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The Global Network of Optical Magnetometers to search for Exotic physics (GNOME) is a network of geographically separated, time-synchronized, optically pumped atomic magnetometers that is being used to search for correlated transient signals heralding exotic physics. The GNOME is sensitive to nuclear- and electron-spin couplings to exotic fields from astrophysical sources such as compact dark-matter objects (for example, axion stars and domain walls). Properties of the GNOME sensors such as sensitivity, bandwidth, and noise characteristics are studied in the present work, and features of the network's operation (e.g., data acquisition, format, storage, and diagnostics) are described. Characterization of the GNOME is a key prerequisite to searches for and identification of exotic physics signatures., Comment: 45 pages, 16 figures, 2 tables

Results are reported from the first full-scale search for transient signals from exotic fields of astrophysical origin using data from a newly constructed Earth-scale detector: the Global Network of Optical Magnetometers for Exotic physics searches (GNOME). Data collected by the GNOME consist of correlated measurements from optical atomic magnetometers located in laboratories all over the world. GNOME data are searched for patterns of signals propagating through the network consistent with exotic fields composed of ultralight bosons such as axion-like particles (ALPs). Analysis of data from a continuous month-long operation of the GNOME finds no statistically significant signals consistent with those expected due to encounters with topological defects (axion domain walls), placing new experimental constraints on such dark matter scenarios.

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.

The systems responsible for the archival and access of data at ITER will play a critical role from the very start of tokamak operations. Ensuring the readiness of the software being developed to handle the large amounts of data to be streamed to the archive systems will be of great importance and require extensive testing. To assist in this validation DIII-D offers the flexibility and means for testing on an operational tokamak. Because of the extensive experience gained in supporting data archiving and retrieval for a large and varied set of diagnostics and in-house expertise in Linux based systems, DIII-D is able to provide a uniquely qualified test platform to fully exercise ITER's software. In this paper, we report on the work that was recently performed to setup a data acquisition and archiving test system connected to live diagnostics for the purpose of evaluating the performance and completeness of the ITER Data Archiving Network Application Program Interface (DAN API). Diagnostic data that was made available and examined for archiving into the ITER DAN included data such as plasma current, bolometry, electron density, soft x-ray signals, visible camera images, data from the Thomson scattering and Electron Cyclotron Heating (ECH) systems. The hardware used for the test system was comprised of two high performing archiver computers and two data acquisition streamer computers. The archiver nodes were configured with 64 GB of RAM and 1 TB of fast solid state drive storage, and networked to the acquisition computers in a local 10 Gb/s network. This configuration was chosen in order to push performance limits and test data streaming and archiving to the ITER DAN at rates of up to 1 Gb/s from multiple archive sources.

A control-oriented, two-timescale, linear, dynamic, response model of the rotational transform ι profile and the normalized beta βN is proposed based on experimental data from the DIII-D tokamak. Dedicated system-identification experiments without feedback control have been carried out to generate data for the development of this model. The data-driven dynamic model, which is both device-specific and scenario-specific, represents the response of the ι profile and βN to the electric field due to induction as well as to the heating and current drive (H&CD) systems during the flat-top phase of an H-mode discharge in DIII-D. The control goal is to use both induction and the H&CD systems to locally regulate the plasma ι profile and βN around particular target values close to the reference state used for system identification. A singular value decomposition (SVD) of the plasma model at steady state is carried out to decouple the system and identify the most relevant control channels. A mixed-sensitivity robust control design problem is formulated based on the dynamic model to synthesize a stabilizing feedback controller without input constraints that minimizes the reference tracking error and rejects external disturbances with minimal control energy. The feedback controller is then augmented with an anti-windup compensator, which keeps the given controller well-behaved in the presence of magnitude constraints in the actuators and leaves the nominal closed-loop system unmodified when no saturation is present. The proposed controller represents one of the first feedback profile controllers integrating magnetic and kinetic variables ever implemented and experimentally tested in DIII-D. The preliminary experimental results presented in this work, although limited in number and constrained by actuator problems and design limitations, as it will be reported, show good progress towards routine current profile control in DIII-D and leave valuable lessons for further advancements in the field.

The plasma control system (PCS) has been one of essential systems in annual KSTAR plasma campaigns: starting from a single-process version in 2008, extensive upgrades are done through the previous 7 years in order to achieve major goals of KSTAR performance enhancement. Major implementations are explained in this paper. In consequences of successive upgrades, the present KSTAR PCS is able to achieve ∼48 s of 500 kA plasma pulses with full real-time shaping controls and real-time NB power controls. It has become a huge system capable of dealing with 8 separate categories of algorithms, 26 actuators directly controllable during the shot, and real-time data communication units consisting of +180 analog channels and +600 digital input/outputs through the reflective memory (RFM) network. The next upgrade of the KSTAR PCS is planned in 2015 before the campaign. An overview of the upgrade layout will be given for this paper. The real-time system box is planned to use the CERN MRG-Realtime OS, an ITER-compatible standard operating system. New hardware is developed for faster real-time streaming system for future installations of actuators/diagnostics.

