport code. It was found that magnetic measurements can be used to quantify the amount of anomalous fast-ion diffusion that is present in DIII-D discharges, and provide an estimate that is consistent with what would be needed for transport simulations to match the measured neutron rates. OMFIT was used to streamline edge-stability analyses, and evaluate the effect of resonant magnetic perturbation (RMP) on the pedestal stability, which have been found to be consistent with the experimental observations. The development of a five-dimensional numerical fluid model for estimating the effects of the interaction between magnetohydrodynamic (MHD) and microturbulence, and its systematic verification against analytic models was also supported by the framework. OMFIT was used for optimizing an innovative high-harmonic fast wave system proposed for DIII-D. For a parallel refractive index , the conditions for strong electron-Landau damping were found to be independent of launched and poloidal angle. OMFIT has been the platform of choice for developing a neural-network based approach to efficiently perform a non-linear multivariate regression of local transport fluxes as a function of local dimensionless parameters. Transport predictions for thousands of DIII-D discharges showed excellent agreement with the power balance calculations across the whole plasma radius and over a broad range of operating regimes. Concerning predictive transport simulations, the framework made possible the design and automation of a workflow that enables self-consistent predictions of kinetic profiles and the plasma equilibrium. It is found that the feedback between the transport fluxes and plasma equilibrium can significantly affect the kinetic profiles predictions. Such a rich set of results provide tangible evidence of how bottom-up approaches can potentially provide a fast track to integrated modeling solutions that are functional, cost-effective, and in sync with the research effort of the community.

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.

Note: Proc. 12th Joint Workshop in Electron Cyclotron Emission and Electron Cyclotron Resonance Heating (EC-12), Aix-en-Provence, May 2002, 41 - 46 (2003), ISBN 981-238-189-9 Reference CRPP-CONF-2003-024 Record created on 2008-05-13, modified on 2017-05-12

A new high-power electron cyclotron heating (ECH) system has been introduced on DIII-D. This system is designed to operate at 110 GHz with a total output power of 2 MW. The system consists of four Varian VGT-8011 gyrotrons, (output power of 500 kW), and their associated support equipment. All components have been designed for up to a 10-s pulse duration. The 110-GHz system is intended to further progress in RF current drive experiments on DIII-D when used in conjunction with the existing 60-GHz ECH (1.6-MW) and 30-60-MHz ICH (2-MW) systems. H-mode physics, plasma stabilization experiments, and transport studies are also to be conducted at 110 GHz. The system design philosophy was based on experience gained from the existing 60-GHz ECH system. The consequences of these design decisions are addressed, as is the performance of various 110-GHz components. >

Two 110 GHz gyrotrons with nominal output power of 1 MW each have been installed on the DIII-D tokamak. The first 110 GHz gyrotron built by Gycom has a nominal rating of 1 MW and a 2 s pulse length, with the pulse length being determined by the maximum temperature allowed on the edge cooled boron nitride window. This gyrotron was first operated into the DIII-D tokamak in late 1996. The second gyrotron was built by Communications and Power Industries (CPI) was commissioned during the spring of 1997. The CPI gyrotron uses a double disc FC-75 cooled sapphire window which has a pulse length rating of 0.8 s at 1 MW, 2 s at 0.5 MW and 10 s at 0.2 MW. Both gyrotrons are connected to the tokamak by a low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HE/sub 11/ mode. Using short pulse lengths to avoid breakdown inside the air filled waveguide, the microwave beam has been measured inside the DIII-D vacuum vessel using a paper target and an IR camera. The resultant microwave beam was found to be well focused with a spot size of approximately 8 cm. The beam can be steered poloidially from the center to the outer edge of the plasma. The initial operation of the Gycom gyrotron with about 0.5 MW delivered to a low density plasma for 0.5 s showed good central electron heating, with peak temperature in excess of 10 keV. A third gyrotron being built by CPI will be installed later this year. Progress with the first CPI tube will also be discussed and future plans for the ECH installation and physics experiments will be presented.

Calculation of electron-cyclotron-current drive (ECCD) with the comprehensive CQL3D Fokker-Planck code for a TCV tokamak shot gives 550 kA of driven toroidal current, in marked disagreement with the 100-kA experimental value. Published ECCD efficiencies calculated with CQL3D in the much larger, higher-confinement DIII-D tokamak are in excellent agreement with experiment. The disagreement is resolved by including in the calculations electrostatic-type radial transport at levels given by global energy confinement in tokamaks. The radial transport of energy and toroidal current are in agreement.

