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High-power plasma waves in the lower hybrid range of frequencies (LHRF) are being used for non-inductive current drive in experiments around the world. In this range, plasmas can support waves with two different polarizations, a quasi-electrostatic wave known as the “slow wave” or simply the lower hybrid wave, and a quasi-electromagnetic wave referred to variously as the “fast wave in the LHRF”, “whistler”, or “helicon”. An improved qualitative understanding of how fast and slow waves in the LHRF propagate can be gained by studying the “inverse wave normal surface” or IWNS. It is shown that the two branches propagate in a qualitatively similar way, with, however, important quantitative differences.

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.

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.

Combined neutral beam injection and fast wave heating at the fourth and fifth cyclotron harmonics accelerate fast ions in the DIII-D tokamak. Measurements with a nine-channel fast-ion D-alpha (FIDA) diagnostic indicate the formation of a fast-ion tail above the injection energy. Tail formation correlates with enhancement of the d–d neutron rate above the value that is expected in the absence of fast-wave acceleration. FIDA spatial profiles and fast-ion pressure profiles inferred from the equilibrium both indicate that the acceleration is near the magnetic axis for a centrally located resonance layer. The enhancement is largest 8–10 cm beyond the radius where the wave frequency equals the cyclotron harmonic, probably due to a combination of Doppler-shift and orbital effects. The fast-ion distribution function calculated by the CQL3D Fokker– Planck code is fairly consistent with the data. (Some figures in this article are in colour only in the electronic version)

We compare, qualitatively and with a simple semi-quantitative model, the difficulty of coupling to fast waves at 0.5 GHz (’whistlers’ or ’helicons’) and to slow waves (’lower hybrid waves’) at 5 GHz and n||=3 under conditions relevant to the DIII-D tokamak. Though the density at which the slow wave begins to propagate is much lower than the fast wave, the stronger evanescence in vacuum for the higher frequency wave makes excitation of the SWs more difficult than for the FW case. Enforcing a minimum density at the coupler face, as might be produced by local gas puffing, is shown to likely be more effective for the SW case. The fact that the surface admittance for the slow wave at 0.5 GHz is much larger than for the FW at the same frequency has possibly significant implications for attempts to excite the FW.

The advanced tokamak programme on DIII-D is to develop the scientific basis for steady state, high performance operation in future devices. We report on experiments attempting to demonstrate sustainment of 100% noninductive current for several seconds at high beta, using up to 2.5 MW of off-axis electron cyclotron current drive and up to 15 MW of neutral beam injection with q95 ≈ 5. A 100% noninductive condition was achieved with βT = 3.6%, βN = 3.5, H89 = 2.4 and improved current drive alignment. However, the duration of this phase was limited by the pressure profile evolution leading to magnetohydrodynamic instabilities after about 0.7 s. In a separate discharge, a nearly (~90%) noninductive, stationary condition was maintained for one current relaxation time (1.8 s) only limited by the duration of the hardware system. These experiments have achieved normalized fusion performance with bootstrap current fraction fBS ≈ 60%, consistent with requirements for the ITER Q = 5 steady-state scenarios. The modelling tools that were successfully employed to devise experiments in DIII-D are applied to ITER, indicating that full noninductive operation is plausible for an ITER steady-state scenario with Q ≈ 5.

The major objective of the National Spherical Torus Experiment (NSTX) is to understand basic toroidal confinement physics at low aspect ratio and high βT in order to advance the spherical torus (ST) concept. In order to do this, NSTX utilizes up to 7.5 MW of neutral beam injection, up to 6 MW of high harmonic fast waves (HHFWs), and it operates with plasma currents up to 1.5 MA and elongations of up to 2.6 at a toroidal field up to 0.45 T. New facility, and diagnostic and modelling capabilities developed over the past two years have enabled the NSTX research team to make significant progress towards establishing this physics basis for future ST devices. Improvements in plasma control have led to more routine operation at high elongation and high βT (up to ~40%) lasting for many energy confinement times. βT can be limited by either internal or external modes. The installation of an active error field (EF) correction coil pair has expanded the operating regime at low density and has allowed for initial resonant EF amplification experiments. The determination of the confinement and transport properties of NSTX plasmas has benefitted greatly from the implementation of higher spatial resolution kinetic diagnostics. The parametric variation of confinement is similar to that at conventional aspect ratio but with values enhanced relative to those determined from conventional aspect ratio scalings and with a BT dependence. The transport is highly dependent on details of both the flow and magnetic shear. Core turbulence was measured for the first time in an ST through correlation reflectometry. Non-inductive start-up has been explored using PF-only and transient co-axial helicity injection techniques, resulting in up to 140 kA of toroidal current generated by the latter technique. Calculated bootstrap and beam-driven currents have sustained up to 60% of the flat-top plasma current in NBI discharges. Studies of HHFW absorption have indicated parametric decay of the wave and associated edge thermal ion heating. Energetic particle modes, most notably toroidal Alfven eigenmodes and fishbone-like modes result in fast particle losses, and these instabilities may affect fast ion confinement on devices such as ITER. Finally, a variety of techniques has been developed for fuelling and power and particle control.

Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with ?N ? 3.5 and ?T ? 3.6% sustained for up to 1?s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by results of these and other recent experiment and simulation efforts.

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these condi...

Energy confinement comparable with tokamak quality is achieved in the Madison Symmetric Torus (MST) reversed field pinch (RFP) at a high beta and low toroidal magnetic field. Magnetic fluctuations normally present in the RFP are reduced via parallel current drive in the outer region of the plasma. In response, the electron temperature nearly triples and beta doubles. The confinement time increases ten-fold (to ~10 ms), which is comparable with L- and H-mode scaling values for a tokamak with the same plasma current, density, heating power, size and shape. Runaway electron confinement is evidenced by a 100-fold increase in hard x-ray bremsstrahlung. Fokker–Planck modelling of the x-ray energy spectrum reveals that the high energy electron diffusion is independent of the parallel velocity, uncharacteristic of magnetic transport and more like that for electrostatic turbulence. The high core electron temperature correlates strongly with a broadband reduction of resonant modes at mid-radius where the stochasticity is normally most intense. To extend profile control and add auxiliary heating, rf current drive and neutral beam heating are in development. Low power lower-hybrid and electron Bernstein wave injection experiments are underway. Dc current sustainment via ac helicity injection (sinusoidal inductive loop voltages) is also being tested. Low power neutral beam injection shows that fast ions are well-confined, even in the presence of relatively large magnetic fluctuations.

The quiescent double barrier (QDB) regime is a high performance regime recently identified in DIII-D and characterized by a double transport barrier structure (core and edge) that can be maintained for several seconds, often limited only by the pulse length capabilities of the DIII-D hardware. The QDB regime has been sustained for up to 25 τE with fusion performance of up to βNH89≈7. The edge barrier is ELM-free, but modulated by low frequency MHD activity that allows density control via an external cryopump. The core barrier is similar to those seen in previous internal transport barrier experiments, but is maintained without complete stabilization of turbulence. Instead, the turbulence correlation lengths become very short so as to minimize the transport length scales. The two barriers are separated by a region of high transport that is a consequence of a zero-crossing in the E×B shearing rate. These discharges typically possess highly peaked density profiles. This has several implications: narrow bootstrap current profile, reduced beta limit and increased impurity retention. We will report on studies of each of these issues.

Modeling work shows that current drive can be done off-axis with high efficiency, as required for FNSF and DEMO, by using very high harmonic fast waves (“helicons” or “whistlers”). The modeling indicates that plasmas with high electron beta are needed in order for the current drive to take place off-axis, making DIII-D a highly suitable test vehicle for this process. The calculations show that the driven current is not very sensitive to the launched value of n∥, a result that can be understood from examination of the evolution of n∥ as the waves propagate in the plasma. Because of this insensitivity, relatively large values (∼3) of n∥ can be launched, thereby avoiding some of the problems with mode conversion in the boundary found in some previous experiments. Use of a traveling wave antenna provides a very narrow n∥ spectrum, which also helps avoid mode conversion.

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.