The DIII-D Plasma Control System (PCS) is a comprehensive software and hardware system used in real-time data acquisition and feedback control of numerous actuators on the DIII-D tokamak. It regulates many plasma characteristics including shape, position, divertor function, and core performance. The custom software developed at DIII-D provides an expandable platform from which new control algorithms can be incorporated. PCS has been expanding with the needs of the DIII-D research program, national, and international institutions that have adapted the PCS for use on their devices. The DIII-D PCS group in collaboration with many national and international groups have been instrumental in steadily improving the effectiveness and capability of the system. Enhancements on two key areas have been made: computer infrastructure upgrades and real time diagnostics control.

One of the most promising candidates to produce clean nuclear fusion energy is the tokamak, a device magnetically confining an extremely hot plasma (i.e. an ionized gas) where the fusion reactions take place. To produce nuclear fusion energy using tokamak devices, it is crucial that the poloidal magnetic flux (characterized by the so-called q profile) and the plasma internal energy are tightly controlled to avoid magnetohydrodynamic instabilities and to reach the high pressures and temperatures that are needed for high fusion-power density. Simultaneous control of the q profile and the internal energy is challenging for a number of reasons: the system is nonlinear, there are significant parameter uncertainties and large disturbances, the available number of actuators is small, and the actuation authority is limited from a control perspective. Therefore, a variable-structure controller is proposed in this work to tackle this plasma control problem since this type of controllers can typically diminish the impact of serious disturbances and nonlinearities while still leading to good performance. Simulations and recent experiments on the DIII-D tokamak in a challenging high-confinement (H-mode) plasma regime show that this control approach does indeed lead to good and repeatable control of the q profile and the internal energy.

One of the most promising devices for realizing power production through nuclear fusion is the tokamak. To maximize performance, it is preferable that tokamak reactors achieve advanced operating scenarios characterized by good plasma confinement, improved magnetohydrodynamic stability, and a largely noninductively driven plasma current. Such scenarios could enable steady-state reactor operation with high fusion gain, the ratio of produced fusion power to the external power provided through the plasma boundary. For certain advanced scenarios, control of the spatial profile of the plasma current will be essential. The complexity of the current profile dynamics, arising due to nonlinearities and couplings with many other plasma parameters, motivates the use of model-based control algorithms that can account for the system dynamics. A first-principles-driven, control-oriented model of the current profile evolution in low-confinement mode (L-mode) discharges in the DIII-D tokamak is employed to address the problem of regulating the current profile evolution around desired trajectories. In the primarily inductive L-mode discharges considered in this paper, the boundary condition, which is dependent on the total plasma current, has the largest influence on the current profile dynamics, motivating the design of a boundary feedback control law to improve the system performance. The backstepping control design technique provides a systematic method to obtain a boundary feedback law through the transformation of a spatially discretized version of the original system into an asymptotically stable target system with desirable properties. Through a nonlinear transformation of the available physical actuators, the resulting control scheme produces references for the total plasma current, total power, and line averaged density, which are tracked by existing dedicated control loops. Adaptiveness is added to the control scheme to improve upon the backstepping controller's disturbance rejection and tracking capability. Prior to experimental testing, a Simserver simulation was carried out to study the controller's performance and ensure proper implementation in the DIII-D Plasma Control System. An experimental test was performed on DIII-D to test the ability of the controller to reject input disturbances and perturbations in initial conditions and to demonstrate the feasibility of the proposed control approach.

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.

The DIII-D plasma control system (PCS), initially deployed in the early 1990s, 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 Experimental Advanced Superconducting Tokamak and Korean Superconducting Tokamak Advanced Research. 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 POs (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 DIII-D PCS toolset enables 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.

Extensive research has been conducted to find operating scenarios that optimize the plasma performance in nuclear fusion tokamak devices with the goal of enabling the success of the ITER project. The development, or planning, of these advanced scenarios is traditionally investigated experimentally by modifying the tokamak's actuator trajectories, such as the auxiliary heating/current-drive (H&CD) scheme, and analyzing the resulting plasma evolution. In this work, a numerical optimization algorithm is developed to complement the experimental effort of advanced scenario planning in the DIII-D tokamak. Two properties related to the plasma stability and performance are the safety factor profile ( q -profile) and the normalized plasma beta ( β N ). The optimization algorithm goal is to design actuator trajectories that steer the plasma to a target q -profile and plasma β N , such that the achieved state is stationary in time, subject to the plasma dynamics (described by a physics-based, nonlinear, control-oriented partial differential equation model) and practical plasma state and actuator constraints, such as the maximum available amount of H&CD power. This defines a nonlinear, constrained optimization problem that we solve by employing sequential quadratic programming. The optimized trajectories are then tested through simulation with the physics-based model and experimentally in DIII-D.

