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Plasma vertical displacement control is essential for the stable operation of tokamak devices. The traditional plasma vertical displacement calculation method is not suitable for balancing speed and accuracy simultaneously, which is necessary for real-time feedback control. In this study, neural networks are used to rapidly detect vertical displacement recognition. Based on a fully connected neural network, the vertical displacement calculation model is trained and tested using magnetic data of approximately 2000 shots. To compare the effects of different inputs on vertical displacement calculation, different magnetic measurement diagnostic signals are used to train and test the model. Compared with a full magnetic measurement dataset, 39 magnetic measurement signals (38 magnetic probes and plasma current) show better accuracy with mean square error

Experiments at DIII-D and EAST have demonstrated simultaneous high confinement, divertor detachment, and active control of detachment level, all of which are required for ITER. Comparing detachment control via T e and J s a t , it appears that T e is the most straightforward sensor to use for accessing detachment onset, while J s a t offers more precise control of degree of detachment. Based on these results, control using nitrogen seeding has so far shown the best ability to follow a target value with the low disruptivity and little to no degradation of performance when an Internal Transport Barrier (ITB) is present, but not all facilities allow its use. Neon seeding also can be paired with feedback control with low impact on core performance as long as there is no disruption, however shots with neon seeding commonly disrupted during these experiments. Argon is effective in EAST, but tends to degrade performance (by ≈ 10 % β p ) when detachment is achieved. With ideal conditions and strike point position control, data from a single Langmuir probe are an acceptable input to the control algorithm, but this simple system is easily defeated by strike point displacement comparable to the T e or J s a t scale lengths. The presence of an ITB seems to be critical to retaining core performance in detachment in these parameter ranges, as the pedestal pressure tends to decrease as a result of impurity seeding.

For the problem of excessively high divertor heat flux, active impurity seeding is an effective method to radiate the plasma energy reaching the divertor and thus achieve the divertor detachment. Neon is a very effective radiation impurity on many current tokamaks, which is also a candidate species to be applied on ITER. In the EAST 2019 experimental campaign, a series of experiments were performed by seeding a mixture of neon and deuterium (Ne-D2) for detachment and core-edge-divertor integration in H-mode plasmas. The divertor partial detachment with high-confinement core plasma has been achieved by using Ne-D2 seeding in EAST with ITER-like tungsten divertor. Both the plasma stored energy and H 98,y2 > 1.1 are maintained, with the divertor electron temperature, heat flux and the surface temperature near the strike point being all significantly reduced. The differences between Ne-D2 seeding at the scrape-off layer (SOL) upstream and downstream have been experimentally investigated in detail. It is found that impurity seeding at SOL downstream is more beneficial to reducing the divertor electron temperature and peak heat flux. By comparison with experiments using divertor D2 fueling, it is further demonstrated that gas seeding in the SOL downstream will enrich more particles near the strike point, while the seeding in the SOL upstream will influence the entire outer target more evenly. Furthermore, in most of the experiments, gas seeding does not cause obvious toroidal asymmetry in the divertor plasma. However, when D2 is injected in an amount similar to that used to build the plasma, it causes the particle flux near the gas-puff to increase locally, i.e., much more than that at the toroidal location far from the gas-puff location. It is a competition between particle source and transport. When the particle source is stronger, it will naturally increase the local particles. In addition, dedicated experiments with different poloidal distances between impurity seeding and strike point on the radiation ability were carried out. Both experimental results and SOLPS simulation show that the seeding close to the strike point is more conducive to neon ionization and energy radiation.

Simultaneous control of transient heat load induced by large-amplitude edge-localized modes (ELMs) and steady-state heat load on divertor targets under metal wall environment is crucial for steady-state operation of future tokamak fusion reactors, such as ITER and the China Fusion Engineering Test Reactor (CFETR). In the recent experiments, sustained partial energy detachment without confinement degradation has been achieved in the Experimental Advanced Superconducting Tokamak (EAST) in high-performance grassy-ELM H-mode with q95 ~ 5.9 by a newly developed detachment feedback control scheme, in which we first used electron temperature (Tet) measured by divertor Langmuir probes to identify the onset of energy detachment, and then the system switched to the feedback control of total radiation power measured by absolute extreme ultraviolet (AXUV) system. Tet around the upper outer strike point was successfully maintained less than 8 eV with seeding of 80% Ne and 20% D2 mixture from upper outer divertor, and the total radiation power was maintained ~1.4 MW, around 52% of injected power. There was no significant decrease of the plasma stored energy and H98,y2 factor (~1) over the entire detachment feedback control process. These experiment results demonstrate good compatibility of the high-performance grassy-ELM regime with radiative divertor. In order to confirm the compatibility in a wider range, stable partial energy detachment in grassy-ELM H-mode with relatively lower q95 (~5.4) was also achieved in EAST through the newly developed integrated-feedback-control technique. The new detachment feedback control without confinement degradation in grassy-ELM H-mode provides a candidate mode for EAST long-pulse operation in the future with well control of ELM-induced transient and steady heat fluxes on the divertor target.