The main aim of the National Spherical Torus Experiment (NSTX) is to establish the fusion physics principles of the spherical torus (ST) concept. The NSTX device began plasma operations in February 1999 and the plasma current Ip was successfully brought up to the design value of 1 MA on 14 December 1999. The planned plasma shaping parameters, elongation κ = 1.6-2.2 and triangularity δ = 0.2-0.4, were achieved in inner wall limited, and single null and double null diverted configurations. The coaxial helicity injection (CHI) and high harmonic fast wave (HHFW) experiments were also initiated. CHI current of 27 kA produced up to 260 kA toroidal current without using an ohmic solenoid. With the injection of 2.3 MW of HHFW power, using 12 antennas connected to six transmitters, electrons were heated from a central temperature of 400 eV to 900 eV at a central density of 3.5 × 1013 cm-3, increasing the plasma energy to 59 kJ and the toroidal β, βT, to 10%. The NBI system commenced operation in September 2000. The initial results with two ion sources (PNBI = 2.8 MW) show good heating, producing a total plasma stored energy of 90 kJ corresponding to βT ≈ 18% at a plasma current of 1.1 MA.

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.

Significant progress has been made in obtaining high-performance discharges for many energy confinement times in the DIII-D tokamak [J. L. Luxon et al., Plasma Physics and Controlled Fusion Research (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159]. Normalized performance (measured by the product of βNH89 and indicative of the proximity to both conventional β limits and energy confinement quality, respectively) ∼10 has been sustained for >5 τE with qmin>1.5. These edge localized modes (ELMing) H-mode discharges have β∼5%, which is limited by the onset of resistive wall modes slightly above the ideal no-wall n=1 limit, with approximately 75% of the current driven noninductively. The remaining Ohmic current is localized near the half-radius. The DIII-D electron cyclotron heating system is being upgraded to replace this inductively driven current with localized electron cyclotron current drive (ECCD). Density control, which is required for effective ECCD, has been successfully demonstrated in long-pulse high-performance ELMing H-mode discharges with βNH89∼7 for up to 6.3 s. In plasma shapes compatible with good density control in the present divertor configuration, the achieved βN is somewhat less than that in the high βNH89=10 discharges.

The mission of the National Spherical Torus Experiment (NSTX) is to extend the understanding of toroidal physics to low aspect ratio (R/a approximately equal to 1.25) in low collisionality regimes. NSTX is designed to operate with up to 6 MW of High Harmonic Fast Wave (HHFW) heating and current drive, 5 MW of Neutral Beam Injection (NBI) and Co-Axial Helicity Injection (CHI) for non-inductive startup. Initial experiments focused on establishing conditions that will allow NSTX to achieve its aims of simultaneous high-bt and high-bootstrap current fraction, and to develop methods for non-inductive operation, which will be necessary for Spherical Torus power plants. Ohmic discharges with plasma currents up to 1 MA and with a range of shapes and configurations were produced. Density limits in deuterium and helium reached 80% and 120% of the Greenwald limit respectively. Significant electron heating was observed with up to 2.3 MW of HHFW. Up to 270 kA of toroidal current for up to 200 msec was produced noninductively using CHI. Initial NBI experiments were carried out with up to two beam sources (3.2 MW). Plasmas with stored energies of up to 140 kJ and bt =21% were produced.

Transport phenomena are studied in Advanced Tokamak (AT) regimes in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research, 1986 (International Atomics Energy Agency, Vienna, 1987), Vol. I, p. 159], with the goal of developing understanding and control during each of three phases: Formation of the internal transport barrier (ITB) with counter neutral beam injection taking place when the heating power exceeds a threshold value of about 9 MW, contrasting to co-NBI injection, where Pthreshold

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.

All high-power ion cyclotron range of frequency (ICRF) heating systems include devices for matching the input impedance of the antenna array to the generator output impedance. For most types of antennas used, the input impedance is strongly time dependent on timescales as rapid as 10-4 s, while the radio frequency (RF) generators used are capable of producing full power only into a stationary load impedance. Hence, the dynamic response of the matching method is of great practical importance. In this paper, world-wide developments in this field over the past decade are reviewed. These techniques may be divided into several classes. The edge plasma parameters that determine the antenna array's input impedance may be controlled to maintain a fixed load impedance. The frequency of the RF source can be feedback controlled to compensate for changes in the edge plasma conditions, or fast variable tuning elements in the transmission line between the generator output and the antenna input connections can provide the necessary time-varying impedance transformation. In `lossy passive schemes', reflected power due to the time-varying impedance of the antenna array is diverted to a dummy load. Each of these techniques can be applied to a pre-existing antenna system. If a new antenna is to be designed, recent advances allow the antenna array to have the intrinsic property of presenting a constant load to the feeding transmission lines despite the varying load seen by each antenna in the array.

