The National Spherical Torus Experiment Upgrade (NSTX-U) will operate at axial toroidal fields of <= 1 T and plasma currents, I-p <= 2 MA. The development of non-inductive (NI) plasmas is a major long-term research goal for NSTX-U. Time dependent numerical simulations of 28 GHz electron cyclotron (EC) heating of low density NI start-up plasmas generated by Coaxial Helicity Injection (CHI) in NSTX-U predict a significant and rapid increase of the central electron temperature (T-e(0)) before the plasma becomes overdense. The increased T-e(0) will significantly reduce the I-p decay rate of CHI plasmas, allowing the coupling of fast wave heating and neutral beam injection. A megawatt-level, 28 GHz electron heating system is planned for heating NI start-up plasmas in NSTX-U. In addition to EC heating of CHI start-up discharges, this system will be used for electron Bernstein wave (EBW) plasma start-up, and eventually for EBW heating and current drive during the I-p flattop.

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)-a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (>= 6.5 T, 1.5 MA), high power density facility (P/S similar to 1.5 MW m(-2)) will test innovative divertor ideas, including an 'X-point target divertor' concept, at the required performance parameters-reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region-while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magneticfield side-the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination-advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions-will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv: 1409.3540)) that makes use of high-temperature superconductor technology-a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental observation of ion cyclotron range of frequency (ICRF) mode-conversion, ICRF flow drive, demonstration of lower-hybrid current drive at ITER-like densities and fields and, using a set of novel diagnostics, extensive validation of advanced RF codes. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. A summary of important achievements and discoveries are included.

This paper describes the operation of a double stub fast ferrite tuner (FFT) that we have designed for the Alcator C-Mod 4.6-GHz lower hybrid current drive system. This FFT is unique because it uses a single electromagnet coil and permanent magnet on each tuning stub. The ferrite is located on the center of the broad face of the waveguide. The FFT is required to withstand over 200 kW of power (20 kW/cm 2 ) at high VSWR for 1-3 s pulses spaced 10-min apart. Breakdown measurements and fabrication considerations will be discussed. In addition, simulation of thermal conditions will be shown. The FFT will be computer controlled and must react to matching a load in a few hundred microseconds. This puts a severe requirement on power supply response time and its variation. In addition, the calculation time of the controlling software algorithms must be considered as well as the diffusion time of the controlling magnetic field through the waveguide wall. We will discuss these requirements and what we have done to meet them.

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

The design, construction and initial results of a new lower hybrid current drive (LHCD) launcher on Alcator C-Mod (Hutchinson et al 1994 Phys. Plasmas 1 1511) are presented. The new LHCD launcher (LH2) is based on a novel splitter concept which evenly distributes the microwave power in four ways in the poloidal direction. This design allows for simplification of the feeding structure while keeping the flexibility to vary the peak launched toroidal index of refraction, N-toroidal, from -3.8 to 3.8. An integrated model predicts good plasma coupling over a wide range of edge densities, while poloidal variations of the edge density are found to affect the evenness of power splitting in the poloidal direction. The measured transmission loss is about 30% lower than the previous launcher, and a clean N-toroidal spectrum has been confirmed. Power handling capability exceeding an empirical weak conditioning limit and reliable operation up to 1.1 MW net LHCD power have been achieved. A survey of antenna-plasma coupling shows the existence of a millimetric vacuum gap in front of the launcher. Fully non-inductive, reversed shear plasma operation has been demonstrated and sustained for multiple current diffusion times. The current drive efficiency, eta(LH) equivalent to n(e)R(0)I(p)/P-LH, of these plasmas is (0.2-0.25) x 10(20) m(-2)AW(-1), which is in agreement with the expected efficiency on the International Thermonuclear Experimental Reactor (ITER).

We are developing a Fast Ferrite Tuner (FFT) for use in the Lower Hybrid current drive system at 4.6GHz. The FFT consists of two stub tuners mounted on the narrow sides of a WR187 waveguide whose phase lengths are electrically controlled by adjusting the bias magnetic field in ferrite blocks mounted in the stubs. The current in the coils producing the bias magnetic fields will be driven by power supplies which are computer controlled. An algorithm in the computer will change the phase length of the stub tuners to eliminate the reflection from the load. The challenging tasks in the development are obtaining the necessary phase shift with minimum field variation, keeping the losses low (<; 0.1 dB), avoiding resonances in the waveguide during field changes, avoiding breakdown at for high power (>;100 KW), and developing a power supply that can respond in sub millisecond timescales. This report presents the design concept, the results of measurements on the tasks, and CST simulations of the design.

