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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.

A key goal in the control of a magnetic fusion reactor is to maintain current profiles that are compatible with a high fraction of self-generated non-inductive current as well as with magnetohydrodynamic (MHD) stability at high plasma pressure. This enables high fusion gain and noninductive sustainment of plasma current for steady-state operation. The approach taken toward establishing such plasma current profiles at the DIII-D tokamak is to create the desired profile during the plasma current ramp-up and early flattop phases. The evolution in time of the current profile is related to the evolution of the poloidal flux, which is modeled in normalized cylindrical coordinates using a nonlinear partial differential equation (PDE) usually referred to as the magnetic diffusion equation. We propose, and test in simulations, an extremum-seeking-based, receding-horizon, diffusivity-interior-boundary control scheme designed to match as close as possible a desired current profile within a prespecified time interval.

Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

Advanced tokamak (AT) research seeks to develop steady-state operating scenarios for ITER and other future devices from a demonstrated scientific basis. Normalized target parameters for steady-state operation on ITER are 100% non-inductive current operation with a bootstrap current fraction f BS ≥ 60%, q 95 ∼ 4-5 and G ≡ β N H scaling /q 2 95 ≥ 0.3. Progress in realizing such plasmas is considered in terms of the development of plasma control capabilities and scientific understanding, leading to improved AT performance. NSTX has demonstrated active resistive wall mode stabilization with low, ITER-relevant, rotation rates below the critical value required for passive stabilization. On DIII-D, experimental observations and GYRO simulations indicate that ion internal transport barrier (ITB) formation at rational-q surfaces is due to equilibrium zonal flows generating high local E x B shear levels. In addition, stability modelling for DIII-D indicates a path to operation at β N ≥ 4 with q min ≥ 2, using broad, hollow current profiles to increase the ideal wall stability limit. Both NSTX and DIII-D have optimized plasma performance and expanded AT operational limits. NSTX now has long-pulse, high performance discharges meeting the normalized targets for an spherical torus-based component test facility. DIII-D has developed sustained discharges combining high beta and ITBs, with performance approaching levels required for AT reactor concepts, e.g. β N = 4, H 89 = 2.5, with f BS > 60%. Most importantly, DIII-D has developed ITER steady-state demonstration discharges, simultaneously meeting the targets for steady-state Q ≥ 5 operation on ITER set out above, substantially increasing confidence in ITER meeting its steady-state performance objective.

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 experimental and modeling results on H-mode edge-localized mode (ELM) instabilities from the DIII-D tokamak project are reviewed. This work has led to the conclusion that the most common type of ELM, called Type I, is triggered by a coupled peeling-ballooning instability driven by the pressure gradient and current density in the H-mode edge pedestal region. Good agreement is found between theoretically predicted stability boundaries and toroidal mode numbers for this instability and experimental observations of edge pedestal parameters and ELM amplitude and frequency as a function of discharge shape and edge-region collisionality. The range of toroidal mode numbers for which there is access to a second stability regime is shown to play an important role. This model of H-mode edge stability has been used to predict the pedestal parameters for ITER and FIRE.

A potential new standard in stationary tokamak performance is emerging from experiments on DIII-D. These experiments have demonstrated the ability to operate near the free boundary, n = 1 stability limit with good confinement quality under stationary conditions. The normalized fusion performance is at or above that projected for Qfus = 10 operation in the International Thermonuclear Experimental Reactor (ITER) design over a wide operating range in both edge safety factor (2.8–4.7) and plasma density (35–70% of the Greenwald density). Projections to ITER based on this data are uniformly positive and indicate that a wide range of operating options may be available on ITER, including the possibility of sustained ignition. Recent experiments have demonstrated the importance of a small m = 3, n = 2 neoclassical tearing mode in avoiding sawteeth and the effect of edge localized modes on tearing mode stability at an edge safety factor near 3. Transport studies using the GLF23 turbulence transport code suggest that E × B shear stabilization is important in reproducing the measured profiles in the simulation. Yet, even in cases in which the toroidal rotation is moderate, confinement quality is robustly better than the standard H-mode confinement scalings.

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...

A high speed data acquisition system which is closely coupled with a high speed digital processor is described. Data acquisition at a rate of 40 million 14 bit data values per second is possible simultaneously with data processing at a rate of 80 million floating point operations per second. This is achieved by coupling a commercially available VME format single board computer based on the Intel i860 microprocessor with a custom designed first‐in, first‐out memory circuit that transfers data at high speed to the processor board memory. Parallel processing to achieve increased computation speed is easily implemented because the data can be transferred simultaneously to multiple processor boards. Possible applications include high speed process control and real time data reduction. A specific example is described in which this hardware is used to implement a feedback control system for 18 parameters which uses 100 input signals and achieves a 100 μs cycle time.