Recent progress in the area of ICRF physics is examined and summarized. A very brief summary of simplified theoretical predictions is given. Illustrative examples of experimental results in the area of ICRF heating and fast-wave current drive are given. ICRF heating is found to be highly effective in heating plasmas to high temperatures in modern tokamaks. Fast-wave current drive has been demonstrated in several experiments and the results are in good agreement with theory. Mode-conversion heating and current drive should be effective for profile control, but experimental results are still preliminary.

The Fast Wave Current Drive System is used to heat and drive non-inductive current in the DIII-D tokamak plasma. The phase and power of the transmitted and reflected RF wave transferred to the antenna are measured at different points in the transmission lines. These measurements are extremely critical to help tune the system to transfer maximum power to the plasma and to provide fast interlocks to protect the transmitter and the transmission line. Currently there are two systems at DIII-D to perform these functions. One system is used for measuring amplitude and phase of the RF signals for initial tuning of the transmission line before start-up and during plasma shots. The second system is used for arc protection. While this system worked well, it was expensive to maintain and offered only fixed functionality. Furthermore, a higher sampling speed would allow higher resolution of the data used for post-shot analysis. Additionally a microsecond level interlock system could have an improved response time. The National Instruments FPGA-based FlexRIO board, digitizer module, and PXI Express platform were chosen to perform both high-accuracy measurements and for interlocks. This commercially available system offers faster ADCs and an on-board FPGA to perform real-time signal processing without suffering the latency of sending the data to a CPU. This allowed the system to achieve microsecond-scale response time and provides the ability for the users to modify the functionality through the user programmable FPGA. This paper will detail the system architecture, performance comparison between the original and new systems, and preliminary testing of the interlock system.

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...

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 noninductive current drive concept, based on internal pressure‐driven currents in a low‐aspect‐ratio toroidal geometry, has been demonstrated on the Current Drive Experiment Upgrade (CDX‐U) [Forest et al., Phys. Rev. Lett. 68, 3559 (1992)] and further tested on DIII‐D [in Plasma Physics and Controlled Nuclear Fusion Research, 1986, Proceedings of the 11th International Conference, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159]. For both experiments, electron cyclotron power provided the necessary heating to breakdown and maintain a plasma with high‐βp and low collisionality (eβp∼1, ν*≤1). A poloidal vacuum field similar to a simple magnetic mirror is superimposed on a much stronger toroidal field to provide the initial confinement for a hot, trapped electron species. With application of electron cyclotron heating (ECH), toroidal currents spontaneously flow within the plasma and increase with applied ECH power. The direction of the generated current is independent of the toroid...

A phasing and matching system has been designed and implemented for the four-element fast wave current drive (FWCD) antenna array on DIII-D, This system permits phased operation using a single transmitter. Coupled power levels of 1.1 Mw have been reached with relative phasing of ±π/2 and equal magnitudes of current and voltage on all four lines. Use of a single power source requires the achievement of amplitude and phase control at high power, without feedback control of these quantities. The system uses only standard components consisting of transmission lines, unmatched tees, and manually controlled phase shifters and stubs. Phasing, matching, and amplitude control for all four current straps are done using a total of five tuning elements. This simplification is achieved through the use of two resonant loops, each connecting a pair of straps. A tuning algorithm developed for the system produces accurate matching, phasing, and amplitude balance within a small number of shots (≤ 5) in cases where the loading is sufficiently high, that is, when the resonant series load resistance (RSLR) > 2 ω, at values of kQ approaching 1. A coupled transmission line model of the antenna array and resonant loops has been created and used to determine changes in resistive and reactive loading, as well as changes in coupling between array elements during plasma shots. The design and modeling of this system and the operating experience are reviewed.

In recent moderate to high harmonic fast wave heating and current drive experiments in DIII‐D, a synergy effect was observed when the 6th harmonic 90 MHz fast wave power is applied to the plasma preheated by neutral beams and the 4th harmonic 60 MHz fast wave. In this paper, we investigate how the synergy can occur using ORBIT‐RF coupled with AORSA. Preliminary simulations suggest that damping of 4th harmonic FW on beam ions accelerates them above the injection energy, which may allow significant damping of 6th harmonic FW on beam ion tails to produce synergy.