The new lower hybrid launcher (LH2) of the Alcator C-Mod tokamak is base d on a novel 4-way-splitter concept. A diagnostic based on the microwave probe s concept [Jacquet et al. 1997] has been installed to verify the LH2 design and study the physic s of LH wave coupling. A total of 32 dedicated probes measure the forward and reflec ted power in a carefully selected set of the active and passive waveguides of the LH2 grill. A new technique which relies only on the microwave probes for measuring the edge density pr ofile in front of the launcher is proposed. Einc = −(EAe +EB)e 1− e−2iK∆ (1) Eref = −(EAe +EB)e 1− e2iK∆ (2) Figure 1: Schematic of the microwave diagnostic; Equations for calculating incident and reflected waves; CAD drawing showing a cutout of the probe assembly A new lower hybrid phased waveguide array launcher (LH2) has been recently installed in the Alcator C-Mod tokamak. The antenna operates at 4.6 GHz and is based on a novel four-way-splitter concept [1], which evenly splits the microwave power in four ways in the poloidal direction. The launcher is made of a stack of 16 four way splitter modules, resulting in a grill of 16x4 waveguides. This design allows the simplification of feeding structure, while keeping the flexibility of launched toroidal spectrum. The LH2 antenna has been designed based on the LH wave linear coupling theory [2]. The TOPLHA and ALOHA codes predict that the evenness of power splitting in the poloidal direction, will be affected by poloidal uneveness of the density profiles in front of the launcher 2. Nonetheless, good 0 0.5 1 1.5 2 2.5 3 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 n e0 [1E18] Γ Reflection coefficients as a function of density profile λ=0.001 m λ=0.005 m λ=0.010 m λ=0.015 m λ=0.020 m λ=0.025 m λ=0.050 m λ=0.100 m Figure 2:A) Simulated average reflection coefficients for 90 degrees phasing as a function of ne0 for different values ofλ ; B) Reflection coefficient at the input of the 4-way-spitters in the case of even and uneven poloidal plasma density profiles. plasma coupling and a clean N ‖ spectrum over a wide range of edge density profiles are expected [3]. The assessment of the performance of the LH2 desi gn and the direct benchmarking of the 4-way-splitter concept requires measurements of the ant nna-plasma coupling. For this purpose, a diagnostic composed of a set of 32 microwave probe s [4] was designed to measure the forward and reflected waves at the grill mouth. These meas ur ments are also planned to be used as a part of the coupler protection system. In addition, the LH2 launcher is equipped with a total of six Langumir probes, and a X-mode reflectometer sys tem for measuring the density profile in front of the launcher. Microwave probes are installed in the waveguides of the two c entermost columns of waveguides and on every other waveguide at the bottom row of grill. In each of the monitored grill waveguides, two probes displaced by one quarter of the guide d wavelengthλg sense the wave field in the waveguide (Fig. 1). The forward and reflected wave s can then be deduced according to Eq. 1 and 2, whereK is the wavevector of theT E10, andL is the location of the reference plane. Two of the 8 dummy columns are also monitored by one pro be each, which is located at λg/4 from the short at the end of the waveguides. Microwaves are coupled to the probe through a small circular coupling hole located on the narrow side of the waveguides. The central conductor of the p robe is welded to the opposite side of the probe housing (Fig. 1). This design allowed good repro ducibility of the probe coupling performance ( ≈ −65±2 dB). Silicon-Dioxide (SiO2) cables are used in-vessel to ensure low losses and the stability of phase with respect to temperatur e variations. The phase and amplitude of each probe signal is finally measured with off-the-sh lf omodyne IQ detectors and are 0 20 40 60 80 100 120 140 160 180 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 phase Γ Γ as a function of phasing (n e0 =4E17[m]) λ=0.001 