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The operational experience of present day high power EC systems is nominally only for a few seconds. This is a long way away from the thousands of seconds required for ITER. It would be beneficial to the ITER program that EC system components be tested to full parameters prior to committing to producing the full set of components. The planned growth in the EC system on DIII-D over the next few years provides the opportunity to assemble a test stand for ITER EC components. By building the DIII-D hardware to the ITER specifications it will allow ITER to gain beneficial prototyping experience on a working tokamak, prior to committing to building the hardware for delivery to ITER.

The performance requirement of 1 (possibly 2) MW cw at 170 GHz for ITER Electron Cyclotron Heating and Current Drive transmission line components is much more demanding than the 1 MW 10 s performance, generally at 110 GHz, that has been demonstrated on present devices. The high ITER heat loads will require enhanced cooling and, for some components, new or modified designs. In addition to thermal management issues, the components must be designed to have very low losses in order to meet the ITER transmission line efficiency requirements. Testing at representative ITER conditions of some components has been initiated at the JAEA 170 GHz gyrotron test stand at Naka, Japan. Plans for testing additional components, both at JAEA and in the US will be presented. The US ITER Project Office (USIPO) plans to test a complete prototype ITER transmission line in order to validate the designs for use on ITER, and GA is providing some components to the USIPO for initial tests.

The efficiency of the transmission line for the 110 GHz ECH system was measured in DIII‐D using a low power rf source. The measured efficiency was about 10% lower than expected from theoretical analysis of the components. The launcher temperature increase during rf pulses was measured and the peak mirror surface temperature was inferred from a simulation.

At DIII-D, five 110 GHz gyrotrons are operating routinely for 2.0 s pulses at generated power levels {ge}750 kW per gyrotron. A sixth gyrotron is being installed, which should bring the generated power level to >4 MW and the injected power to about 3.0 MW. The output power now can be modulated by the plasma control system to fix T{sub e} at a desired value. The system is being used as a tool for control of current diffusion, for current profile control and other experiments leading to advanced tokamak operation.

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...

Artificially grown diamond has extremely low absorption for microwaves in the millimeter wave range, making this material an attractive candidate for output windows on high power gyrotrons. Several windows have failed in this application due to higher than expected losses. Infrared measurements of the window temperature on a high power gyrotron operating at 110 GHz have been performed. The peak central temperature and time to equilibrium during the rf pulse were consistent with the low loss properties of the material determined from low power cavity measurements.

Three 110 GHz gyrotrons with nominal output power of 1 MW each have been installed and are operational on the DIII-D tokamak. One gyrotron is built by Gycom and 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. The second and third gyrotrons were built by Communications and Power Industries (CPI). The first 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 0.35 MW. The second CPI gyrotron, utilizes a single disc chemical-vapor-deposition diamond window, that employs water cooling around the edge of the disc. Calculation predict that the diamond window should be capable of full 1 MW cw operation. All gyrotrons are connected to the tokamak by a low-loss-windowless evacuated transmission line using circular corrugated waveguide for propagation in the HE11 mode. Each waveguide system incorporates a two mirro...

Three different transmission line configurations for operating a four-element antenna array with a single rf power source are compared. The goal of this study is to obtain a system that presents a matched load to the generator despite variation of the loading of the array elements due to changing plasma conditions.

The first of three MW-level 110 GHz gyrotrons was operated into the DIII-D tokamak in late 1996. Two additional units will be commissioned during 1997. Each gyrotron is connected to the tokamak by a low loss, windowless, evacuated transmission line using circular corrugated waveguide carrying the HE11 mode. The microwave beam spot is well focused with a spot size of approximately 6 cm and can be steered poloidally from the center to the outer edge of the plasma. The initial operation 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. The injection was 19° off perpendicular for current drive.

The physics of non-inductive current drive and current profile control using the fast magnetosonic wave has been demonstrated on the DIII-D tokamak. In non-sawtoothing discharges formed by neutral beam injection (NBI), the radial profile of the fast wave current drive (FWCD) was determined by the response of the loop voltage profile to co, counter, and symmetric antenna phasings, and was found to be in good agreement with theoretical models. The application of counter FWCD increased the magnetic shear reversal of the plasma and delayed the onset of sawteeth, compared to co FWCD. The partial absorption of fast waves by energetic beam ions at high harmonics of the ion cyclotron frequency was also evident from a build up of fast particle pressure near the magnetic axis and a correlated increase in the neutron rate. The anomalous fast particle pressure and neutron rate increased with increasing NBI power and peaked when a harmonic of the deuterium cyclotron frequency passed through the center of the plasma. The experimental FWCD efficiency was highest at 2 T where the interaction between the fast waves and the beam ions was weakest; as the magnetic field strength was lowered, the FWCD efficiency decreased to approximately half of the maximum theoretical value.