The single‐feed, end‐grounded straps of the NSTX 12‐strap HHFW antenna array have been replaced with double‐feed, center‐grounded straps to reduce the voltages in the vicinity of the Faraday shield (FS) for a given strap current. The strap spacings to the FS and to the back plate were increased by 3 mm to decrease the electric fields for a given voltage. The electric fields near the FS have been roughly halved for the same strap currents, permitting a direct examination of the roles that internal fields play in determining antenna power limits in plasmas. Extensive RF/plasma conditioning of the antenna was required to remove enough of the evaporated Li deposits from prior wall conditioning to permit coupling in excess of 4 MW to L‐ and H‐mode plasmas in 2009. Most arcs were associated with expulsion of Li from the FS/antenna frame surfaces. The center‐grounded straps were less susceptible to arcing during ELMing H‐mode plasmas. Reliable operation above 2 MW was difficult after the installation of the Liqu...

Fast Waves (FWs) at 60 MHz and 90 MHz are used in DIII‐D for central electron heating and current drive. Coupling of FWs to high‐performance discharges is limited by low antenna loading in these regimes. To extend the application of high‐power FWs to such regimes, methods of increasing the antenna loading in these regimes are needed. A systematic study of loading enhancement techniques has been carried out in DIII‐D, including reduction of the antenna∕plasma distance, gas puffing into the far scrape‐off layer (SOL), and control of other parameters that affect the particle balance in the far SOL. Quantitative understanding of the physics of the loading resistance and its dependence on edge density profiles is demonstrated. The core FW heating efficiency appeared to be ∼100% in the Advanced Inductive regime, consistent with the high first‐pass direct electron absorption of ∼75% that is predicted by the ray‐tracing code GENRAY in this high electron beta regime., United States. Dept. of Energy (Grant DE-FC02-04ER54698), United States. Dept. of Energy (Grant DE-FG02-940ER54084), United States. Dept. of Energy (Grant DE-AC05-00OR22725), United States. Dept. of Energy (Grant DE-AC02-09CH11466), United States. Dept. of Energy (Grant DE-FG02-08ER54984)

Fast waves (FWs) at 60 MHz and at 90 MHz are coupled to DIII‐D discharges for central heating and current drive at net FW power up to 3.5 MW. The primary absorption mechanism is intended to be direct electron damping in the plasma core. In discharges at B = 2 T with fast deuteron populations from neutral beam injection, 4th and 6th deuterium cyclotron harmonic absorption on the fast ions competes with direct electron damping. Previous experiments have shown that the 6ΩD absorption of the 90 MHz FWs is weaker than the 4ΩD absorption of 60 MHz FWs, in agreement with a model that includes unspecified edge losses. Recent experiments have shown that if the fast deuterons are accelerated by absorption of 60 MHz (4ΩD) FWs, adding 90 MHz power (6ΩD) can increase the fusion neutron rate by a larger increment than is obtained with 90 MHz power alone. Details of this synergy between 4ΩD and 6ΩD absorption are presented.

Combined neutral beam injection and fast wave heating at cyclotron harmonics accelerate deuterium fast ions in the National Spherical Torus Experiment (NSTX) and in the DIII‐D tokamak. Acceleration above the injected energy is evident in fast‐ion D‐alpha (FIDA) and volume‐average neutron data. The FIDA diagnostic measures spatial profiles of the accelerated fast ions. In DIII‐D, the acceleration is at a 4th or 5th cyclotron harmonic; the maximum enhancement in the high‐energy FIDA signal is 8–10 cm beyond the resonance layer. In NSTX, acceleration is observed at five harmonics (7–11) simultaneously; overall, the profile of accelerated fast ions is much broader than in DIII‐D. The energy distribution predicted by the CQL3D Fokker‐Planck code agrees fairly well with measurements in DIII‐D. However, the predicted profiles differ from experiment, presumably because the current version of CQL3D uses a zero‐banana‐width model.