Efficient, high-gain operation of a tokamak device requires the achievement of certain radial shapes for the toroidal current profile. The evolution in time of the toroidal current profile in tokamaks is related to the evolution of the poloidal magnetic flux profile. A model-based control approach for the regulation of the poloidal magnetic flux profile at the DIII-D tokamak is proposed in this work. The model describing the poloidal flux evolution is based on a control-oriented formulation of the magnetic diffusion equation. Auxiliary heating and current drive (H&CD) systems including electron cyclotron (EC) and neutral beam injection (NBI) along with the total plasma current are used as actuators to manipulate the profile shape. Optimal state feedback control with integral action is used to design a controller to regulate the profile around a target while rejecting disturbances. Combining the profile controller with control of the plasma stored energy is found to improve tracking performance. Simulations and experimental results are presented to demonstrate the controller's effectiveness.

Active closed-loop control of the plasma safety factor profile ( q -profile) and internal energy dynamics in nuclear fusion tokamak devices has the potential to significantly impact the success of the ITER project. These plasma properties are related to both the stability and performance of a given plasma operating scenario. In this work, we develop integrated feedback control algorithms to control the q -profile and internal energy dynamics in DIII-D advanced tokamak (high performance) scenarios. The feedback controllers are synthesized by embedding a nonlinear, physics-based, control-oriented partial differential equation model of the plasma dynamics into the control design and to be robust to uncertainties in the plasma electron density, electron temperature, and plasma resistivity profiles. The auxiliary heating and current-drive system and the total plasma current are the actuators utilized by the feedback controllers to control the plasma dynamics. Finally, the feedback controllers are tested both through simulations based on the physics-based model and experimentally in the DIII-D tokamak.

The development and operation of the neoclassical tearing mode (NTM) avoidance and control system for DIII-D, which uses six sets of real-time steerable mirrors in order to move the electron cyclotron current drive (ECCD) deposition location in plasma, is described. The real-time DIII-D NTM control algorithm residing in the Plasma Control System (PCS) automatically detects an NTM by analysis of the Mirnov diagnostics, employs motional Stark effect (MSE) EFIT MHD equilibrium reconstruction to locate the rational q-surface where the NTM island can be found, then calculates the appropriate mirror position for alignment of the ECCD with the island using ray tracing. The control commands from PCS are sent to the electron cyclotron system to switch on and off or modulate the gyrotrons and to the steerable mirror system to move the steerable mirrors to the requested positions. Successful NTM suppression has been achieved in DIII-D using this control system to rapidly align the NTM island and the ECCD deposition location, and to actively maintain the alignment as plasma conditions change.

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.

The number of tokamak discharges required to carry out meaningful experiments can be significantly reduced by current profile control in the early startup phase. The tokamak is a plasma-confinement device, suitable for confining plasma at the requisite high temperature necessary for initiating fusion. It is currently the most promising device for realizing sustained fusion power generation at a commercial grade level, though it is still in the experimental stage. Presently, difficult to achieve plasma conditions, such as the shape of the plasma current profile, are achieved in a trial and error fashion, which can be a lengthy, wasteful process. In this work we make use of model-based control techniques such as optimal feedforward control via nonlinear programming and linearized feedback control to obtain a target current profile at a specified time in low-confinement-mode (L-mode) discharges. The effectiveness of the controller is demonstrated experimentally.

A new generation fault processor is in development which is intended to increase fault handling flexibility and reduce the number of incomplete DIII-D shots due to gyrotron faults. The processor, which is based upon a field programmable gate array device, will analyze signals for aberrant operation and ramp down high voltage to try to avoid hard faults. The processor will then attempt to ramp back up to an attainable operating point. The new generation fault processor will be developed during an expansion of the electron cyclotron heating (ECH) areas that will include the installation of a depressed collector gyrotron and associated equipment. Existing systems will also be upgraded. Testing of real-time control of the ECH launcher poloidal drives by the DIII-D plasma control system will be completed. The ECH control system software will be upgraded for increased scalability and to increase operator productivity. Resources permitting, all systems will receive an extra layer of interlocks for the filament and magnet power supplies, added shielding for the tank electronics, programmable filament boost shape for long pulses, and electronics upgrades for the installation of the advanced fault processor.

Major developments on the Mega Amp Spherical Tokamak (MAST) have enabled important advances in support of ITER and the physics basis of a spherical tokamak (ST) based component test facility (CTF), as well as providing new insight into underlying tokamak physics. For example, L–H transition studies benefit from high spatial and temporal resolution measurements of pedestal profile evolution (temperature, density and radial electric field) and in support of pedestal stability studies the edge current density profile has been inferred from motional Stark effect measurements. The influence of the q-profile and E × B flow shear on transport has been studied in MAST and equilibrium flow shear has been included in gyro-kinetic codes, improving comparisons with the experimental data. H-modes exhibit a weaker q and stronger collisionality dependence of heat diffusivity than implied by IPB98(y,2) scaling, which may have important implications for the design of an ST-based CTF. ELM mitigation, an important issue for ITER, has been demonstrated by applying resonant magnetic perturbations (RMPs) using both internal and external coils, but full stabilization of type-I ELMs has not been observed. Modelling shows the importance of including the plasma response to the RMP fields. MAST plasmas with q > 1 and weak central magnetic shear regularly exhibit a long-lived saturated ideal internal mode. Measured plasma braking in the presence of this mode compares well with neo-classical toroidal viscosity theory. In support of basic physics understanding, high resolution Thomson scattering measurements are providing new insight into sawtooth crash dynamics and neo-classical tearing mode critical island widths. Retarding field analyser measurements show elevated ion temperatures in the scrape-off layer of L-mode plasmas and, in the presence of type-I ELMs, ions with energy greater than 500 eV are detected 20 cm outside the separatrix. Disruption mitigation by massive gas injection has reduced divertor heat loads by up to 70%.