The techniques used in implementing two applications of real time digital analysis of data from the DIII‐D tokamak are described. These tasks, which are demanding in both the speed of data acquisition and the speed of computation, execute on hardware capable of acquiring 40 million data samples per second and executing 80 million floating point operations per second. In the first case, a feedback control algorithm executing at a 10 kHz cycle frequency is used to specify the current in the poloidal field coils in order to control the discharge shape. In the second, fast Fourier transforms of Mirnov probe data are used to find the amplitude and frequency of each of eight toroidal mode numbers as a function of time during the discharge. Data sampled continuously at 500 kHz are used to produce results at 2 ms intervals.

The National Spherical Torus Experiment (NSTX) has as one of its primary goals the demonstration of the attractiveness of the spherical torus concept as a fusion power plant. Central to this goal is the achievement of high plasma /spl beta/ (=2/spl mu//sub 0/ B/sup 2/ a measure of the efficiency of a magnetic plasma confinement system). It has been demonstrated both theoretically and experimentally that the maximum achievable /spl beta/ is a strong function of both local and global plasma parameters. It is therefore important to optimize control of the plasma. To this end a phased development plan for digital plasma control on NSTX is presented. The relative level of sophistication of the control system software and hardware will be increased according to the demands of the experimental program in a three phase plan. During Day 0 (first plasma), a simple coil current control algorithm will initiate plasma operations. During the second phase (Day 1) of plasma operations the control system will continue to use the preprogrammed algorithm to initiate plasma breakdown but will then change over to a rudimentary plasma control scheme based on linear combinations of measured plasma fields and fluxes. The third phase of NSTX plasma control system development will utilize the rtEFIT code, first used on DIII-D, to determine, in real-time, the full plasma equilibrium by inverting the Grad-Shafranov equation. The details of the development plan, including a description of the proposed hardware are presented.

It is demonstrated that the radial electric field in the Phaedrus Tandem Mirror [Plasma Physics and Controlled Nuclear Fusion Research 1984 (IAEA, Vienna, 1985), Vol. 2, p. 265] can be altered using plasma potential control rings (PPC rings) situated at each end of the device, and the azimuthal plasma rotational velocity may thus be varied. Low‐frequency (ω≪ωci), low azimuthal mode number (m=−1 and m=−2) instabilities driven by E×B rotation are observed and shown to be in qualitative agreement with the theory of Freidberg and D’Ippolito [Phys. Fluids 26, 2657 (1983)], and Phillips [Phys. Fluids 27, 1783 (1984)] for the case when Phaedrus is operated as a conventional tandem mirror with minimum‐‖B‖ end cells.

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 advanced tokamak program on DIII-D is targeting the full integration of high-beta and high-bootstrap/noninductive current fraction for long-pulse lengths and the high confinement consistent with these features. Central to achieving these simultaneously is access to the highest ideal beta limits possible to maximize the headroom for experimental operation with RWM control. A study of the ideal-MHD stability is done for plasmas modeled after DIII-D advanced tokamak plasmas, varying the plasma elongation, triangularity, and outboard squareness. The highest beta(sub)N limits reach 6-7 for the n=1 kink mode for all elongation, outer squareness values, and plasma triangularity equals 0.8.

First plasma, with a maximum current of 300kA, was achieved on NSTX in February 1999. These results were obtained using preprogrammed coil currents. The first controlled plasmas on NSTX were made starting in August 1999 with the full 1MA plasma current achieved in December 1999. The controlled quantities were plasma position (R, Z) and current (Ip). Variations in the plasma shape are achieved by adding preprogrammed currents to those determined by the control parameters. The control system is fully digital, with plasma position and current control, data acquisition, and power supply control all occurring in the same four-processor real time computer. The system uses the PCS (Plasma Control Software) system designed at General Atomics. Modular control algorithms, specific to NSTX, were written and incorporated into the PCS. The application algorithms do the actual control calculations, with the PCS handling data passing. The control system, including planned upgrades, will be described, along with results of the initial controlled plasma operations. Analysis of the performance of the control system will also be presented.

The DIII-D Advanced Tokamak (AT) program in the DIII-D tokamak [J. L. Luxon, Plasma Physics and Controlled Fusion Research, 1986, Vol. I (International Atomic Energy Agency, Vienna, 1987), p. 159] is aimed at developing a scientific basis for steady-state, high-performance operation in future devices. This requires simultaneously achieving 100% noninductive operation with high self-driven bootstrap current fraction and toroidal beta. Recent progress in this area includes demonstration of 100% noninductive conditions with toroidal beta, βT=3.6%, normalized beta, βN=3.5, and confinement factor, H89=2.4 with the plasma current driven completely by bootstrap, neutral beam current drive, and electron cyclotron current drive (ECCD). The equilibrium reconstructions indicate that the noninductive current profile is well aligned, with little inductively driven current remaining anywhere in the plasma. The current balance calculation improved with beam ion redistribution that was supported by recent fast ion diagno...