A goal in the 5‐year plan for the fast wave program on DIII‐D is to couple a total of 3.6 MW of RF power into a long pulse, H‐mode plasma for central electron heating. The present short‐pulse 285/300 antenna array would need to be replaced with one capable of at least 1.2 MW, 10 s operation at 60 MHz into an H‐mode (low resistive loading) plasma condition. The primary design under consideration uses a poloidally‐segmented strap (3 sections) for reduced strap voltage near the plasma/Faraday screen region. Internal capacitance makes the antenna structure self‐resonant at 60 MHz, strongly reducing peak E‐fields in the vacuum coax and feed throughs.

A convenient method of tuning resonant loops has been developed for use on the Long‐Pulse Fast Wave Current Drive antennas on DIII‐D. These 4‐element antennas are fed from a single RF source, with a 90° hybrid junction feeding two resonant loops. Each resonant loop connects a pair of non‐adjacent current straps, and a decoupler allows independent impedance matching of each output leg of the hybrid junction. The tuning method involves only measurement of the reflection coefficient of each resonant loop and the transmission from one loop to the other. It will be shown that this method of tuning equalizes the currents in all four current straps and 90° phasing between adjacent elements.

We evaluate the accuracy of the Porcelli sawtooth model using more realistic numerical models from the ORBIT‐RF and GATO codes in DIII‐D fast wave heating experiments. Simulation results confirm that the fast wave‐induced energetic trapped particles may stabilize the sawtooth instability. The crucial kinetic stabilizing contribution strongly depends on both the experimentally reconstructed magnetic shear at the q = 1 surface and the calculated poloidal beta of energetic trapped particles inside the q = 1 surface.

Recent DIII‐D experiments on the absorption of fast waves (FWs) by moderate to high ion cyclotron harmonic damping [R. I. Pinsker, et al., Nucl. Fusion 46, S416‐S424 (2006)] have motivated a substantial modeling effort within the RF SciDAC community, but as of yet, the time dependence of the plasma response has not been modeled. In this work, a simple 0‐D time‐dependent absorption and transport model is used to predict the plasma response to a step in FW power when multiple FW absorption mechanisms are present. When the plasma has a nonlinear response to at least one of the damping mechanisms, both the partition between the various absorption channels in the steady‐state solution and the time required to reach the final state depend strongly on the initial conditions.

Advanced Tokamak research in DIII-D seeks to develop a scientific basis for steady-state high-performance tokamak operation. Fully noninductive (fNI≈ 100%) in-principle steady-state discharges have been maintained for several confinement times. These plasmas have weak negative central shear with qmin≈ 1.5-2, βN≈ 3.5, and large, well-aligned bootstrap current. The loop voltage is near zero across the entire profile. The remaining current is provided by neutral beam current drive (NBCD) and electron cyclotron current drive (ECCD). Similar plasmas are stationary with fNI≈ 90-95% and duration up to 2 s, limited only by hardware. In other experiments, βN≈ 4 is maintained for 2 s with internal transport barriers, exceeding previously achieved performance under similar conditions. This is allowed by broadened profiles and active magnetohydrodynamic instability control. Modifications now underway on DIII-D are expected to allow extension of these results to higher performance and longer duration. A new pumped divertor will allow density control in high triangularity double-null divertor configurations, facilitating access to similar in-principle steady-state regimes with βN> 4. Additional current drive capabilities, both off-axis ECCD and on-axis fast wave current drive (FWCD), will increase the magnitude, duration, and flexibility of externally driven current. © 2006 Elsevier B.V. All rights reserved.

The DIII-D ECH system includes three launcher assemblies each of which can accommodate the RF beams from two gyrotrons. The launchers use four different designs for the mirrors which focus and direct the beams into the tokamak. The designs use molybdenum brazed to graphite, thin Glidcop or variable thickness Glidcop. A fourth design with laminated Glidcop/stainless steel construction has been operated, but no thermal data are available. All the mirrors operate without active cooling. This paper presents preliminary analyses and measurements of the thermal performance of the three designs for which data have been obtained.

The theory originally developed for the MST EBW waveguide coupling experiments is used to model coupling to the EBW in the National Spherical Torus eXperiment (NSTX).

In ECH H‐mode discharges with essentially no toroidal momentum input, counter toroidal rotation is measured in the interior, ρ < 0.5, while co rotation is measured outside this region. This is contrasted with Ohmic H‐modes where the rotation is everywhere in the co direction. A simple parameterization is suggested to facilitate comparison between machines.