The DIII-D digital plasma control system (PCS) has been adapted and used at a number of tokamaks around the world over the past several years. Implementations of the DIII-D PCS at KSTAR and EAST in particular have made use of, and improved upon, the methodology for off-normal event/fault detection and response that was originally incorporated into the DIII-D PCS. This work discusses the lessons learned implementing and installing codes for detecting and responding to off-normal events in both plasma discharges and tokamak systems. In particular, we discuss the implementation details for codes at DIII-D, EAST, and KSTAR, and discuss how those experiences are being incorporated into a proposed finite state machine architecture in support of ITER research at DIII-D.

In order to realize the low latency and distortion-free signal transmission between the plasma control system (PCS) and servo systems, the digital output structure configured with reflective memory board (RFM) was adopted in EAST PCS. And the enhanced performances are reported. Another achievement made in the latest EAST PCS was the implementation of density control algorithm, which controlled the line average density in either voltage or width modulation mode. The new integrated algorithm improved the precision of density calculation and control performance greatly. The details and experiment results are presented in this paper.

This paper briefly outlines the current status of CODAC system of the EAST tokamak. Integrated central system has been working for synchronizing all the subsystems and responsible for the integral safety protection by interlock. Data acquisition system has been established for the discharge with duration up to 100 seconds. Various data visualization tools such as EASTVIEWER to view the flux surfaces, RTSCOPE to view plasma boundary and CCD image in real-time, WEBSCOPE to view the diagnostic data via a web browser and EASTVOD to view and search plasma discharge image at different time, haven developed for EAST operation. Plasma control system is inherited from DIII-D, and has been matured to a stage to feedback control the EAST plasma shape, density and current in real time.

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).

The Experimental Advanced Superconducting Tokamak (EAST) was the first shaped tokamak of mega-ampere scale to achieve plasma utilizing a fully superconducting poloidal field coil system, and it is ...

The EAST plasma control system (PCS) is in continuous development to satisfy the EAST experimental requirements. In order to realize low latency and distortion-free signal transmission between PCS and servo systems such as the poloidal field power supply, in-vessel coil power supply and real-time scope, reflective memory boards (RFM) were applied. The new hardware layout and enhanced performance are reported. Newly implemented PCS control algorithms for gas control and real-time data display are also presented.

Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in collaboration between experimental sites and laboratories worldwide. This scientific collaboration activity is strong at existing experimental sites, is a major element of machines just coming on line, and is also a thrust of experiments that will come on line in the next decade. Computer science research into enhancing the ability to scientifically participate in a fusion experiment remotely has been growing in size in an attempt to better address the needs of fusion scientists worldwide. The natural progression of this research is to examine how to move from remote scientific participation to remote hardware control. This paper examines the challenges associated with remote experimental device control and proposes a solution based on a semantic approach that defines a Gatekeeper software system that will be the only channel of interaction for incoming requests to the experimental site. The role of the Gatekeeper is to validate the identification and access privilege of the requestor and to ensure the validity of the proposed request. The Gatekeeper will be a modular system, transparent to end-users, and allow a high volume of activity.

A complete plasma control system (PCS) has been developed for KSTAR’s first plasma campaign as a collaborative project with the DIII-D team. The KSTAR real time plasma control system is based on a conceptual design by Jhang and Choi [Hogun Jhang, I.S. Choi, Fusion Engineering and Design 73 (2005) 35–49] and consists of a fast real-time computer/communication cluster and software derived from the GA-PCS [Penaflor, B.G., et.al., Fusion Engineering and Design, 83 (2) (2008) 176]. The system has been used for simulation testing, poloidal field (PF) coil power supply commissioning and first plasma control. The seven sets of up-down symmetric, superconducting PF coil/power supply systems have been successfully tested. Reflective memory (RFM) is utilized as the primary actuator/PCS real-time communication layer and PCS synchronization with KSTAR timing system and slower control devices is achieved through an EPICS implementation. Consistent feedback loop times of 100 microseconds has been achieved during PF coil power supply testing and first plasma commissioning. Here we present the “Day-One” plasma control system in its final form for the first plasma experimental campaign of KSTAR and describe how the system has been utilized during magnet commissioning and plasma startup experiments.