It is shown that the predictions of a numerical code (ANTENA) and the data of wave field measurements in the Phaedrus-B tandem mirror are consistent (±25%) for right-handed (−) wave fields and less so (±40%) for left-handed (+) wave fields in the plasma core, and that they disagree for + fields near the column edge. Shorting out or reduction of the wave azimuthal electric fields by limiters is the probable cause of this discrepancy. The ICRF fluctuating wave fields are shown as || contour maps in the r-z plane, where the + data peak at a smaller radius than predicted. The waves are characterized by different dominant axial wave numbers for the left- and right-handed circularly polarized fields.

A thermal barrier plasma potential structure in a three-cell axisymmetric tandem mirror has been produced using only the ion cyclotron range of frequencies (ICRF), with a central cell density of up to 1013 cm−3 and a central cell ion temperature of approximately 40 eV. Thermal barrier like potential dips occurred in the region of a magnetic field gradient, at a magnetic beach of a slow wave launched in the end cell. The barrier formation is a result of a combination of radiofrequency trapping, radiofrequency enhanced potentials and μ∇B forces. The barrier pumping is consistent with μ∇B forces and θ × drifts.

Plasma with significant central-cell beta can be sustained in a tandem mirror composed of three axisymmetric simple mirror cells by the use of ion-cyclotron resonant heating. Radial ponderomotive force due to the rf electric field opposes the centrifugal force due to the field-line curvature to ensure interchange stability. This is indicated by the sensitive dependence of the plasma stability on the sign of the difference between the rf frequency and the ion-cyclotron frequency.

It is shown that wave eigenmode fields can provide RF stabilization of the interchange instability. The excitation level of an m = +1 fast magnetosonic wave is controlled by tailoring of the k/sub X/ spectrum of an axial ion-cyclotron radio-frequency antenna array in the Phaedrus-B tandem mirror system. When the k/sub X/ spectrum is chosen to excite the wave strongly, density-fluctuation levels in the plasma are greatly reduced.

Electrostatic ion confinement is demonstrated in an rf-sustained tandem mirror and shown to have a dynamic nature. Measurements of plasma potential and ion end loss on particular flux tubes show that confinement can be enhanced by a factor of 3 over direct-flow loss rates. The axial confining potential can vary with radial position and time resulting in changes in ion confinement with abrupt reductions in central-cell plasma density when axial potential confinement is reduced.

Electrostatic end plugging is observed in a completely axisymmetric, three cell tandem mirror under conditions where the central‐cell plasma density is always larger than the end‐cell density. A factor of 4 increase in the central‐cell density, to a maximum of 1.2×1013 cm−3 with simultaneous plasma beta of 13%, occurs upon application of the end plugging potential. Ion confining potentials of 25 V and 80 V at the two ends of the device, respectively, result in a factor of 2.5 increase in the axial confinement time for Tic =40 eV in agreement with the collisional flow model for ion confinement. The non‐Boltzmann ion confining potential is caused by electron heating in the end cells by rf near the ion‐cyclotron frequency. The initial central‐cell density rise is caused by an increase in the ionization rate that occurs because of an increase in the electron temperature. The density remains high throughout the end‐cell heating pulse as a result of increased particle confinement time. There is no nonambipolar ...

The possibility of fuelling pure RF-powered tandem mirror end plugs by gas puffing is considered. It is shown experimentally that gas puffing can lead to significant increases in plug density and that the details of the plug density buildup depended on the antenna locations relative to the central cell and the plug midplane. Large charge-exchange energy loss resulting from the puff gas makes it an inefficient technique for fuelling high-temperature end plugs. The possibility of employing gas puffing for building up high-density end plugs composed of relatively cold ions is also considered as a means of achieving electrostatic confinement for an ICRF-stabilized central cell. Experimental enhancement of the central-cell density, which shows no significant cooling of the central cell by the colder plug plasmas, is demonstrated. The RF power requirements for refuelling an ICRF-powered end plug to a given density and temperature are estimated.

Reduced recycling, reduced edge neutral pressure, improved density control, and improved discharge reproducibility have been achieved in the DIII-D tokamak by in situ helium condition of the graphite tiles. An improvement in energy confinement has been observed in hydrogen discharges with hydrogen beam injection after helium preconditioning. After the graphite wall coverage in DIII-D was increased to 40%, helium glow wall conditioning, routinely applied before each tokamak discharge, has been necessary to reduce recycling and obtain H-mode. The utilization of helium glow wall conditioning was an important factor in the achievement of an ohmic H-mode, i.e. no auxiliary heating, with significant improvement in ohmic energy confinement.

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.