The Fast Wave Current Drive (FWCD) system at DIII-D has increased its available radio frequency (RF) power capabilities with the addition of two new high power transmitters along with their associated transmission line systems. A Sun Sparc-10 workstation, functioning as the FWCD operator console, is being used to control transmitter operating parameters and transmission line tuning parameters, along with acquiring data and making data available for integration into the DIII-D data acquisition system. Labview, a graphical user interface application, is used to manage and control the above processes. This paper will discuss the three primary branches of the FWCD computer control system: transmitter control, transmission line tuning control, and FWCD data acquisition. The main control program developed uses VXI, GPIB, CAMAC, Serial, and Ethernet protocols to blend the three branches together into one cohesive system. The control of the transmitters utilizes VXI technology to communicate with the transmitter's digital interface. A GPIB network allows for communication with various instruments and CAMAC crate controllers. CAMAC crates are located at each phase-shifter/stub-tuner station and are used to digitize transmission line parameters along with transmission line fault detection during RF transmission. The phase-shifter/stub-tuner stations are located through out the DIII-D facility and are controlled from the FWCD operator console via the workstation's Serial port. The Sun workstation has an Ethernet connection allowing for the utilization of the DIII-D data acquisition "Open System" architecture and of course providing communication with the rest of the world.

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.

Radio-frequency power at 30 MHz has been applied in a variety of situations to National Spherical Torus Experiment (NSTX) plasmas. The response of the plasma is observed in order to study both the physics of High Harmonic Fast Wave (HHFW) heating and as a tool to extend the performance of NSTX plasmas. In this paper we will discuss the progress made to date towards these goals.

The 110 GHz system on the DIII-D tokamak comprises three Gycom Centaur class gyrotrons producing 750 kW for 2.0 sec pulses and one CPI VGT8110 series gyrotron producing 800 kW for 2.0 ec pulses. Total injected power is in excess of 2 MW. An additional two CPI gyrotrons are being prepared for operation following failure of their diamond output windows and these failures have resulted in operational limits on power and pulse length for the operating CPI gyrotron. The system has poloidally steerable launchers for four gyrotrons, two with oblique launch for current drive and two with perpendicular launch for heating only. A fully articulating launcher with poloidal scan and ±19 deg toroidal scan is also operational. Radiatively cooled mirrors capable of 1 MW for 10 s pulses at 1% duty cycle are being tested on the oblique launcher. New dummy loads based on conversion of the HE1,1 fundamental mode to surface modes and having fast time response and 80% absorption of the incident rf are now in service. A new sin...

Conventional electron cyclotron heating using the O- and X-modes to carry energy from the plasma edge to the cyclotron resonance layer is not possible for high density, low magnetic field devices (RFPs and STs, for example), since these modes are evanescent in most of the plasma. As an alternative, we consider coupling to the electron Bernstein wave (EBW) with a single waveguide, the mouth of which is inserted to the vicinity of the upper hybrid resonance.

Direct observation of the generation of sheared ExB flows via Ion Bernstein Waves (IBW) at the fourth ion cyclotron harmonic is reported in the toroidal magnetized plasma Thorello. The measured ExB shearing rate from IBW induced Reynolds stress increases with increasing RF power. The experimental results show a decrease of the radial correlation length in the region of the shear layer.

Combined neutral beam injection and fast wave heating at the fourth cyclotron harmonic produce an energetic deuterium beam ion tail in the DIII-D tokamak. When the concentration of thermal hydrogen exceeds approx. 5%, the beam ion absorption is suppressed in favour of second harmonic hydrogen absorption. As theoretically expected, the beam absorption increases with beam ion gyroradius; also, central absorption at the fifth harmonic is weaker than central absorption at the fourth harmonic. For central heating at the fourth harmonic, an energetic, perpendicular, beam population forms inside the q = 1 surface. The beam ion tail transiently stabilizes the sawtooth instability but destabilizes toroidicity induced Alfven eigenmodes (TAEs). Saturation of the central heating correlates with the onset of the TAEs. Continued expansion of the q = 1 radius eventually precipitates a sawtooth crash; complete magnetic reconnection is observed.