This paper describes a remote control room built at DIII-D to support remote participation activities of DIII-D research staff. In order to create a persistent, efficient, and reliable remote participation environment for DIII-D scientists, a remote control room has been built in a 640-ft 2 dedicated area. The purpose of this room is to experiment and define a remote control room framework that can facilitate the remote participation needs of current and future fusion experiments such as ITER. A variety of hardware equipment has been installed and several remote participation and collaboration technologies have been deployed. Objectivity and practical consideration has been the key while designing the room and deploying the technologies. Although, the DIII-D remote control room is still a work in progress and new software tools are being implemented, it has been already useful for a number of international remote participation activities. For example, it has been used for remote support of the EAST Tokamak in China during the start up operation and proven effective for other collaborative experiment activities. The description of the remote control room design is given along with technologies deployed for remote collaboration needs. We will also discuss our recent experiences involving the DIII-D remote control room as well as future plans for improvements.

Plasma control design approaches and solutions developed at DIII-D to address its control-intensive advanced tokamak (AT) mission are applicable to many problems facing ITER and other next-generation devices. A systematic approach to algorithm design, termed “integrated plasma control,” enables new tokamak controllers to be applied operationally with minimal machine time required for tuning. Such high confidence plasma control algorithms are designed using relatively simple (“control-level”) models validated against experimental response data and are verified in simulation prior to operational use. A key element of DIII-D integrated plasma control, also required in the ITER baseline control approach, is the ability to verify both controller performance and implementation by running simulations that connect directly to the actual plasma control system (PCS) that is used to operate the tokamak itself. The DIII-D PCS comprises a powerful and flexible C-based realtime code and programming infrastructure, as well as an arbitrarily scalable hardware and realtime network architecture. This software infrastructure provides a general platform for implementation and verification of realtime algorithms with arbitrary complexity, limited only by speed of execution requirements. We present a complete suite of tools (known collectively as TokSys) supporting the integrated plasma control design process, along with recent examples of control algorithms designed for the DIII-D PCS. The use of validated physics-based models and a systematic model-based design and verification process enables these control solutions to be directly applied to ITER and other next-generation tokamaks.

Experimental advanced superconducting tokamak (EAST), the first tokomak with ITER-like fully superconductive poloidal and toroidal coils, successfully completed its engineering commission in March 2006; had its first plasma in September 2006 [B.N. Wan, et al., The first plasma of EAST, in: 21st IAEA Fusion Energy Conference, Chengdu, China, 16–21 October, 2007. [1] ] and further generated diverted plasma in January 2007. The plasma control system was adapted from DIII-D plasma control system architecture and jointly developed by the DIII-D and EAST plasma control team. After briefly outlining the hardware and software architecture of the EAST plasma control system, this paper summarizes the plasma control results in the first plasma and first diverted plasma campaigns.

The DIII-D advanced tokamak physics program requires extremely high performance from the DIII-D plasma control system, including simultaneous accurate regulation of plasma shape, stored energy, density and divertor characteristics, as well as coordinated suppression of magnetohydrodynamic instabilities. To satisfy these demanding control requirements, we apply the integrated plasma control method, consisting of construction of physics-based plasma and system response models, validation of models against operating experiments, design of integrated controllers that operate in concert with one another, simulation of control action against off-line and actual machine control platforms, and optimization through iteration of the design-test loop. The present work describes progress in development of physics models and development and experimental application of new model-based plasma controllers on DIII-D. We also describe the development of the control software, hardware, and model-based control algorithms for the superconducting EAST and KSTAR tokamaks.

The requirements of the DIII-D physics program have led to the development of many operational control results with direct relevance to ITER. These include new algorithms for robust and sustained stabilization of neoclassical tearing modes with electron cyclotron current drive, model-based controllers for stabilization of the resistive wall mode in the presence of ELMs, coupled linear–nonlinear algorithms to provide good dynamic axisymmetric control while avoiding coil current limits, and adaptation of the DIII-D plasma control system (PCS) to operate next-generation superconducting tokamaks. Development of integrated plasma control (IPC), a systematic approach to modelbased design and controller verification, has enabled successful experimental application of high reliability control algorithms requiring a minimum of machine operations time for testing and tuning. The DIII-D PCS hardware and software and its versions adapted for other devices can be connected to IPC simulations to confirm control function prior to experimental use. This capability has been important in control system implementation for tokamaks under construction and is expected to be critical for ITER.

A complete software suite for integrated tokamak plasma control has been developed within the DIII-D program. The suite consists of software for real-time control of all aspects of the plasma, modeling, simulation and design tools for analysis and development of controllers, a flexible and modular architecture for implementation and testing of algorithms and many fully validated models. Many elements of the system have been applied to and implemented on NSTX and MAST. The DIII-D realtime plasma control system together with the integrated modeling and simulation suite have been selected for operational use by both the KSTAR and EAST tokamaks, and are also being used at General Atomics to investigate control issues for ITER.

