A. W. Hyatt

General Atomics, San Diego, California, United States

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Publications (277)325.81 Total impact

  • [Show abstract] [Hide abstract]
    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; · 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. · 3.24 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. · 7.73 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; · 1.15 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-. · 2.25 Impact Factor
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    Nuclear Fusion 10/2013; 53(10):104005. · 3.24 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-. · 2.25 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:S166–S169. · 2.02 Impact Factor
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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; 53:043009. · 3.24 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
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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; 01/2013
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    ABSTRACT: In recent 2 years, various algorithms to control plasma shape, current and density have been implemented or improved for EAST tokamak. These plasma control performances have been verified by either simulated or actual experimental operation, and thus plasma control basis has been established for the long pulse operation and high performance H-mode plasma operation with low hybrid wave (LHW) and ion cyclotron resonance frequency (ICRF) heating. Startup simulation has been done by using TOKSYS code for the plasma breakdown in either 3.1 Wb or 4.5 Wb initial poloidal flux state and the scenarios proved to be robust and used for routine operation. Various shape configurations have been well feedback controlled by using ISOFLUX limited, double-null or single null algorithms based on RTEFIT equilibrium reconstruction. For the long pulse operation, strike point control and magnetics drift compensation have been implemented in the plasma control system (PCS). To improve the operation safety and efficiency, the verification of magnetic diagnostics before plasma breakdown has been demonstrated adequate to prevent a discharge in case of key sensor failure.
    Fusion Engineering and Design 12/2012; 87(12):1887–1890. · 1.15 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 and these data-driven models were used to synthesize integrated magnetic and kinetic profile controllers. Closed-loop experiments were performed for the regulation of (a) the poloidal flux profile, Ψ(x), (b) the inverse of the safety factor profile, ι(x)=1/q(x), and (c) either the Ψ(x) profile or the ι(x) profile together with the normalized pressure parameter, β N . The neutral beam injection (NBI), electron cyclotron current drive (ECCD) systems and ohmic coils provided the heating and current drive (H&CD) sources. The first control actuator was the plasma surface loop voltage or current (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). The control method was also applied on simulated ITER discharges using a simplified transport code (METIS).
    24th IAEA Fusion Energy Conference; 10/2012
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    ABSTRACT: Plasmas stable to m/n=2/1 tearing modes (TMs), in the ITER baseline scenario (IBS) with ITER similar Ip/aBT, have been demonstrated in DIII-D, evolving to stationary conditions. Previous studies showed the possibility that long pulse IBS plasmas might be susceptible to TMs. However within a defined stability boundary, most of these longer duration discharges have achieved stationary conditions (δτduration<=7.5 s and <=11τR) with high Co-beam torque and without the need for ECCD. To mitigate 2/1 TMs at reduced torque, broad ECCD deposition was found to be most effective when positioned near the q=3/2 flux surface, although a subset of low torque pulses were also obtained without ECCD. The DIII D internal coils (I coils) were used to achieve ELM suppression in IBS plasmas with ECCD at q95=3.15 for durations up to 1 s with only the upper row of I coils, providing a broad n=3 spectrum. Conditions stable to n=1 tearing modes in IBS discharges and the effect of Zeff, density, and other parameters are discussed.
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    ABSTRACT: The spatial distribution of runaway electron (RE) strikes to the wall during argon pellet-initiated rapid shutdown of diverted and limited plasma shapes in DIII-D is studied using a new array of hard x-ray (HXR) scintillators. Two plasma configurations were investigated: an elongated diverted H-mode and a low-elongation limited L-mode. HXR emission from MeV level REs generated during the argon pellet injection is observed during the thermal quench (TQ) in diverted discharges from REs lost into the divertor. In limiter discharges, this prompt TQ loss is reduced, suggesting improved TQ confinement of REs in this configuration. During the plateau phase when the plasma current is carried by REs, toroidally symmetric HXR emission from remaining confined REs is seen. Transient HXR bursts during this RE current plateau suggest the presence of a small level of wall losses due to the presence of an unidentified instability. Eventually, an abrupt final loss of the remaining RE current occurs. This final loss HXR emission shows a strong toroidal peaking and a consistent spatiotemporal evolution that suggests the development of a kink instability.