m λ=0.005 m λ=0.010 m λ=0.015 m λ=0.020 m λ=0.025 m λ=0.050 m λ=0.100 m 5 10 15 20 25 30 10 20 30 40 50 60 70 80 90 100 RMS error for n e0 and λ fit n e0 [10 m] λ [1 0 − 3 m ] Figure 3:A) Simulated average reflection coefficient as a function of the antenna ph asing for different values ofλ ; B) RMS error of the fitting procedure as a function of the fitting parameters ne0 andλ . digitized at 250 KHz. From a physics standpoint, the microwave probes diagnostic allows a detailed study of the LH wave coupling problem. The coupling efficiency strongly d epends on the parallel wavenumber N‖ of the launched wave, the plasma edge density n e0 a d the plasma density scale length λ = ne0/e ∂ r at the mouth of the LH antenna [2]. Figure 2 shows the predicte reflection coefficient for LH2 as a function of ne0 andλ for 90 degrees phasing. Currently, it is common practice to use the ambiguity of the edge density profile to ma ke coupling simulations fit to the experimental measurements [5, 6, 7, 8, 9, 10]. In C-Mod, a SOL X -mode reflectometer [11] will be used to measure the actual density profile in front of the la uncher at three poloidal locations. The availability of accurate density profile measurements w ill finally eliminate these free parameters and enable a self-consistent validation of the LH c oupling codes. In the experiments, to study coupling as a function of the SOL density profile, the radial positioning of the launcher behind the limiters, gas puffing, and radial movement of the p lasma column will be used. Also, Langmuir probes of different length (1 mm and 2 mm) will give a crude estimate of the density and its gradient at the grill mouth. In addition, the idea of a new technique has been developed fo r measuring the density profile in front of the launcher, without relying on other diagno stics but the microwave probes themselves. In this technique the phase of the LH waves is qui ckly swept while the plasma parameters are kept constant. Under the assumption that lin e r theory holds true, the density profile can be inferred by the least square fitting of the measu red reflection coefficients to the ones predicted by linear coupling theory (Fig. 3A). As an exa mple, lets consider a conventional linear density model with two parameters to be determined: t he edge density ne0 and the density scale lengthλ . The ALOHA code has been used to compute the plasma impedance for th complete set of density profiles in the range 1 .2×1017< ne0< 30×1017m−3 and 0.1< λ < 10cm. Figure 3B shows a plot of RMS error of the fitting procedure if an ideal linear density profile with ne0 = 9E17 [m−3] andλ = 15E−3 [mm] was to be measured, thus giving an idea of the sensitivity of this method. For this example a phase scan bet ween 50 and 145 degrees has been considered, which would keeping the reflection coefficient i n the experiment within acceptable limits. The average density profile can be inferred by consid ering the average reflection coefficients of the whole antenna. However, for the specific case of LH2, this technique can be used to deduce the density profiles at four poloidal locations using measurements from the RF probes located on columns 8 and 9 of the LH2. The X-mode reflectometer will measure the density profile in front of the launcher at three poloidal locations a d will be used to validate the phase scan method. If successful, this technique could be used rou tinely at the beginning of every LH shot and may be the building block for a phase feedback cont rol system which minimizes reflection. The microwave probes diagnostic will also offer a unique opp ortunity to study coupling at high power, a regime where linear coupling theory is known to break down [2]. It has been conjectured that this is due to the effect of ponderomotive f orces on the edge plasma density, however experimental observations are often contradictor y and a definitive theoretical model in this regime is still missing [2, 7, 10]. Work supported by USDOE awards DE-FC02-99ER54512 and DE-AC 02-76CH03073.

