A. W. Hyatt

General Atomics, San Diego, California, United States

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Publications (290)356.58 Total impact

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    ABSTRACT: Non-rotating (`locked') magnetic islands often lead to complete losses of confinement in tokamak plasmas, called major disruptions. Here locked islands were suppressed for the first time, by a combination of applied three-dimensional magnetic fields and injected millimetre waves. The applied fields were used to control the phase of locking and so align the island O-point with the region where the injected waves generated non-inductive currents. This resulted in stabilization of the locked island, disruption avoidance, recovery of high confinement and high pressure, in accordance with the expected dependencies upon wave power and relative phase between O-point and driven current.
    Physical Review Letters 10/2015; 115(17). DOI:10.1103/PhysRevLett.115.175002 · 7.51 Impact Factor
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    ABSTRACT: The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.
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    ABSTRACT: A new approach has been experimentally demonstrated to control the stored energy by applying a non-axisymmetric magnetic field using the DIII-D in-vessel coils to modify the energy confinement time. In future burning plasma experiments as well as magnetic fusion energy power plants, various concepts have been proposed to control the fusion power. The fusion power in a power plant operating at high gain can be related to the plasma stored energy and hence, is a strong function of the energy confinement time. Thus, an actuator that modifies the confinement time can be used to adjust the fusion power. In relatively low collisionality DIII-D discharges, the application of non-axisymmetric magnetic fields results in a decrease in confinement time and density pumpout. Gas puffing was used to compensate the density pumpout in the pedestal while control of the stored energy was demonstrated by the application of non-axisymmetric fields.
    Nuclear Fusion 05/2015; 55(5):053001. DOI:10.1088/0029-5515/55/5/053001 · 3.06 Impact Factor
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    ABSTRACT: Sustaining the plasma in a stable and a high performance condition is one of the important control issues for future steady state tokamaks. In the 2014 KSTAR campaign, we have developed a real-time poloidal beta (βp) control technique and carried out preliminary experiments to identify its feasibility. In the control system, the βp is calculated in real time using the measured diamagnetic loop signal, and compared with the target value leading to the change of the neutral beam (NB) heating power using a feedback PID control algorithm. To match the requested power of NB which is operated with constant voltage, the on-time periods of the intervals were adjusted as the ratio of the required power to the maximum achievable one. This paper will present the overall procedures of the βp control, the βp estimation process and NB operation scheme implemented in the plasma control system (PCS), and the analysis on the preliminary experimental results.
    Fusion Engineering and Design 04/2015; 95. DOI:10.1016/j.fusengdes.2015.04.004 · 1.15 Impact Factor
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    ABSTRACT: DIII-D has made significant progress in developing the techniques required to operate ITER, and in understanding their impact on performance when integrated into operational scenarios at ITER-relevant parameters. Long duration plasmas, stable to m/n = 2/1 tearing modes, with an ITER-similar shape and I p/aB T, have been demonstrated in DIII-D, that evolve to stationary conditions. The operating region most likely to reach stable conditions has normalized pressure, β N ≈ 1.9–2.1 (compared to the ITER baseline design of 1.6–1.8), and a Greenwald normalized density fraction, f GW 0.42–0.70 (the ITER design is f GW ≈ 0.8). The evolution of the current profile, using internal inductance (l i) as an indicator, is found to produce a smaller fraction of stable pulses when l i is increased above ≈1.1 at the beginning of β N flattop. Stable discharges with co-neutral beam injection are generally accompanied with a benign n = 2 magnetohydrodynamic mode. However if this mode exceeds ≈10 G, the onset of a m/n = 2/1 tearing mode occurs with a loss of confinement. In addition, stable operation with low applied external torque, at or below the extrapolated value expected for ITER has also been demonstrated. With electron cyclotron injection, the operating region of stable discharges has been further extended at ITER equivalent levels of torque and to edge-localized mode (ELM) free discharges at higher torque but with the addition of an n = 3 magnetic perturbation from the DIII-D internal coil set. The characterization of the ITER baseline scenario evolution for long pulse duration, extension to more ITER-relevant values of torque and electron heating, and suppression of ELMs have significantly advanced the physics basis of this scenario, although significant effort remains in the simultaneous integration of all these requirements.