The DIII-D mission to explore the advanced tokamak (AT) regime places significant demands on the DIII-D plasma control system (PCS) [D.A. Humphreys, et al., Fusion Eng. Des. 66–68 (2003) 633], including simultaneous and accurate regulation of plasma shape, stored energy, and density, as well as coordinated suppression of magnetohydrodynamic (MHD) instabilities. The present work describes selected new solutions to key control problems in DIII-D AT operation. These include novel nonrigid, resistive linear plasma response models, nonlinear algorithms for avoidance of current limiting, and improved neoclassical tearing mode control algorithms.

This paper describes the improvements being made in the capabilities of the DIII–D plasma control system (PCS) towards achieving optimization of current and pressure profiles in advanced tokamak discharges. Key improvements have been increased processing power and the ability to include profile diagnostic data. The recently completed upgrade of the PCS to Linux based Intel computers connected with 2 Gigabit/s Myrinet networking technology has been successful in achieving the goals of increasing the overall performance and flexibility of the system. The new Intel computing system has increased processing power by a factor 30 over the older i860 based systems. The Myrinet fiber based network has allowed the inclusion of data in real-time from DIII–D diagnostics, situated in remote locations within the DIII–D research facility. The PCS now collects 32 channels of motional stark effect (MSE) data and uses these data for real-time computation of the safety factor (q) profile. Electron temperature and density profile data from the Thomson scattering diagnostic and electron temperature profile data from the electron cyclotron emission (ECE) diagnostic are presently being added. Addition of an ion temperature and toroidal rotation profile data from the charge exchange recombination diagnostic is planned. Feedback control by the PCS of the electron temperature at two points has been demonstrated using either ECH or neutral beam power. This has been used to modify current profile evolution during plasma current ramp up. Specifics of the latest improvements to the DIII–D PCS are detailed here.

An extensive set of software tools for integrated plasma control, developed and validated on the DIII-D tokamak, has been applied to several next-generation fusion device designs including KSTAR, EAST, and ITER. These devices will require elements of integrated plasma control in order to achieve high reliability advanced tokamak or burning plasma operation. Plasma Control Systems (PCS) based on the DIII-D PCS [1] have been designed for each of these devices. The integrated plasma control approach uses validated physics models to design controllers for plasma shape and both axisymmetric and nonaxisymmetric MHD instabilities [2] and confirms control performance by operating actual machine control hardware and software against detailed tokamak system simulations. The physics-based models include conductors, diagnostics, power supplies, and both linear and nonlinear plasma models. These models can be implemented in the detailed control simulations to verify event handling and demonstrate functioning of control action under realistic hardware (CPU and network) conditions. Results of simulations are shown, illustrating control performance characteristics produced by each device design, engineering choices, and control system algorithms and hardware. Such simulations allow confirmation of performance prior to actual implementation on an operating device.

A271 ADVANCED TOKAMAK OPERATION USING THE DIII-D PLASMA CONTROL SYSTEM. The principal focus of experimental operations in the DIII-D tokamak is the advanced tokamak (AT) regime to achieve, which requires highly integrated and flexible plasma control. In a high performance advanced tokamak, accurate regulation of the plasma boundary, internal profiles, pumping, fueling, and heating must be well coordinated with MHD control action to stabilize such instabilities as tearing modes and resistive wall modes. Sophisticated monitors of the operational regime must provide detection of off-normal conditions and trigger appropriate safety responses with acceptable levels of reliability. Many of these capabilities are presently implemented in the DIII-D plasma control system (PCS), and are now in frequent or routine operational use. The present work describes recent development, implementation, and operational experience with AT regime control elements for equilibrium control, MHD suppression, and off-normal event detection and response.

This paper describes an upgrade for the real-time computing system responsible for monitoring and controlling plasma, discharges in the DIII–D tokamak (J.L. Luxon, L.G. Davis, Fusion Technol. 8 (1985) 441) at General Atomics (GA). The current system employs six CSPI i860 VME format processors working in parallel to acquire data in real-time and perform feed back control of plasma shape and position parameters. Work has commenced on integration of a new computing system based on commonly available PCI bus based processors that communicate over the 2 Gbit/s Myrinet (Myricom, Inc., CA, USA) network. The new system will greatly improve the processing power available to the algorithms required for computing plasma equilibrium reconstructions in real-time. A factor of 20 anticipated performance increase will allow for improved accuracy and frequency response for plasma shape estimation and control. The migration from VME to PCI Myrinet computer clustering will improve the data acquisition capabilities by opening up access to additional DIII–D temperature and density diagnostics. The upgrade will increase the inter-processor communication speed and provide the flexibility to integrate additional processors to match the cost and computing needs of the tokamak research program.

A model-based multivariable controller for plasma shape control has been successfully tested on the DIII–D tokamak. Good steady-state control of plasma boundary shape and X-point position was demonstrated in lower single-null ohmic plasmas. Quality of control for rapid plasma shape variation was mixed, but was robustly stable for all degrees of freedom explored. The control design was based on a linear plasma response model derived from fundamental physics assumptions, which was extensively validated against DIII–D experimental data. A linear controller produced with robust control design methods was tested and improved using results of closed loop simulations prior to experimental tests. A modification of the linear controller which addressed one of several practical DIII–D nonlinear constraints was tested during the experimental discharges.