Fast waves at frequencies far above the ion cyclotron frequency and approaching the lower hybrid frequency (also called 'helicons' or 'whistlers') have application to off-axis current drive in tokamaks with high electron beta. The high frequency causes the whistler-like behaviour of the wave power nearly following field lines, but with a small radial component, so the waves spiral slowly towards the plasma centre. The high frequency also contributes to strong damping. Modelling predicts robust off-axis current drive with good efficiency compared to alternatives in high performance discharges in DIII-D and Fusion Nuclear Science Facility (FNSF) when the electron beta is above about 1.8%. Detailed analysis of ray behaviour shows that ray trajectories and damping are deterministic (that is, not strongly affected by plasma profiles or initial ray conditions), unlike the chaotic ray behaviour in lower frequency fast wave experiments. Current drive was found to not be sensitive to the launched value of the parallel index of refraction n||, so wave accessibility issues can be reduced. Use of a travelling wave antenna provides a very narrow n||spectrum, which also helps avoid accessibility problems.

Real-time steerable electron cyclotron current drive (ECCD) has been demonstrated to reduce the power requirements and time needed to remove 3/2 and 2/1 neoclassical tearing modes (NTMs) in the DIII-D tokamak. In a world first demonstration of the techniques required in ITER, the island formation onset is detected automatically, gyrotrons are turned on and the real-time steerable ECCD launcher mirrors are moved promptly to drive current at the location of the islands. This shrinks and suppresses the modes well before saturation using real-time motional Stark effect constrained equilibria reconstruction with advanced feedback and search algorithms to target the deposition. In ITER, this method will reduce the ECCD energy requirement and so raise Q by keeping the EC system off when the NTM is not present. Further, in the experiments with accurate tracking of pre-emptive ECCD to resonant surfaces, both 3/2 and 2/1 modes are prevented from appearing with much lower ECCD peak power than required for removal of a saturated mode.

Physics based integrated modelling of the baseline scenario for a Fusion Nuclear Science Facility based on the Advanced Tokamak concept (FNSF-AT) (Chan et al 2010 Fusion Sci. Technol. 57 66) has found steady-state equilibria with good stability and controllability properties at the fusion performance required to accomplish FNSF's nuclear science mission with margin. 2D divertor analysis for this baseline scenario predicts that peak heat flux 1. Two current drive scenarios, two divertor configurations, and two blanket concepts have been analysed. FNSF-AT would complement ITER in addressing science and technology gaps to a commercially attractive DEMO, and could enable a DEMO construction decision triggered by the achievement of Q = 10 in ITER.

The system for controlling neoclassical tearing modes (NTMs) in DIII-D now catches the NTM the moment it becomes unstable by turning on the stabilizing electron cyclotron current drive (ECCD) and promptly bringing it back to stable before it has grown to a large size. Between NTMs, the ECCD can be turned off to save power, which will improve the fusion gain, Q, when used in ITER. This technique, named ?catch and subdue? (CS this is achieved by algorithms that probe how the NTM responds to changes in the alignment. Secondly, the alignment must be maintained even when the NTM is gone so that the ECCD will immediately stabilize when turned on in response to a new NTM. This is made possible by real-time equilibrium reconstructions that include measurements of the motional Stark effect and by a refraction estimator. Thirdly, real-time steerable mirrors are now fast and accurate actuators for the alignment adjustments. Fourthly, early NTM detection is made possible by a real-time mode analysis that filters noise to minimize false positives. These various control elements will be described and followed by a discussion of the further development needed for NTM control on ITER.

Two of the eight neutral-beam sources on the DIII-D tokamak were modified to allow injection below the midplane. To validate off-axis beam performance, the various beams are injected sequentially into low-power plasmas that are optimized for accurate neutron, neutral–particle, fast-ion D-alpha and fast-ion pressure measurements. As expected, the fast-ion profile is broader with off-axis injection than with on-axis injection. The driven toroidal rotation also broadens with off-axis injection and the central fast-ion density is several times smaller. The number of trapped ions in the core depends sensitively on the pitch of the magnetic field lines. Comparisons with classical predictions agree with the measurements for some diagnostics but are discrepant for others.