We have developed high power four and eight way splitters for a new Lower Hybrid launcher. The motivation for the new launcher was the need to provide more power and reliability to the launcher structure. In addition there was a desire to simplify and increase the reliability of the implementation of the alumina windows. The launcher consists of 64 waveguide apertures powered by 8 klystrons with maximum power of 250 kW each at 4.6 GHz. Hence, it is necessary to split the power from each Klystron into eight separate waveguides. The outputs of the splitter have a difference in power less than 0.1dB and phase less than 2 degree. The design analysis of the splitter was done with the computer code CST. Structure analysis was performed using Ansys. The splitter is fabricated by machining an open cavity into a thick stainless steel plate creating the specified internal geometry. It is machined to a tight tolerance of +/- 0.005″. A fitted lid is then welded on top of the open cavity using electron beam welding. The excess metal is removed with Electro discharge machining (EDM) creating the external geometry. The waveguides are then butt-welded to the splitter. Welding fixtures/parameters are being developed to achieve the desired tolerances. Two methods for attaching the ceramic windows are being evaluated, brazing and electro-forming.

The Ion Cyclotron RF Transmitter System (ICRF) at Alcator C-Mod comprises four separate transmitters each capable of driving 2 MW of power into plasma loads. Four separate transmission lines guide RF power into three antennas, each mounted in a separate horizontal port, in the C-Mod Tokamak. Protection for the antennas, matching elements and transmission line is accomplished by two unique but interdependent subsystems encompassed by the ICRF Fault System. The Antenna Protection System evaluates antenna phasing and voltage, sets fault thresholds, generates fault signals, and passes fault information to the Master Fault Processor. During operation, the Master Fault Processor is responsible for detecting hazards along the transmission line, generating faults, processing faults from the Antenna Protection System, terminating RF drive and extinguishing faults within 10 mus. In addition, the system controls various delays and sets the boundaries for RF retries. The ICRF Control System provides amplitude regulation for all antennas and phase control for a four-strap antenna. We are modifying some of the fault processing components and control elements of these systems in an effort to improve reliability and serviceability, and increase flexibility. This upgrade will reduce wired interconnections, add remote features to improve access to key operating parameters, improve RF isolation with new switching components, simplify phase control, and expand the RF regulation system to an active control regime whereby plasma parameters may become direct feedback elements for RF regulation. Details of the proposed upgrade to the system will be presented, and implementation of any new technological tools will be discussed.

Abstract This paper reviews the physics and technology of wave-particle-interaction experiments in the ion cyclotron range of frequencies (ICRF) and the lower hybrid (LH) range of frequencies (LHRF) on the Alcator C-Mod tokamak. Operation of fixed frequency (80 MHz) and tunable (40- to 80-MHz) ICRF transmitters and the associated transmission system is described. Key fabrication issues that were solved in order to operate a four-strap ICRF antenna in the compact environment of C-Mod are discussed in some detail. ICRF heating experiments utilizing the hydrogen (H) and helium-3 (3He) minority heating schemes are described, and data are presented demonstrating an overall heating efficiency of 70 to 90% for the (H) minority scheme and somewhat lower efficiency for (3He) minority heating. Mode conversion electron heating experiments in D(3He), D(H), and H(3He) discharges are also reported as well as simulations of these experiments using an advanced ICRF full-wave solver. Measurements of mode-converted ion cyclotron waves and ion Bernstein waves using a phase contrast imaging diagnostic are presented and compared with the predictions of a synthetic diagnostic code that utilizes wave electric fields from a full-wave solver. The physics basis of the LH current profile control program on Alcator C-Mod is also presented. Computer simulations using a two-dimensional (velocity space) Fokker Planck solver indicate that ~200 kA of LH current can be driven in low-density H-mode discharges on C-Mod with ~3 MW of LHRF power. It is shown that this off-axis LH current drive can be used to create discharges with nonmonotonic profiles of the current density and reversed shear. An advanced tokamak operating regime near the ideal no-wall β limit is described for C-Mod, where ~70% of the current is driven through the bootstrap effect. The LH power is coupled to C-Mod through a waveguide launcher consisting of four rows (vertically) with 24 guides per row (toroidally). A detailed description of the LH launcher fabrication is given in this paper along with initial operation results.

S. Scott 1), A. Bader 2), M. Bakhtiari 3), N. Basse 4), W. Beck 2), T. Biewer 5), S. Bernabei 1), P. Bonoli 2), B. Bose 2), R. Bravenec 6), I. Bespamyatnov 6), R. Childs 2), I. Cziegler 2), R. Doerner 7), E. Edlund 2), D. Ernst 2), A. Fasoli 8), M. Ferrara 2), C. Fiore 2), T. Fredian 2), A. Graf 9), T. Graves 2), R. Granetz 2), N. Greenough 1), M. Greenwald 2), M. Grimes 2), O. Grulke 10), D. Gwinn 11), R. Harvey 12), S. Harrison 3), T. C. Hender 13), J. Hosea 1), D. F. Howell 13), A. E. Hubbard 2), J. W. Hughes 2), I. Hutchinson 2), A. Ince-Cushman 2), J. Irby 2), T. Jernigan 5), D. Johnson 2), J. Ko 2), P. Koert 2), B. LaBombard 2), A. Kanojia 2), L. Lin 2), Y. Lin 2), B. Lipschultz 2), J. Liptac 2), A. Lynn 14), P. MacGibbon 2), E. Marmar 2), K. Marr 2), M. May 15), D. R. Mikkelsen 1), R. McDermott 2), A. Parisot 2), R. Parker 2), C. K. Phillips 1), P. Phillips 6), M. Porkolab 2), M. Reinke 2), J. Rice 2), W. Rowan 6), M. Sampsell 6), G. Schilling 1), A. Schmidt 2), N. Smick 3), A. Smirnov 12), J. Snipes 2), D. Stotler 1), J. Stillerman 2), V. Tang 15), D. Terry 2), J. Terry 2), M. Ulrickson 16), R. Vieira 2), G. Wallace 2), D. Whyte 2), J. R. Wilson 1), G. Wright 3), J. Wright 2), S. Wolfe 2), S. Wukitch 2), G. Wurden 17), H. Yuh 18), K. Zhurovich 2), J. Zaks 2), S. Zweben 1)

