The EAST plasma control system (PCS) undergoes continuous development and upgrade to achieve major goals of EAST steady-state advanced operation. According to the development of computer multi-core processor technology, the four nodes PCS cluster was simplified to one host and one real time computing node structure. Besides, the cluster internal data exchange network Myrinet was replaced by high speed 10 Giga-bit Ethernet for inter machine and shared memory technology for multi-core communication. Such new upgraded system is called standalone PCS which has less hardware but be more efficient and easy to backup. In the new upgraded system, a real time archiving mode using data segment technology of MDSplus was realized which provided the possibility to save all data in segments without increasing the computer memory or reducing the saving frequency in steady–state operation. To reduce the heat flux and surface temperature on the divertor targets, plasma radiation control using divertor inert gas puffing and mid-plane supersonic molecular beam injection (SMBI) was integrated in PCS. In this paper, the control design and experiment results are discussed.

To accomplish experimental advanced superconducting tokamak (EAST) physical targets, the plasma control system (PCS), adapted from DIII-D PCS and deployed on EAST in 2005, keeps in continuous development. Some new control abilities for steady-state operation have been achieved. To avoid the integrator linear drift in long pulse discharge, the linear rate for each integrator channel is precalculated and used to decrease the effect of the acquired raw data during shots. Another strategy applied for long pulse operation is real-time data archiving using data segment technology of MDSplus, which provides the possibility to save all data in segments without increasing the computer memory or reducing the saving frequency. For plasma high-performance and noninductive operations, the plasma beta and loop voltage control was implemented in PCS. Using low hybrid wave, the control algorithm was successfully verified in 2016 EAST campaign. Besides, another two control algorithms are integrated to reduce the divertor heat flux. One is radiation power control, which is successfully feedback controlled by using divertor inert gas puffing and mid-plane supersonic molecular beam injection. The other is quasi-snowflake shape control using PEFIT/ISOFLUX, which shows significant heat load reduction to divertor target according to the experimental result. The present EAST PCS has become a huge system capable of long pulse, high-performance advanced plasma control operation, which is ready to demonstrate ITER-like control contents.

Results on the integration and the operation of the KSTAR plasma control system (PCS) upgrade are given. Realtime hardware, new realtime-capable operating system, and a brand-new data acquisition are assembled in order to extend the performance and compatibility with modern computer systems. The first full commissioning and eventual routine use performed in 2016 so that the system can now acquire more than 400 channels for more than 100 seconds of data with 5 kHz sampling. The performance test results are summarized, featuring In/Out streaming echo tests and the synchronization verifications. Examples of general performance improvements are demonstrated, and additional features added to the software are also described.

Disruption avoidance is one of the most critical issues in KSTAR due to its high current, temperature and long operation characteristics of the superconducting tokamak. To minimize possible damages to the machine, we imposed a real time handling algorithm of off-normal events, named "forced landing" to ramp down the plasma with well-controlled plasma current and position. Two different plasma control schemes are implemented for different purposes and situations. Both schemes have been successfully demonstrated with careful classifications of actions on the machine in real experiments and they show routine performance with effective reduction of impacts on the machine. Such "machine-protection" schemes enable more aggressive operations such as mega-ampere or high beta experiment, resulting in an expansion of the KSTAR operation regime due to reduced less concerns regarding mechanical safety issues.

The upgrade to the National Spherical Torus eXperiment (NSTX-U) included two main improvements: a larger center-stack, enabling higher toroidal field and longer pulse duration, and the addition of three new tangentially aimed neutral beam sources, which increase available heating and current drive, and allow for flexibility in shaping power, torque, current, and particle deposition profiles. To best use these new capabilities and meet the high-performance operational goals of NSTX-U, major upgrades to the NSTX-U control system (NCS) hardware and software have been made. Several control algorithms, including those used for real-time equilibrium reconstruction and shape control, have been upgraded to improve and extend plasma control capabilities. As part of the commissioning phase of first plasma operations, the shape control system was tuned to control the boundary in both inner-wall limited and diverted discharges. It has been used to accurately track the requested evolution of the boundary (including the size of the inner gap between the plasma and central solenoid, which is a challenge for the ST configuration), X-point locations, and strike point locations, enabling repeatable discharge evolutions for scenario development and diagnostic commissioning.

Recent experiments on DIII-D demonstrate the potential of physics-model-based q-profile control to improve reproducibility of plasma discharges. A combined feedforward + feedback control scheme is employed to optimize the current ramp-up phase by consistently achieving target q profiles (Target 1: q(min) = 1.3, q(95) = 4.4; Target 2: q(min) = 1.65, q(95) = 5.0; Target 3: qmin = 2.1, q(95) = 6.2) at prescribed times during the plasma formation phase (Target 1: t = 1.5 s; Target 2: t = 1.3 s; Target 3: t = 1.0 s). At the core of the control scheme is a nonlinear, first-principles- driven, physics-based, control-oriented model of the plasma dynamics valid for low confinement (L-mode) scenarios. To prevent undesired L-H transitions, a constraint on the maximum allowable total auxiliary power is imposed in addition to the maximum powers for the individual heating and current-drive sources. Experimental results are presented to demonstrate the effectiveness of the combined feedforward + feedback control scheme to consistently achieve the desired target profiles at the predefined times. These results also show how the addition of feedback control significantly improves upon the feedforward-only control solution by reducing the matching error and also how the feedback controller is able to reduce the matching error as the constraint on the maximum allowable total auxiliary power is relaxed while keeping the plasma in L-mode.