    Nuclear Fusion 02/2015; 55(2):023004. DOI:10.1088/0029-5515/55/2/023004 · 3.06 Impact Factor
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    ABSTRACT: Recent DIII-D experiments assessed the snowflake divertor (SF) configuration in a radiative regime in H-mode discharges with seeding. The SF configuration was maintained for many energy confinement times (2–3 s) in H-mode discharges ( MA, MW, and down (favorable direction toward the divertor)), and found to be compatible with high performance operation (H98y2 ⩾ 1). The two studied SF configurations, the SF-plus and the SF-minus, have a small finite distance between the primary X-point and the secondary null located in the private flux region or the common flux region, respectively. In H-mode discharges with the SF configurations (cf. H-mode discharges with the standard divertor with similar conditions) the stored energy lost per the edge localized mode (ELM) was reduced, and significant divertor heat flux reduction between and during ELMs was observed over a range of collisionalities, from lower density conditions toward a higher density H-modes with the radiative SF divertor.
    Journal of Nuclear Materials 12/2014; 463. DOI:10.1016/j.jnucmat.2014.12.052 · 1.87 Impact Factor
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    ABSTRACT: An accelerated fusion energy development program, a “fast-track” approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation, and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 m, κ = 2.3, BT = 5.4 T, IP = 6.6 MA, βN = 2.75, Pfus = 127 MW. The modest bootstrap fraction of ƒBS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ∼ 10 in ITER.
    Fusion Engineering and Design 10/2014; 89(7-8). DOI:10.1016/j.fusengdes.2014.03.055 · 1.15 Impact Factor
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    ABSTRACT: On DIII-D (Luxon 2005 Fusion Sci. Technol. 48 828), a high β scenario with minimum safety factor (qmin) near 1.4 has been optimized with new tools and shown to be a favourable candidate for long pulse or steady state operation in future devices. The new capability to redirect up to 5 MW of neutral beam injection (NBI) from on- to off-axis improves the ability to sustain elevated qmin with a less peaked pressure profile. These changes increase the ideal magnetohydrodynamics (MHD) n = 1 mode βN limit thus providing a path forward for increasing the noninductive current drive fraction by operating at high βN. Quasi-stationary discharges free of tearing modes have been sustained at βN = 3.5 and βT = 3.6% for two current profile diffusion timescales (about 3 s) limited by neutral beam duration. The discharge performance has normalized fusion performance expected to give fusion gain Q ≈ 5 in a device the size of ITER. Analysis of the poloidal flux evolution and current drive balance show that the loop voltage profile is almost relaxed even with 25% of the current driven inductively, and qmin remains elevated near 1.4. These observations increase confidence that the current profile will not evolve to one unstable to a tearing mode. In preliminary tests a divertor heat flux reduction technique based on producing a radiating mantle with neon injection appears compatible with this operating scenario. 0D model extrapolations suggest it may be possible to push this scenario up to 100% noninductive current drive by raising βN. Similar discharges with qmin = 1.5–2 were susceptible to tearing modes and off-axis fishbones, and with qmin > 2 lower normalized global energy confinement time is observed.
    Nuclear Fusion 08/2014; 54(9):093009. DOI:10.1088/0029-5515/54/9/093009 · 3.06 Impact Factor
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    ABSTRACT: Magnetic feedback control of the resistive-wall mode has enabled the DIII-D tokamak to access stable operation at safety factor q_{95}=1.9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at a given toroidal magnetic field. In tokamaks with a divertor, the limit occurs at q_{95}=2, as confirmed in DIII-D. Since the energy confinement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a whole new high-current regime not accessible before. This result brings significant possible benefits in terms of fusion performance, but it also extends resistive-wall mode physics and its control to conditions never explored before. In present experiments, the q_{95}<2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.