    Nuclear Fusion 01/2012; 52(1). · 3.24 Impact Factor
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    ABSTRACT: High-power electron cyclotron (EC) waves are used to increase performance in several Advanced Tokamak (AT) regimes on DIII-D where there is a simultaneous need for high noninductive current and high beta. In the Quiescent High-confinement mode (QH-mode), a direct measurement of the electron cyclotron current drive (ECCD) profile is made using modulation techniques, and a trapped electron mode (TEM) dominated regime with core T{sub e}>T{sub i} is created. In the 'highq{sub min}' AT scenario, ECCD provides part of the off-axis noninductive current and helps to produce a tearing stable equilibrium. In the hybrid regime, strong central current drive from EC waves and other sources increases the noninductive current fraction to {approx_equal}100%. Surprisingly, the core safety factor remains above unity, meaning good alignment between the current drive profile and the desired plasma current profile is not necessary in this scenario.
    AIP Conference Proceedings. 12/2011; 1406(1).
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    ABSTRACT: Future generation tokamaks that produce significant fusion power will require a means of reducing damaging levels of both transient (ELM-induced) and steady power loads at the divertor targets. We report here on experiments at DIII-D that examine how certain variations in the divertor geometry affect both the capability to reduce heat flux at the outer divertor target and the H-mode quality of the main plasma. We focus specifically on documenting how core and divertor plasmas respond to changes in both the parallel path length of the outer divertor leg and the radial location of the outer divertor strike point, both in ``standard" ELMing H-mode (without gas puffing) and in radiating divertor. We also investigate how these geometric changes may mitigate transient ELM-induced heat fluxes.
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    ABSTRACT: Disruptivity in ITER must be minimized to limit downtime and maximize use of the limited number of discharges. Minimizing disruptivity requires sufficient control capability, including robustness to disturbances and disruption avoidance through prediction of controllability limits. Robust control implies a balance of passively stable nominal scenarios, robust operation near or beyond open loop stability limits, and responses to off-normal events to avoid disruptive termination. Such a solution is possible because disruptions result from deterministic loss of controllability due to many proximal causes (e.g. loss of hardware resources, human error, or uncontrollable disturbances), most of which can be addressed with good physics models and known control methods. We illustrate the required approach with DIII-D experiments to assess ITER controllability and pre-qualify ITER scenarios, and with design and analysis ensuring sufficiently robust vertical control for ITER.
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    ABSTRACT: Two of the eight neutral beam sources on DIII-D have been modified to allow vertical steering, with the injection angle varying from horizontal to downward at an angle of 16.5 degrees for off-axis deposition. Initial experiments to assess the basic beam functionality, geometry, and confinement were carried out. Dα images of beam into gas and plasma yield beam neutral profiles and are key in assessing beam shape and clipping. Neutron and fast-ion Dα (FIDA) diagnostics verify classical behavior of the off-axis beam ions in MHD-quiescent conditions. An initial physics experiment takes advantage of the downward steered beams to vary the fast-ion gradient ∇βf from centrally peaked to hollow. Systematic scans determine the stability and impact of reversed shear Alfvén eigenmodes and toroidal Alfvén eigenmodes as a function of ∇βf.

Publication Stats

2k Citations
325.81 Total Impact Points


  • 1991–2014
    • General Atomics
      San Diego, California, United States
  • 2010
    • Northeast Institute of Geography and Agroecology
      • Institute of Plasma Physics
      Beijing, Beijing Shi, China
  • 1994–2007
    • University of California, San Diego
      • Department of Mechanical and Aerospace Engineering (MAE)
      San Diego, CA, United States
  • 2002
    • Columbia University
      • Department of Applied Physics and Applied Mathematics
      New York City, NY, United States
  • 1998
    • Lawrence Livermore National Laboratory
      • Physics Division
      Livermore, California, United States
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, NJ, United States
  • 1993
    • Oak Ridge National Laboratory
      • Fusion Energy Division
      Oak Ridge, FL, United States
  • 1989
    • University of California, Los Angeles
      • Department of Mechanical and Aerospace Engineering
      Los Angeles, CA, United States