Self‐diffusivities of carbon dioxide at 1180°—1680°K have been determined by a new version of the continuous flow method. Results can be correlated with previous low‐temperature data (mean error 1.5%) in terms of a single set of parameters of the Lennard‐Jones (6–12) potential (σ=2.70 A, e/k=871°K) valid for the entire temperature range (195°—1680°K). Literature values of viscosity (200°—1500°K) may also be correlated by a single set of parameters (σ=3.94 A, e/k=200°K). It is suggested that one can include both viscosity and self‐diffusivity data in the same correlation if the nonsphericity of the CO2 molecule, which may have the strongest effect at high temperatures, is accounted for. This can be done approximately by use of the nonsphericity factors calculated by Curtiss and co‐workers for rigid, spherocylindrical molecules, if one assumes a temperature dependence for the factor such that it becomes about unity at room temperature.

It is found that ion cyclotron rf can create an ion plugging plasma potential equal to several times T/sub e//e(approx. =T/sub i//e) in the Phaedrus tandem mirror end cells or at the end of the central cell. A model based on a local balance of electron expulsion by the rf field E/sub z/ and trapping by Coulomb scattering predicts the magnitude and scaling of the potential with density and rf antenna current. This model is in reasonable agreement with experimental observations. This ion-cyclotron rf enhanced potential increases as the plasma density decreases and as the antenna current increases.

The central cell of the Phaedrus‐B tandem mirror is equipped with four dual half‐turn ICRF antennas. The antennas are arranged in two close‐spaced pairs, with the antenna pairs located 50 cm to either side of central cell midplane. In these experiments, two antennas (one from each pair) with an axial separation of 1 m are driven by separate RF amplifiers, each with ≳200 KW power output. The relative phase of the driven antenna currents may be fixed throughout a shot, or rapidly swept (180° in

A device has been constructed for the study of the interaction between a fast ion beam and a target plasma of separately controllable parameters. The beam of either hydrogen or helium ions has an energy of 1-4 keV and a total current of 0.5-2 A. The beam energy and beam current can be varied separately. The ion source plasma is created by a pulsed (0.2-10-ms pulse length) discharge in neutral gas at up to 3 x 10(-3) Torr. The neutrals are pulsed into the source chamber, allowing the neutral pressure in the target region to remain less than 5 x 10(-5) Torr at a 2-Hz repetition rate. The creation of the source plasma can be described by a simple set of equations which predict optimum source design parameters. The target plasma is also produced by a pulsed discharge. Between the target and source chambers the beam is neutralized by electrons drawn from a set of hot filaments. Currently under study is an unstable wave in a field-free plasma excited when the beam velocity is nearly equal to the target electron thermal velocity (v(beam) approximately 3.5 x 10(7) cm/s, Te = 0.5 eV).

As a complement to ITER, the existing large and medium-size tokamaks are expected to explore more advanced operation scenarios compatible with a future power reactor. Specifically, the following urgent issues are required to be solved hopefully with an experimental demonstration: to work out one or more solutions to maintain a steady-state plasma with high performance (high beta and high bootstrap current fraction), and to suppress plasma instabilities almost completely. The developed solutions have to be confirmed using a DT burning plasma in ITER. In this article, we first review on system configurations and essential functions used for plasma control systems in existing tokamaks, from the viewpoint of technical inheritance of both hardware and software. Second, we discuss what plasma control system is relevant to ITER and fusion power plants from the two viewpoints: operation supervisor and plasma control tool. Finally, we will provide the current design of the ITER plasma control system.

Stabilization of interchange instabilities in the Phaedrus tandem mirror [Phys. Rev. Lett. 51, 1955 (1983)] through the use of externally applied rf at a single frequency is demonstrated over a wide range of frequency (1.03Ωi 1.3Ωi it is shown that the force acts primarily on electrons. Stable plasmas are only achieved when the net radial ponderomotive force is calculated to be stabilizing and comparable to the curvature force. Results are also reported for rf applied simultaneously at two frequencies, one which is stabilizing and one which is destabilizing. Abrupt changes in stability are observed as the rf power is increased whenever one of the rf frequencies is effective at plasma heating.

Fueling and heating of tandem mirror end cells by rf at the ion‐cyclotron frequency have been studied experimentally in the Phaedrus Tandem Mirror. The end cell density is found to increase linerly with rf voltage. The total plasma energy is observed to increase with rf power with no evidence of saturation at high power. The plasma axial length decreases with increasing rf power down to a length of approximately the distance between the two resonance locations in the end cell. The highest density and average ion energy are achieved with the resonance closest to the midplane. It is necessary to assume that the rf electric field at the resonance decreases with increasing density and with distance from the antenna in order to model the ion particle and power balance. The particle and power balance model predicts that dense, hot end cells may be maintained with E+∼1 V/cm and a fueling efficiency of a few percent when the resonance is close to the midplane because of better ion confinement and small plasma vol...