    Physical Review Letters 07/2014; 113(4):045003. DOI:10.1103/PhysRevLett.113.045003 · 7.51 Impact Factor
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    ABSTRACT: Recent DIII-D and RFX-mod experiments have demonstrated stable tokamak operation at very low values of the edge safety factor q(a) near and below 2. The onset of n = 1 resistive wall mode (RWM) kink instabilities leads to a disruptive stability limit, encountered at q(a) = 2 (limiter plasmas) and q 95 = 2 (divertor plasmas). However, passively stable operation can be attained for q(a) and q 95 values as low as 2.2. RWM damping in the q(a) = 2 regime was measured using active MHD spectroscopy. Although consistent with theoretical predictions, the amplitude of the damped response does not increase significantly as the q(a) = 2 limit is approached, in contrast with damping measurements made approaching the pressure-driven RWM limit. Applying proportional gain magnetic feedback control of the n = 1 modes has resulted in stabilized operation with q 95 values reaching as low as 1.9 in DIII-D and q(a) reaching 1.55 in RFX-mod. In addition to being consistent with the q(a) = 2 external kink mode stability limit, the unstable modes have growth rates on the order of the characteristic wall eddy-current decay timescale in both devices, and a dominant m = 2 poloidal structure that is consistent with ideal MHD predictions. The experiments contribute to validating MHD stability theory and demonstrate that a key tokamak stability limit can be overcome with feedback.
    Physics of Plasmas 07/2014; 21(7):072107. DOI:10.1063/1.4886796 · 2.14 Impact Factor
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    ABSTRACT: Superconducting tokamaks like KSTAR, EAST and ITER need elaborate magnetic controls mainly due to either the demanding experiment schedule or tighter hardware limitations caused by the superconducting coils. In order to reduce the operation runtime requirements, two types of plasma simulators for the KSTAR plasma control system (PCS) have been developed for improving axisymmetric magnetic controls. The first one is an open-loop type, which can reproduce the control done in an old shot by loading the corresponding diagnostics data and PCS setup. The other one, a closed-loop simulator based on a linear nonrigid plasma model, is designed to simulate dynamic responses of the plasma equilibrium and plasma current (Ip) due to changes of the axisymmetric poloidal field (PF) coil currents, poloidal beta, and internal inductance. The closed-loop simulator is the one that actually can test and enable alteration of the feedback control setup for the next shot. The simulators have been used routinely in 2012 plasma campaign, and the experimental performances of the axisymmetric shape control algorithm are enhanced. Quality of the real-time EFIT has been enhanced by utilizations of the open-loop type. Using the closed-loop type, the decoupling scheme of the plasma current control and axisymmetric shape controls are verified through both the simulations and experiments. By combining with the relay feedback tuning algorithm, the improved controls helped to maintain the shape suitable for longer H-mode (10–16 s) with the number of required commissioning shots largely reduced.
    Fusion Engineering and Design 05/2014; 89(5). DOI:10.1016/j.fusengdes.2013.12.040 · 1.15 Impact Factor
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    ABSTRACT: A long-pulse high confinement plasma regime known as H-mode is achieved in the Experimental Advanced Superconducting Tokamak (EAST) with a record duration over 30 s, sustained by Lower Hybrid wave Current Drive (LHCD) with advanced lithium wall conditioning and divertor pumping. This long-pulse H-mode plasma regime is characterized by the co-existence of a small Magneto-Hydrodynamic (MHD) instability, i.e., Edge Localized Modes (ELMs) and a continuous quasi-coherent MHD mode at the edge. We find that LHCD provides an intrinsic boundary control for ELMs, leading to a dramatic reduction in the transient power load on the vessel wall, compared to the standard Type I ELMs. LHCD also induces edge plasma ergodization, broadening heat deposition footprints, and the heat transport caused by ergodization can be actively controlled by regulating edge plasma conditions, thus providing a new means for stationary heat flux control. In addition, advanced tokamak scenarios have been newly developed for high-performance long-pulse plasma operations in the next EAST experimental campaign.