The first systematic investigation of core electron thermal transport and the role of local ion temperature gradient/trapped electron mode/electron temperature gradient (ITG/TEM/ETG)-scale core turbulence is performed in high temperature, low collisionality H-mode plasmas in the DIII-D tokamak. Wavenumber spectra of L-mode and H-mode density turbulence are measured by Doppler backscattering. H-mode wavenumber spectra are directly contrasted for the first time with nonlinear gyrokinetic simulation results. Core ITG/TEM-scale turbulence is substantially reduced/suppressed by E × B shear promptly after the L–H transition, resulting in reduced electron thermal transport across the entire minor radius. For small k θρs, both experiment and nonlinear gyrokinetic simulations using the GYRO code show density fluctuation levels increasing with k θρs in H-mode (r/a = 0.6), in contrast to ITG/TEM-dominated L-mode plasmas. GYRO simulations also indicate that a significant portion of the remaining H-mode electron heat flux results directly from residual intermediate/short-scale TEM/ETG turbulence. Electron transport at substantially increased electron-to-ion temperature ratio (T e/T i ⩾ 1, r/a ⩽ 0.35) has been investigated in ECH-assisted, quiescent H-mode plasmas. A synergistic increase in core electron and ion thermal diffusivity (normalized to the gyro-Bohm diffusivity) is found with applied ECH. From linear stability analysis, the TEM mode is expected to become the dominant linear instability with ECH due to increased electron-to-ion temperature ratio and a reduction in the ion temperature gradient. This is consistent with increased electron temperature fluctuations and core electron thermal diffusivity observed experimentally. The reduced ion temperature gradient likely results from a reduction in the ITG critical gradient due to increased T e/T i and reduced E × B shear. These studies are performed at collisonality ( , r/a ⩽ 0.6) and address transport in electron heat-dominated regimes, thought to be important in ITER due to α-particle heating.

Recent progress on ITER steady-state (SS) scenario modelling by the ITPA-IOS group is reviewed. Code-to-code benchmarks as the IOS group's common activities for the two SS scenarios (weak shear scenario and internal transport barrier scenario) are discussed in terms of transport, kinetic profiles, and heating and current drive (CD) sources using various transport codes. Weak magnetic shear scenarios integrate the plasma core and edge by combining a theory-based transport model (GLF23) with scaled experimental boundary profiles. The edge profiles (at normalized radius ρ = 0.8–1.0) are adopted from an edge-localized mode-averaged analysis of a DIII-D ITER demonstration discharge. A fully noninductive SS scenario is achieved with fusion gain Q = 4.3, noninductive fraction f NI = 100%, bootstrap current fraction f BS = 63% and normalized beta βN = 2.7 at plasma current I p = 8 MA and toroidal field B T = 5.3 T using ITER day-1 heating and CD capability. Substantial uncertainties come from outside the radius of setting the boundary conditions (ρ = 0.8). The present simulation assumed that βN (ρ) at the top of the pedestal (ρ = 0.91) is about 25% above the peeling–ballooning threshold. ITER will have a challenge to achieve the boundary, considering different operating conditions (T e/T i ≈ 1 and density peaking). Overall, the experimentally scaled edge is an optimistic side of the prediction. A number of SS scenarios with different heating and CD mixes in a wide range of conditions were explored by exploiting the weak-shear steady-state solution procedure with the GLF23 transport model and the scaled experimental edge. The results are also presented in the operation space for DT neutron power versus stationary burn pulse duration with assumed poloidal flux availability at the beginning of stationary burn, indicating that the long pulse operation goal (3000 s) at I p = 9 MA is possible. Source calculations in these simulations have been revised for electron cyclotron current drive including parallel momentum conservation effects and for neutral beam current drive with finite orbit and magnetic pitch effects.

A fusion development facility (FDF) based on the tokamak approach with normal conducting magnetic field coils is presented. FDF is envisioned as a facility with the dual objective of carrying forward advanced tokamak (AT) physics and enabling the development of fusion energy applications. AT physics enables the design of a compact steady-state machine of moderate gain that can provide the neutron fluence required for FDF's nuclear science development objective. A compact device offers a uniquely viable path for research and development in closing the fusion fuel cycle because of the demand to consume only a moderate quantity of the limited supply of tritium fuel before the technology is in hand for breeding tritium.