This paper describes the current status of real-time digital plasma control for the DIII-D tokamak. The digital plasma control system (PCS) has been in place at DIII-D since the early 1990s and continues to expand and improve in it's capabilities to monitor and control plasma parameters for DIII-D fusion science experiments. The PCS monitors over 200 tokamak parameters from the DIII-D experiment using a real-time data acquisition system that acquires a new set of samples once every 60 /spl mu/s. This information is then used in a number of feedback control algorithms to compute and control a variety of parameters including those affecting plasma shape and position. A number of recent improvements to the control capabilities of the PCS along with general system enhancements have contributed to the advancement and understanding of fusion energy science at DIII-D. Among the more significant improvements to the PCS control capabilities has been the successful incorporation of advanced techniques for reconstruction of plasma equilibrium parameters in realtime. An isoflux control method using these improved plasma parameter calculations has greatly enhanced the ability of achieving desired shaped plasma on the DIII-D tokamak and even made it possible to achieve new and different ones. A number of system related improvements has improved the usability and flexibility of the DIII-D PCS. These include more graphical user interfaces to assist in entering and viewing the large and ever growing number of parameters controlled by the PCS, increased interaction and accessibility from other DIII-D applications, and upgrades to the computer hardware and vended software. Future plans for the system include possible upgrades of the real-time computers, further links to other DIII-D diagnostic measurements such as real-time Thomson scattering analysis, and joint collaborations with other tokamak experiments including the NSTX at Princeton.

OAK A271 NEXT-GENERATION PLASMA CONTROL IN THE DIII-D TOKAMAK. The advanced tokamak (AT) operating mode which is the principal focus of the DIII-D tokamak requires highly integrated and complex plasma control. Simultaneous high performance regulation of the plasma boundary and internal profiles requires multivariable control techniques to account for the highly coupled influences of equilibrium shape, profile, and stability control. This paper describes progress towards the DIII-D At mission goal through both significantly improved real-time computational hardware and control algorithm capability.

During the tokamak discharge, especially the ramp-up phase, the plasma state equilibrium continually evolves. As a consequence, the plasma response model should evolve as well. We first identified a linear plasma response model of the rotational transform ι profile and βN around a desired equilibrium. Then, an uncertainty is introduced to the identified model to partially account for the dynamic character of the plasma state equilibrium evolution. A robust controller is designed to stabilize this family of plasma models, which are reformulated into a nominal model with uncertainty. A singular value decomposition (SVD) of the nominal identified model is carried out to decouple and identify the most relevant control channels in steady-state. The DK-iteration method, combining H∞ synthesis and μ analysis, is applied to synthesize a closed-loop controller that minimizes the tracking error and input effort. The feedback controller is then augmented with an anti-windup compensator, which keeps the given profile controller well-behaved in the presence of magnitude constraints in the actuators and leaves the nominal closed-loop unmodified when no saturation is present. PTRANSP simulations and experimental results in DIII-D illustrate the performance of the model-based controller.

The paper proposes a model-based control approach for the coupled evolution of the poloidal magnetic flux profile and the normalized pressure ratio βN. The model is determined by a system identification method which is shown to sufficiently reproduce the plasma response to variations in particular actuators. Data for model identification is collected during the plasma current flattop in a large βN, high-confinement scenario (H-mode) with the actuators modulated in open loop. Using this data, a linear state-space plasma response model for the poloidal magnetic flux profile and βN dynamics around a plasma equilibrium state is identified. An optimal state feedback controller with integral action is designed for the purpose of simultaneous control of the poloidal flux profile and βN. Experimental results showing the performance of the proposed controller implemented in the DIII-D tokamak are presented.

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.

System identification techniques have been successfully used to obtain linear dynamic plasma response models around a particular equilibrium in different tokamaks. This paper identifies a two-time-scale dynamic model of the rotational transform ι profile and β N in response to the electric field due to induction as well as to heating and current drive (H&CD) systems based on experimental data from DIII-D. The control goal is to regulate the plasma ι profile and β N around a particular target value. A singular value decomposition (SVD) of the plasma model at steady state is carried out to decouple the system and identify the most relevant control channels. A mixed sensitivity H ∞ control design problem is solved to determine a stabilizing feedback controller that minimizes the reference tracking error and rejects external disturbances with minimal control energy. The feedback controller is augmented with an anti-windup compensator, which keeps the given controller well-behaved in the presence of magnitude constraints in the actuators and leaves the nominal closed-loop unmodified when no saturation is present. Experimental results illustrate the performance of the proposed controller, which is one of the first profile controllers integrating magnetic and kinetic variables ever implemented in DIII-D.