The resilience of a maximally flat externally coupled traveling wave antenna (TWA) is contrasted with the sensitivity of a simple directly driven resonant loop array to vacuum and plasma conditions in DIII-D. We find a unique synergy between standing and traveling wave resonant TWA components. This synergy extends TWA operation to several passbands between 60 and 120 MHZ, provides 60{degrees}- 120{degrees} tunability between elements within a 1-2 MHZ bandwidth and permits efficient and continuous operation during ELMing H-mode.

The physics of electron heating and current drive with the fast magnetosonic wave has been demonstrated on DIII-D, in reasonable agreement with theoretical modeling. A recently completed upgrade to the fast wave capability should allow full noninductive current drive in steady state advanced confinement discharges and provide some current density profile control for the Advanced Tokamak Program. DIII-D now has three four-strap fast wave antennas and three transmitters, each with nominally 2 MW of generator power. Extensive experiments have been conducted with the first system, at 60 MHz, while the two newer systems have come into operation within the past year. The newer systems are configured for 60 to 120 MHz. The measured FWCD efficiency is found to increase linearly with electron temperature as {gamma} = 0.4 {times} 10{sup 18} T{sub eo} (keV) [A/m{sup 2}W], measured up to central electron temperature over 5 keV. A newly developed technique for determining the internal noninductive current density profile gives efficiencies in agreement with this scaling and profiles consistent with theoretical predictions. Full noninductive current drive at 170 kA was achieved in a discharge prepared by rampdown of the Ohmic current. Modulation of microwave reflectometry signals at the fast wave frequency is beingmore » used to investigate fast wave propagation and damping. Additionally, rf pick-up probes on the internal boundary of the vessel provide a comparison with ray tracing codes, with dear evidence for a toroidally directed wave with antenna phasing set for current drive. There is some experimental evidence for fast wave absorption by energetic beam ions at high cyclotron harmonic resonances.« less

Recent results from the DIII-D tokamak have provided significant contributions to the understanding of many of the elements of tokamak physics and the application of this understanding to the design of next generation devices including ITER and CIT. The limitations of magnetohydrodynamic stability on the values of plasma beta (the ratio of kinetic pressure to the containing pressure of the magnetic field) that can be attained has been experimentally demonstrated and found to be described by existing theory. Values of beta (10.7%) well in excess of those required for proposed devices (ITER and CIT) have been demonstrated. Regimes of confinement (H-mode) have been established that scale favorably to proposed next generation devices, and experiments demonstrating the dependence of the energy confinement on plasma size have been completed. Understanding of confinement is rapidly developing especially in the areas of bulk transport and the role of turbulence in the plasma edge. Key experimental results in areas of plasma transport and edge plasma phenomena are in agreement with theories based on short wavelength turbulence. Control of the divertor heat loads and impurity influx has been demonstrated, and new progress has been made in the understanding of plasma edge phenomena. Experiments with ion Bernstein wave heating have not found regimes in which these waves can produce effective central ion heating. Electron cyclotron current drive experiments have demonstrated 70 kA of driven current in 400 kA discharges.

The authors' data indicate that the L-mode to H-mode transition in the DIII-D tokamak is associated with the sudden reduction in anomalous, fluctuation-connected transport across the outer midplane of the plasma. In addition to the reduction in edge density and magnetic fluctuations observed at the transition, the edge radial electric field becomes more negative after the transition. They have determined the scaling of the H-mode power threshold with various plasma parameters; the roughly linear increase with plasma density and toroidal field are particularly significant. Control of the ELM frequency and duration by adjusting neutral beam input power has allowed us to produce H-mode plasmas with constant impurity levels and durations up to 5 s. Energy confinement time in ohmic H-mode plasmas and in deuterium H-mode plasmas with deuterium beam injection can exceed saturated ohmic confinement times by at least a factor of two. Energy confinement times above 0.3 s have been achieved in these beam-heated plasmas with plasma currents in the range of 2.0 to 2.5 MA. Local transport studies have shown that electron and ion thermal diffusivities and angular momentum diffusivity are comparable in magnitude and all decrease with increasing plasma current.

Tokamak discharges using the expanded boundary divertor in the DIII-D device exhibit H-mode confinement. With neutral-beam power up to 6 MW, energy confinement remains comparable to the Ohmic value at a plasma current of 1 MA. Confinement is also independent of plasma density and toroidal field. Confinement increases with plasma current, but the exact functional dependence is, as yet, uncertain. These results show that the H mode can be achieved in a reactor-compatible open divertor configuration.

Neutral-beam current-drive experiments in the DIII-D tokamak with a single null poloidal divertor are described. A plasma current of 0.34 MA has been sustained by neutral beams alone, and the energy confinement is of $H$-mode quality. Poloidal $\ensuremath{\beta}$ values reach 3.5 without disruption or coherent magnetic activity suggesting that these plasmas may be entering the second stability regime.