    Physics of Plasmas 05/2014; 21(5):056107. DOI:10.1063/1.4872195 · 2.14 Impact Factor
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    ABSTRACT: Mueller-Stokes theory can be used to calculate the polarization evolution of an electromagnetic (EM) wave as it propagates through a magnetized plasma. Historically, the theory has been used to interpret polarimeter signals from systems operating on fusion plasmas. These interpretations have mostly employed approximations of Mueller-Stokes theory in regimes where either the Faraday rotation (FR) or the Cotton-Mouton (CM) effect is dominant. The current paper presents the first systematic comparison of polarimeter measurements with the predictions of full Mueller-Stokes theory where conditions transition smoothly from a FR-dominant (i.e., weak CM effect) plasma to one where the CM effect plays a significant role. A synthetic diagnostic code, based on Mueller-Stokes theory accurately reproduces the trends evident in the experimentally measured polarimeter phase over this entire operating range, thereby validating Mueller-Stokes theory. The synthetic diagnostic code is then used to investigate the influence of the CM effect on polarimetry measurements. As expected, the measurements are well approximated by the FR effect when the CM effect is predicted to be weak. However, the code shows that as the CM effect increases, it can compete with the FR effect in rotating the polarization of the EM-wave. This results in a reduced polarimeter response to the FR effect, just as observed in the experiment. The code also shows if sufficiently large, the CM effect can even reverse the handedness of a wave launched with circular polarization. This helps to understand the surprising experimental observations that the sensitivity to the FR effect can be nearly eliminated at high enough BT (2.0 T). The results also suggest that the CM effect on the plasma midplane can be exploited to potentially measure magnetic shear in tokamak plasmas. These results establish increased confidence in the use of such a synthetic diagnostic code to guide future polarimetry design and interpret the resultant experimental data.
    Physics of Plasmas 10/2013; 20(10):2519-. DOI:10.1063/1.4826956 · 2.14 Impact Factor
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    ABSTRACT: Since the first H-mode discharges in 2010, the duration of the H-mode state has been extended and a significantly wider operational window of plasma parameters has been attained. Using a second neutral beam (NB) source and improved tuning of equilibrium configuration with real-time plasma control, a stored energy of W-tot similar to 450 kJ has been achieved with a corresponding energy confinement time of tau(E) similar to 163 ms. Recent discharges, produced in the fall of 2012, have reached plasma beta(N) up to 2.9 and surpassed the n = 1 ideal no-wall stability limit computed for H-mode pressure profiles, which is one of the key threshold parameters defining advanced tokamak operation. Typical H-mode discharges were operated with a plasma current of 600 kA at a toroidal magnetic field B-T = 2 T. L-H transitions were obtained with 0.8-3.0MW of NB injection power in both single-and double-null configurations, with H-mode durations up to similar to 15 s at 600 kA of plasma current. The measured power threshold as a function of line-averaged density showed a roll-over with a minimum value of similar to 0.8 MW at (n) over bar (e) similar to 2 x 10(19) m(-3). Several edge-localized mode (ELM) control techniques during H-mode were examined with successful results including resonant magnetic perturbation, supersonic molecular beam injection (SMBI), vertical jogging and electron cyclotron current drive injection into the pedestal region. We observed various ELM responses, i.e. suppression or mitigation, depending on the relative phase of in-vessel control coil currents. In particular, with the 90 degrees phase of the n = 1 RMP as the most resonant configuration, a complete suppression of type-I ELMs was demonstrated. In addition, fast vertical jogging of the plasma column was also observed to be effective in ELM pace-making. SMBI-mitigated ELMs, a state of mitigated ELMs, were sustained for a few tens of ELM periods. A simple cellular automata ('sand-pile') model predicted that shallow deposition near the pedestal foot induced small-sized high-frequency ELMs, leading to the mitigation of large ELMs. In addition to the ELM control experiments, various physics topics were explored focusing on ITER-relevant physics issues such as the alteration of toroidal rotation caused by both electron cyclotron resonance heating (ECRH) and externally applied 3D fields, and the observed rotation drop by ECRH in NB-heated plasmas was investigated in terms of either a reversal of the turbulence-driven residual stress due to the transition of ion temperature gradient to trapped electron mode turbulence or neoclassical toroidal viscosity (NTV) torque by the internal kink mode. The suppression of runaway electrons using massive gas injection of deuterium showed that runaway electrons were avoided only below 3 T in KSTAR. Operation in 2013 is expected to routinely exceed the n = 1 ideal MHD no-wall stability boundary in the long-pulse H-mode (>= 10 s) by applying real-time shaping control, enabling n = 1 resistive wall mode active control studies. In addition, intensive works for ELM mitigation, ELM dynamics, toroidal rotation changes by both ECRH and NTV variations, have begun in the present campaign, and will be investigated in more detail with profile measurements of different physical quantities by techniques such as electron cyclotron emission imaging, charge exchange spectroscopy, Thomson scattering and beam emission spectroscopy diagnostics.
