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The first real-time profile control experiments integrating magnetic and kinetic variables were performed on DIII-D in view of regulating and extrapolating advanced tokamak scenarios to steady-state devices and burning plasma experiments. Device-specific, control-oriented models were obtained from experimental data using a generic twotime- scale method that was validated on JET, JT-60U and DIII-D under the framework of the International Tokamak Physics Activity for Integrated Operation Scenarios (Moreau et al 2011 Nucl. Fusion 51 063009). On DIII-D, these data-driven models were used to synthesize integrated magnetic and kinetic profile controllers. The neutral beam injection (NBI), electron cyclotron current drive (ECCD) systems and ohmic coil provided the heating and current drive (H&CD) sources. The first control actuator was the plasma surface loop voltage (i.e. the ohmic coil), and the available beamlines and gyrotrons were grouped to form five additional H&CD actuators: co-current on-axis NBI, co-current off-axis NBI, counter-current NBI, balanced NBI and total ECCD power from all gyrotrons (with off-axis current deposition). Successful closed-loop experiments showing the control of (a) the poloidal flux profile, 𝞇(x), (b) the poloidal flux profile together with the normalized pressure parameter, ßN, and (c) the inverse of the safety factor profile, ῑ(x) = 1/q(x), are described. [ABSTRACT FROM AUTHOR]

The DIII-D device began operation in 1986, and a fully digital plasma control system (PCS) was implemented in 1993. Over time, the success of the PCS to exploit the inherent versatility of the DIII-D device led to a philosophy of using the PCS to control all available plasma system actuators. This made the PCS a very powerful physics tool that is at the core of Physics Operations at DIII-D. The complexity of the DIII-D device and all the systems the PCS must control makes the proper setup of the PCS for new experiments a daunting task. A cadre of physicists specially trained in PCS operation forms the bulk of the Physics Operations staff at DIII-D. They are the interface between experimental plans and successful execution and, as such, are a critical component of each experiment. Physics Operations is also a set of tools and procedures. We will briefly examine some of those tools, such as the TokSys control design and modeling environment and the "smart" PCS setup checklist, that greatly reduce human error in reconfiguring the PCS for a new experiment. We will examine the procedures that allow efficient use of those tools and some of the human factors that can affect productivity. [ABSTRACT FROM AUTHOR]

Recent commissioning of two major fully superconducting (SC)-shaped tokamaks, Experimental Advanced Superconducting Tokamak (EAST) and Korean Superconducting Tokamak Advanced Research (KSTAR), represents a significant advance in magnetic fusion research. The key to commissioning success in these complex and unique tokamaks was as follows: 1) use of a robust, flexible plasma control system (PCS) based on the validated DIII-D design; 2) use of the TokSys design and modeling environment, which is tightly coupled with the DIII-D PCS architecture for first-plasma scenario development and plasma diagnosis; and 3) collaborations with experienced internationally recognized teams of tokamak operations and control experts. We provide an overview of the generic modeling environment and plasma control tools developed and validated within the DIII-D experimental program and applied through an international collaborative program to successfully address the unique constraints associated with the startup of these next-generation tokamaks. The unique characteristics of each tokamak and the machine constraints that must be included in device modeling and simulation, such as SC coil current slew rate limits and the presence of nonlinear magnetic materials, are discussed, along with commissioning and initial operational results. Lessons learned from the startup experience in these devices are summarized, with special emphasis on ramifications for International Thermonuclear Experimental Reactor (ITER). [ABSTRACT FROM AUTHOR]

The achievement of suitable toroidal-current-density profiles in tokamak plasmas plays an important role in enabling high fusion gain and noninductive sustainment of the plasma current for steady-state operation with improved magnetohydrodynamic stability. The evolution in time of the current profile is related to the evolution of the poloidal magnetic flux, which is modeled in normalized cylindrical coordinates using a partial differential equation (PDE) usually referred to as themagnetic flux diffusion equation. The dynamics of the plasma current density profile can be modified by the total plasma current and the power of the noninductive current drive. These two actuators, which are constrained not only in value and rate but also in their initial and final values, are used to drive the current profile as close as possible to a desired target profile at a specific final time. To solve this constrained finite-time open-loop PDE optimal control problem, model reduction based on proper orthogonal decomposition is combined with sequential quadratic programming in an iterative fashion. The use of a low-dimensional dynamical model dramatically reduces the computational effort and, therefore, the time required to solve the optimization problem, which is critical for a potential implementation of a real-time receding-horizon control strategy. [ABSTRACT FROM AUTHOR]

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 diagnostic measurements. The duration of this state is limited by pressure profile evolution, leading to magnetohydrodynamic (MHD) instabilities after about 1 s or half of a current relaxation time (τCR). Stationary conditions are maintained in similar discharges (∼90% noninductive), limited only by the 2 s duration (1τCR) of the present ECCD systems. By discussing parametric scans in a global parameter and profile databases, the need for low density and high beta are identified to achieve full noninductive operation and good current drive alignment. These experiments achieve the necessary fusion performance and bootstrap fraction to extrapolate to the fusion gain, Q=5 steady-state scenario in the International Thermonuclear Experimental Reactor (ITER) [R. Aymar et al., Fusion Energy Conference on Controlled Fusion and Plasma Physics, Sorrento, Italy (International Atomic Energy Agency, Vienna, 1987), paper IAEA-CN-77/OV-1]. The modeling tools that have been successfully employed to both plan and interpret the experiment are used to plan future DIII-D experiments with higher power and longer pulse ECCD and fast wave and co- and counterneutral beam injection in a pumped double-null configuration. The models predict our ability to control the current and pressure profiles to reach full noninductivity with increased beta, bootstrap fraction, and duration. The same modeling tools are applied to ITER, predicting favorable prospects for the success of the ITER steady-state scenario. [ABSTRACT FROM AUTHOR]

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 conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. [ABSTRACT FROM AUTHOR]

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 P[sub threshold]<2.5 MW. Expansion of the ITB is enhanced compared to similar co-injected discharges. Both differences are believed to arise from modification of the ExB shear dynamics when the sign of the rotation contribution is reversed. Sustainment of an AT regime with β[sub N]H[sub 89]=9 for 16 confinement times has been accomplished in a discharge combining an ELMing H-mode (edge localized, high confinement mode) edge and an ITB, and exhibiting ion thermal transport down to 2-3 times neoclassical. The microinstabilities usually associated with ion thermal transport are predicted stable, implying that another mechanism limits performance. High frequency magnetohydrodynamic (MHD) activity is identified as the probable cause. © 2000 American Institute of Physics. [ABSTRACT FROM AUTHOR]

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 radial ion loss in the core plasma (r≤16 cm) but inward radial transport of ions is observed at a rate consistent with ion–neutral collisions and a radial electric field in the negative radial direction. Steady-state thermal-barrier-like potential dips that are in agreement with the Boltzmann model for potentials are observed in the transitions between the central cell and the end cells. [ABSTRACT FROM AUTHOR]

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 <ω<4.1Ωi). Here rf stabilized plasmas are reported with ion temperature up to 550 eV and beta up to 8%. Calculations of the rf electric fields show that the dominant mode is the m=+1 fast wave. When the rf frequency is near Ωi the calculated ponderomotive forces on ions and electrons are shown to be comparable. For ω>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. [ABSTRACT FROM AUTHOR]

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 linearly 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[sub +] ∼ 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 volume. Monte Carlo simulation of the trapping process shows that E²[sub +] ∞ T[sup 1.5, sub ic] is required to maintain a given rf trapping efficiency. [ABSTRACT FROM AUTHOR]