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.

On the DIII-D National Fusion Facility tokamak plasma diagnostics continue to improve and experiments increase in complexity. Hence the utility of dynamic control of the beam energy (and therefore also the injected torque, ion heating fraction, etc.) has become apparent. Here we report on upgrades that have been incorporated into the DIII-D Plasma Control System (PCS) and Neutral Beam Injection (NBI) systems in order to allow the beam acceleration voltage (Vaccel) to be varied continuously in a <= 20 kV range during a shot for the first time, generating new capabilities such as smooth plasma transitions and controllable interactions with Alfven waves.

The upgrade to the National Spherical Torus eXperiment (NSTX-U) [1, 2] included a larger center-stack, enabling higher toroidal field and longer pulse duration, and three new tangentially aimed neutral beam sources, which increase heating and current drive, and allow for flexibility in shaping deposition profiles. To meet the high-performance goals of NSTX-U, major upgrades to the Plasma Control System (PCS) hardware [3] and software have been made. Several control algorithms, including those used for vertical control, real-time equilibrium reconstruction, and shape control, have been upgraded to improve and extend control capabilities. The shape controller has been tuned to control inner-wall limited and diverted discharges and has been used with the vertical position controller to produce repeatable discharge evolutions, contributing to achieving 1MA, 0.65T scenarios on NSTX-U with 2s pulse length.

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 similar to 48s 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. (C) 2016 Elsevier B.V. All rights reserved.

DIII-D experimental results are reported to demonstrate the potential of physics-model-based safety factor profile control for robust and reproducible sustainment of advanced scenarios. In the absence of feedback control, variability in wall conditions and plasma impurities, as well as drifts due to external disturbances, can limit the reproducibility of discharges with simple preprogrammed scenario trajectories. The control architecture utilized is a feedforward + feedback scheme where the feedforward commands are computed off-line and the feedback commands are computed on-line. In this work, a first-principles-driven (FPD), physics-based model of the q profile and normalized beta (beta(N)) dynamics is first embedded into a numerical optimization algorithm to design feedforward actuator trajectories that steer the plasma through the tokamak operating space to reach a desired stationary target state that is characterized by the achieved q profile and beta(N). Good agreement between experimental results and simulations demonstrates the accuracy of the models employed for physics-model-based control design. Second, a feedback algorithm for q profile control is designed following an FPD approach, and the ability of the controller to achieve and maintain a target q profile evolution is tested in DIII-D high confinement (H-mode) experiments. The controller is shown to be able to effectively control the q profile when beta(N) is relatively close to the target, indicating the need for integrated q profile and beta(N) control to further enhance the ability to achieve robust scenario execution. The ability of an integrated q profile + beta(N) feedback controller to track a desired target is demonstrated through simulation.

The plasma control system (PCS) plays a vital role at EAST for fusion science experiments. Its software application consists of two main parts: an IDL graphical user interface for setting a large number of plasma parameters to specify each discharge, several programs for performing the real-time feedback control and managing the whole control system. The PCS user interface can be used from any X11 Windows client with privileged access to the PCS computer system. However, remote access to the PCS system via the IDL user interface becomes an extreme inconvenience due to the high network latency to draw or operate the interfaces. In order to realize lower latency for remote access to the PCS system, a web-based system has been developed for EAST recently. The setup data are retrieved from the PCS system and client-side JavaScript draws the interfaces into the user's browser. The user settings are also sent back to the PCS system for controlling discharges. These technologies allow the web-based user interface to be viewed by authorized users with a web browser and have it communicate with PCS server processes directly. It works together with the IDL interface and provides a new way to aid remote participation.

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.

Control of the current profile in tokamak plasmas has been shown to play an important role in achieving advanced scenarios that could enable steady-state operation. The nonlinearity and spatially distributed nature of the current profile dynamics motivate the use of model-based control designs. In this work, we consider a control-oriented model of the current profile evolution in DIII-D high-confinement (H-mode) discharges, and the problem of regulating the current profile around a desired trajectory. The PDE model is discretized in space with a finite difference method and a backstepping design is applied to obtain a transformation from the original system into a particular target system with desirable properties. The resulting boundary condition control law is complemented with control laws for the available distributed actuators. The combined control strategy uses nonlinear combinations of the total plasma current, total power, and line averaged density as actuators. Simulation and experimental results show the ability of the controller to track desired targets and to reject input disturbances.

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 (10–16s) with the number of required commissioning shots largely reduced.

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 (PF) coils, and power supplies, and is used to verify the control algorithm and optimize control parameters and PF 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.