A series of carefully designed experiments on DIII-D have taken advantage of a broad set of turbulence and profile diagnostics to rigorously test gyrokinetic turbulence simulations. In this paper the goals, tools and experiments performed in these validation studies are reviewed and specific examples presented. It is found that predictions of transport and fluctuation levels in the mid-core region (0.4 < ρ < 0.75) are in better agreement with experiment than those in the outer region (ρ ⩾ 0.75) where edge coupling effects may become increasingly important and multiscale simulations may also be necessary. Validation studies such as these are crucial in developing confidence in a first-principles based predictive capability for ITER.

Two techniques were developed at DIII-D [J. L. Luxon, Nucl. Fusion 42, 64 (2002)] to tackle ITER-specific aspects of neoclassical tearing mode (NTM) control, namely, (1) the relatively small size of the rotating islands, smaller than the electron cyclotron current drive (ECCD) deposition region, and (2) the increased tendency of the islands, compared to present devices, to lock to the wall or to the residual error field, in a position not necessarily accessible to ECCD. Modulated ECCD is known to suppress small islands more efficiently, when “broad,” than continuous ECCD. At DIII-D, a NTM of poloidal/toroidal mode numbers m/n=3/2 was completely stabilized by a new technique where oblique electron cyclotron emission acted at the same time as an indicator of good alignment between ECCD and the island, and as a waveform generator, for modulation in synch and in phase with the island O-point. In another experiment, after locking in an unfavorable position, a 2/1 island was steered by externally generated magn...

Neutral beam injection into reversed magnetic shear DIII-D plasmas produces a variety of Alfvénic activity including toroidicity and ellipticity induced Alfvén eigenmodes (TAE/EAE, respectively) and reversed shear Alfvén eigenmodes (RSAE) as well as their spatial coupling. These modes are studied during the discharge current ramp phase when incomplete current penetration results in a high central safety factor and strong drive due to multiple higher order resonances. It is found that ideal MHD modelling of eigenmode spectral evolution, coupling and structure are in excellent agreement with experimental measurements. It is also found that higher radial envelope harmonic RSAEs are clearly observed and agree with modelling. Some discrepancies with modelling such as that due to up/down eigenmode asymmetries are also pointed out. Concomitant with the Alfvénic activity, fast ion (FIDA) spectroscopy shows large reductions in the central fast ion profile, the degree of which depends on the Alfvén eigenmode amplitude. Interestingly, localized electron cyclotron heating (ECH) near the mode location stabilizes RSAE activity and results in significantly improved fast ion confinement relative to discharges with ECH deposition on axis. In these discharges, RSAE activity is suppressed when ECH is deposited near the radius of the shear reversal point and enhanced with deposition near the axis. The sensitivity of this effect to deposition power and current drive phasing as well as ECH modulation are presented. © 2009 IAEA, Vienna.

For the first time, profiles (0.3

The role of interchange and internal kink modes in the sawtooth oscillations is explored by comparing bean- and oval-shaped plasmas. The n=1 instability that results in the collapse of the sawtooth has been identified as a quasi-interchange in the oval cases and the internal kink in the bean shape. The ion and electron temperature profiles are followed in detail through the sawtooth ramp. It is found that electron energy transport rates are very high in the oval and quite low in the bean shape. Ion energy confinement in the oval is excellent and the sawtooth amplitude (δT∕T) in the ion temperature is much larger than that of the electrons. The sawtooth amplitudes for ions and electrons are comparable in the bean shape. The measured q profiles in the bean and oval shapes are found to be consistent with neoclassical current diffusion of the toroidal current, and the observed differences in q largely result from the severe differences in electron energy transport. For both shapes the collapse flattens the q ...

The grating spectrometer installation on the DIII‐D tokamak uses fundamental circular waveguide propagating the TE11 lowest‐order mode followed by oversized circular guide carrying the low‐loss TE01 mode. The short section of fundamental guide permits use of an electronic chopper operating at 100 kHz for both calibration and plasma operation. By using ac‐coupled amplifiers tuned to the chopping frequency, the background signal generated in the indium antimonide detectors by neutrons and x rays is automatically subtracted and the system noise bandwidth is reduced. Compared with a quasi‐optical system, the much smaller fundamental horn and front‐end waveguide allow the waveguide system to be located outside a gate valve. With this configuration the entire waveguide run, including the actual horn and vacuum window used during plasma operations, can be included in the calibration setup.