In tokamak fusion plasmas, controlling the spatial distribution profile of the toroidal current is key to achieving advanced scenarios characterized by confinement improvement and possible steady-state operation. The dynamics of the current profile are nonlinear and coupled with other plasma parameters, motivating the use of model-based control strategies. In this work, we use a control-oriented model of the current profile evolution in DIII-D to design a feedback controller for regulating the profile around a desired trajectory. Without feedback, the response of the current profile to disturbances, model uncertainty, and perturbed initial conditions can be undesirable. To improve tracking performance of the system, a nonlinear input transformation is combined with a linear-quadratic-integral (LQI) optimal controller designed to minimize a weighted combination of the tracking error and controller effort. The resulting control law utilizes the total plasma current, total external heating power, and line averaged plasma density as actuators. We present simulation and experimental results showing successful rejection of perturbed initial conditions and input disturbances.

The tokamak is a high order, distributed parameter, nonlinear system with a large number of instabilities. Therefore, accurate theoretical plasma models are difficult to develop. However, linear plasma response models around a particular equilibrium can be developed by using data-driven modeling techniques. This paper introduces a linear model of the rotational transform ι profile evolution based on experimental data from the DIII-D tokamak. The model represents the response of the ι profile to the electric field due to induction as well as to heating and current drive (H&CD) systems. The control goal is to use both induction and H&CD systems to regulate the plasma ι profile around a particular target profile. A singular value decomposition (SVD) of the plasma model at steady state is carried out to decouple the system and identify the most relevant control channels. A mixed sensitivity H ∞ control design problem is formulated to synthesize a stabilizing feedback controller without input constraint that minimizes the reference tracking error and rejects external disturbances with minimal control energy. The feedback controller is then augmented with an anti-windup compensator, which keeps the given profile controller well-behaved in the presence of magnitude constraints in the actuators and leaves the nominal closed-loop unmodified when no saturation is present. Finally, computer simulations and experimental results illustrate the performance of the model-based profile controller.

Setting up a suitable toroidal current profile in a fusion tokamak reactor is vital to the eventual realization of a commercial nuclear fusion power plant. Creating the desired current profile during the ramp-up and early flat-top phases of the plasma discharge and then actively maintaining this target profile for the remainder of the discharge is the goal at the DIII-D tokamak. The evolution of the toroidal current profile in toka-maks is related to the evolution of the poloidal magnetic flux profile, which is modeled by the magnetic diffusion equation. A simplified first-principles-driven, nonlinear, dynamic, control-oriented, partial differential equation model of the poloidal flux profile evolution is obtained by combining the magnetic diffusion equation with empirical correlations obtained from experimental data at DIII-D and is used to synthesize a robust H ∞ feedback controller to track a desired reference trajectory of the poloidal magnetic flux gradient profile. We employ a singular value decomposition of the static gain matrix of the plant model to identify the most relevant channels which we control with the feedback controller. A framework for real-time feedforward + feedback control was implemented in the DIII-D Plasma Control System and experimental results in the DIII-D tokamak are presented to illustrate the capabilities of the feedback controller. These experiments mark the first time ever a first-principles-driven model-based magnetic profile controller was successfully implemented and tested in a tokamak device.

Control of the spatial profile of plasma current in tokamak plasmas has been demonstrated to be a key condition for achieving advanced scenarios with improved confinement and possible steady-state operation. The dynamics of the current profile are nonlinear and coupled with several other plasma parameters, motivating the design of model-based controllers that can account for these complexities. In this work, we consider a control-oriented model of the current profile evolution in DIII-D and the problem of regulating the current profile around a desired feed-forward trajectory. In open-loop, the response of the system to disturbances and perturbed initial conditions may be undesirable. To improve the performance of the system, the PDE model is discretized in space using a finite difference method and a backstepping design is applied to obtain a discrete transformation from the original system into an asymptotically stable target system with desirable properties. Through a nonlinear transformation, the resulting boundary control law utilizes the total plasma current, total power, and line averaged density as actuators. A Simserver simulation study is done to test the controller's performance and its implementation in the DIII-D plasma control system. Finally, experimental results showing the ability of the controller to reject input disturbances and perturbations in initial conditions are presented.

First-principles predictive models based on flux-averaged transport equations often yield complex expressions not suitable for real-time control implementations. It is however always possible to reduce these models to forms suitable for control design while preserving the dominant physics of the system. If further model simplification is desired at the expense of less model accuracy and controller capability, data-driven modeling emerges as an alternative to first-principles modeling. System identification techniques have the potential of producing low-complexity, linear models that can capture the system dynamics around an equilibrium point. This paper focuses on the control of the poloidal magnetic flux profile evolution in response to the heating and current drive (H&CD) systems and the total plasma current. Open-loop data for model identification is collected during the plasma current flattop in a high-confinement scenario (H-mode). Using this data a linear state-space plasma response model for the poloidal magnetic flux profile dynamics around a reference profile is identified. The control goal is to use the H&CD systems and the plasma current to regulate the magnetic profile around a desired target profile in the presence of disturbances. The target profile is defined close enough to the reference profile used for system identification in order to stay within the range of validity of the identified model. An optimal state feedback controller with integral action is designed for this purpose. Experimental results showing the performance of the proposed controller implemented in the DIII-D tokamak are presented.