    Nuclear Fusion 10/2013; 53(10):104005. DOI:10.1088/0029-5515/53/10/104005 · 3.06 Impact Factor
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    ABSTRACT: The initial experiments on off-axis neutral beam injection into high noninductive current fraction (fNI), high normalized pressure (βN) discharges in DIII-D [J. L. Luxon, Fusion Sci. Technol. 48, 828 (2005)] have demonstrated changes in the plasma profiles that increase the limits to plasma pressure from ideal low-n instabilities. The current profile is broadened and the minimum value of the safety factor (qmin) can be maintained above 2 where the profile of the thermal component of the plasma pressure is found to be broader. The off-axis neutral beam injection results in a broadening of the fast-ion pressure profile. Confinement of the thermal component of the plasma is consistent with the IPB98(y,2) scaling, but global confinement with qmin>2 is below the ITER-89P scaling, apparently as a result of enhanced transport of fast ions. A 0-D model is used to examine the parameter space for fNI=1 operation and project the requirements for high performance steady-state discharges. Fully noninductive solutions are found with 4<βN<5 and bootstrap current fraction near 0.5 for a weak shear safety factor profile. A 1-D model is used to show that a fNI=1 discharge at the top of this range of βN that is predicted stable to n=1, 2, and 3 ideal MHD instabilities is accessible through further broadening of the current and pressure profiles with off-axis neutral beam injection and electron cyclotron current drive.
    Physics of Plasmas 09/2013; 20(9):2504-. DOI:10.1063/1.4821072 · 2.14 Impact Factor
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    ABSTRACT: We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L||, and the radial location of the outer divertor target, RTAR, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L|| and RTAR should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., nTAR ∝ [RTAR]2[L||]6/7 and TTAR ∝ [RTAR]−2[L||]−4/7. The dependence of nTAR and TTAR on L|| was consistent with our data, but the dependence of nTAR and TTAR on RTAR was not. The surprising result that the divertor plasma parameters did not depend on RTAR in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger RTAR cases. Modeling results using the SOLPS code support this postulate.
    Journal of Nuclear Materials 07/2013; 438(11):S166–S169. DOI:10.1016/j.jnucmat.2013.01.051 · 1.87 Impact Factor
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    ABSTRACT: 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 two-time- 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, Psi(x), (b) the poloidal flux profile together with the normalized pressure parameter, beta(N), and (c) the inverse of the safety factor profile, (i) over bar (x) = 1/q(x), are described.