This paper summarizes highlights of research results from the Alcator C-Mod tokamak covering the period 2006–2008. Active flow drive, using mode converted ion cyclotron waves, has been observed for the first time in a tokamak plasma, using a mix of D and 3He ion species; toroidal and poloidal flows are driven near the location of the mode conversion layer. ICRF induced edge sheaths are implicated in both the erosion of thin boron coatings and the generation of metallic impurities. Lower hybrid range of frequencies (LHRF) microwaves have been used for efficient current drive, current profile modification and toroidal flow drive. In addition, LHRF has been used to modify the H-mode pedestal, increasing temperature, decreasing density and lowering the pedestal collisionality. Studies of hydrogen isotope retention in solid metallic plasma facing components reveal significantly higher retention than expected from ex situ laboratory studies; a model to explain the results, based on plasma/neutral induced lattice damage, has been developed and tested. During gas-puff mitigation of disruptions, induced MHD instabilities cause the magnetic field to become stochastic, resulting in reduction of halo currents, spreading of plasma power loading and loss of runaway electrons before they cause damage. Detailed pedestal rotation profile measurements have been used to infer E r profiles, and correlation with global H-mode confinement. An improved L-mode regime, obtained at q 95 ⩽ 3 with ion drift away from the active X-point, shows very good energy confinement with a strong temperature pedestal, a weak density pedestal, and no evidence of particle or impurity accumulation, without the need for ELMs or any additional edge density regulation mechanism.

MIT and PPPL have joined together to fabricate a high-power lower hybrid current drive (LHCD) system for supporting steady-state AT regime research on Alcator C-Mod. The goal of the first step of this project is to provide 1.5 MW of 4.6 GHz rf [radio frequency] power to the plasma with a compact launcher which has excellent spectral selectivity and fits into a single C-Mod port. Some of the important design, construction, calibration and testing considerations for the launcher leading up to its installation on C-Mod are presented here.

We have developed a compact, mass-producible probe inserted into two holes 1/4 wavelength apart in the narrow side of the reduced height waveguide. The probe consists of two current loops and a directional coupler mounted directly on a single microstrip circuit board thereby greatly reducing the physical size of the probe. Calibrations have shown directivity greater than 20 dB at 4.6 GHz for each of the over 100 probes built.

Recent Alcator C-Mod experimental campaigns have focused upon the study of the Advanced Tokamak regimes, which includes characterization of the RF heating, the formation and dynamics of internal barriers, H-mode edge pedestal, and divertor and scrape-off physics. The ICRF system has been recently upgraded with the improved performance of the 4-strap antenna. Total ICRF power in excess of 5 MW has been launched successfully into the plasma during this campaign. Due to the compact nature of C-Mod, the power feeds for the antenna are vacuum strip lines. Their orientation, to the tokamak B-field, is governed by maintaining E<15 kV/cm in locations where the RF E-field is parallel to tokamak B-field. Other modifications included improved protection tile grounding and installation of protective shields for Faraday screen ceramic isolators. The antennas also make use of BN protection tiles to eliminate high Z impurities from the antennas. The present empirical power limit results from arcing in a region of the antenna strap where E/spl sim/15 kV/cm and parallel to B and injections from the metallic fasteners used to attach the BN tiles to the antenna.

A 3 MW (upgradeable to 4 MW) 4.6 GHz Lower Hybrid Current Drive (LHCD) system is being implemented on Alcator C-Mod to control and sustain current profile evolution. The LHCD low-power microwave and active control system, using fast vector modulators, will provide a phase and amplitude controlled driver for each of twelve 4.6 GHz, 250 kW klystrons. Resulting phase and power outputs of each klystron will be monitored by I-Q detectors and made available for closed-loop control of the klystron phase and amplitude during plasma shots. Synchronized digital controllers will be implemented for each control loop and operation will be coordinated with external fast protection circuitry provided in the LHCD transmitter and coupler protection systems.