    Nuclear Fusion 06/2013; 53(6):063020. DOI:10.1088/0029-5515/53/6/063020 · 3.06 Impact Factor
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    ABSTRACT: The DIII-D Plasma Control System (PCS [1]), initially deployed in the early 1990's, now controls nearly all aspects of the tokamak and plasma environment. Versions of this PCS, supported by General Atomics, are presently used to control several tokamaks around the world, including the superconducting tokamaks EAST and KSTAR. The experimental challenges posed by the advanced tokamak mission of DIII-D and the variety of devices supported by the PCS have driven the development of a rich array of control algorithms, along with a powerful set of tools for algorithm design and testing. Broadly speaking, the PCS mission is to utilize all available sensors, measurements and actuators to safely produce a plasma state trajectory leading to and then maintaining the desired experimental conditions. Often new physics understanding leads to new or modified control requirements that use existing actuators in new ways. We describe several important DIII-D PCS design and test tools that support implementation and optimization of algorithms. We describe selected algorithms and the ways they fit within the PCS architecture, which in turn allows great flexibility in designing, constructing and using the algorithms to reliably produce a desired complex experimental environment. Control algorithms, PCS interfaces, and design and testing tools are described from the perspective of the physics operator (PO), who must operate the PCS to achieve experimental goals and maximize physics productivity of the tokamak. For example, from a PO's (and experimental team leader's) standpoint, a PCS algorithm interface that offers maximum actuator, algorithmic and measurement configuration flexibility is most likely to produce a successful experimental outcome. However, proper constraints that limit flexibility in use of the PCS can also help to maximize effectiveness. For example, device limits and safety must be built into the PCS, sometimes at the algorithm level. We show how the D3D PCS toolset - nables rapid offline testing of a new or modified algorithm in a simulated tokamak environment. Finally, we illustrate usage of PCS-based checklists and procedures that enhance experimental productivity and we describe an asynchronous condition detector system within the PCS that enhances device safety and enables complex experiment design.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 06/2013
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    ABSTRACT: In this paper, a linear model for plasma current, position and shape control based on the plasma rigid motion assumption is presented and implemented in an EAST tokamak simulator. The simulator models the plasma, poloidal field coils, and power supplies, and is used to verify the control algorithm and optimize control parameters and poloidal field coil current trajectories. Plasma position and shape control has been achieved during the last several EAST operation campaigns due to successful decoupling of plasma current, plasma position and shape. The control logic used and experimental results are described in detail. Diverted plasma shapes, including double null, upper and lower single null, and with elongation up to 2.0, triangularity in the range 0.4-0.6 and X point control accuracy of 1 cm, were successfully controlled. Smooth shape transition in the current ramp up ensures that volt-seconds are saved and that plasma disruptions are avoided. Such control capability provides the basis for future high performance plasma operation.
    Nuclear Fusion 03/2013; 53(4):043009. DOI:10.1088/0029-5515/53/4/043009 · 3.06 Impact Factor
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    ABSTRACT: A number of important software and hardware changes have recently been made to the DIII-D Plasma Control System (PCS) to further its capabilities in support of fusion research. The PCS is a highly customizable real-time control application developed at General Atomics used to manage the many parameters that affect plasmas produced on the DIII-D tokamak. Included in the most recent updates to the PCS are refinements to the real-time Electron Cyclotron Heating (ECH) capabilities which have improved overall performance and reliability for fast and precise aiming of the mirrors used to control direct ECH power into the plasma. The introduction of new real-time streaming data acquisition hardware has provided a means for acquiring plasma electron temperatures and densities from the Thomson scattering System along with data from the Electron Cyclotron Emission (ECE) diagnostic for use in PCS feedback control algorithms. The new fiber optically connected streaming digitizers allow PCS computers located in one part of the tokamak facility to easily communicate with remotely located diagnostic systems in other parts of the lab, in addition to being able to transfer high frequency data (sampled at 500 Hz) for a large number of channels in real-time. Details of the most recent PCS enhancements will be provided along with a more thorough description of the latest software and hardware architecture.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013

Publication Stats

3k Citations
356.58 Total Impact Points


  • 1991-2015
    • General Atomics
      San Diego, California, United States
  • 2002
    • University of California, San Diego
      • Department of Mechanical and Aerospace Engineering (MAE)
      San Diego, California, United States
  • 1998
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, NJ, United States
    • Lawrence Livermore National Laboratory
      • Physics Division
      Livermore, California, United States
  • 1994
    • Oak Ridge National Laboratory
      Oak Ridge, Florida, United States
  • 1989
    • University of California, Los Angeles
      • Department of Mechanical and Aerospace Engineering
      Los Angeles, CA, United States