The compatibility of efficient divertor detachment with high-performance core plasma is vital to the development of magnetically controlled fusion energy. The joint research on the EAST and DIII-D tokamaks demonstrates successful integration of divertor detachment with excellent core plasma confinement quality, a milestone towards solving the critical plasma–wall-interaction (PWI) issue and core-edge integration for ITER and future reactors. In EAST, actively controlled partial detachment with T et,div ∼ 5 eV around the strike point and H 98 > 1 in different H-mode scenarios including the high β P H-mode scenario have been achieved with ITER-like tungsten divertor, by optimizing the detachment access condition and performing detailed experiments for core-edge integration. For active long-pulse detachment feedback control, a 30 s H-mode operation with detachment-control duration being 25 s has been successfully achieved in EAST. DIII-D has achieved actively controlled fully detached divertor with low plasma electron temperature (T et,div ⩽ 5 eV across the entire divertor target) and low particle flux (degree of detachment, DoD > 3), simultaneously with very high core performance (β N ∼3, β P > 2 and H 98 ∼ 1.5) in the high βP scenario being developed for ITER and future reactors. The high-β P high confinement scenario is characterized by an internal transport barrier (ITB) at large radius and a weak edge transport barrier (ETB, or pedestal), which are synergistically self-organized. Both the high-β P scenario and impurity seeding facilitate divertor detachment. The detachment access leads to the reduction of ETB, which facilitates the development of an even stronger ITB at large radius in the high β P scenario. Thus, this strong large radius ITB enables the core confinement improvement during detachment. These significant joint DIII-D and EAST advances on the compatibility of high confinement core and detached divertor show a great potential for achieving a high-performance core plasma suitable for long-pulse operation of fusion reactors with controllable steady-state PWIs.

The efficient and safe operation of large fusion devices relies on plasma configuration inside the vacuum chamber. Due to EAST PF coils distribution, it is difficult to physical decouple plasma control parameters and poloidal field (PF) coils current. In this paper, we present a MIMO decoupling controller aimed to overcome the intrinsic limitations of simpler SISO controller, this new controller has been validated by means of the Tokamak Simulation Code (TSC) and implemented on the EAST tokamak. Preliminary results show the potential of this approach for the EAST plasma shape control systems. © 2018 Elsevier B.V.

Robust real-time control algorithms for tracking plasma kinetic parameters in advanced tokamak scenarios are developed based on linear state-space dynamic models. The real-time control algorithms under study comprise the H ∞ robust control, the linear quadratic integral (LQI) control and the internal model control (IMC). The plasma models used in this work are restricted to linear state-space models identified from dedicated simulation/experimental data, though the proposed control algorithms can conveniently be extrapolated to broadly incorporate linear models obtained from first-principles plasma theory. The control objective is to track plasma kinetic parameters of interest to desired operating points in advanced tokamak scenarios by actuating additional heating & current drive (H&CD) systems in real-time. Plasma kinetic parameters involve the poloidal pressure parameter β p , the internal inductance li, the average toroidal rotation angular speed Ω φ and the electron temperature on axis T e,0 while the actuators are the ion cyclotron resonance heating (ICRH) and lower hybrid current drive (LHCD) systems. In order to achieve enhanced control performance, two control layers are designed. The outer layer, i.e. an internal model-based proportional-integral (PI) actuator controller, operating on a fast timescale ( the energy confinement time τ E ) aiming at tracking the commands requested by the inner kinetic controllers, while the inner layer, i.e. a kinetic controller chosen from various alternatives, running on a slow timescale (∼ τ E ) is dedicated to tailoring plasma kinetic parameters. Simulation results for the EAST tokamak are provided and compared to show the capabilities of each control approach. Dedicated kinetic control experiments conducted in an H-mode scenario on EAST are reported as well. The advantages and limits of these control algorithms are discussed and summarised.

A real-time disruption predictor using random forest was developed for high-density disruptions and used in the plasma control system (PCS) of the EAST tokamak for the first time. The disruption predictor via random forest (DPRF) ran in piggyback mode and was actively exploited in dedicated experiments during the 2019–2020 experimental campaign to test its real-time predictive capabilities in oncoming high-density disruptions. During dedicated experiments, the mitigation system was triggered by a preset alarm provided by DPRF and neon gas was injected into the plasma to successfully mitigate disruption damage. DPRF’s average computing time of ∼250 μs is also an extremely relevant result, considering that the algorithm provides not only the probability of an impending disruption, i.e. the disruptivity, but also the so-called feature contributions, i.e. explainability estimates to interpret in real time the drivers of the disruptivity. DPRF was trained with a dataset of disruptions in which the electron density reached at least 80% of the Greenwald density limit, using the zero-dimensional signal routinely available to the EAST PCS. Through offline analysis, an optimal warning threshold on the DPRF disruptivity signal was found, which allows for a successful alarm rate of 92% and a false alarm rate of 9.9%. By analyzing the false alarm causes, we find that a fraction (∼15%) of the misclassifications are due to sudden transitions of plasma confinement from H- to L-mode, which often occur during high-density discharges in EAST. By analyzing DPRF feature contributions, it emerges that the loop voltage signal is that main cause of such false alarms: plasma signals more apt to characterize the confinement back-transition should be included to avoid false alarms.

Implementation of P-EFIT code for plasma control in EAST is described. P-EFIT is based on the EFIT framework, but built with the CUDA™ architecture to take advantage of massively parallel Graphical Processing Unit (GPU) cores to significantly accelerate the computation. 65 × 65 grid size P-EFIT can complete one reconstruction iteration in 300 μs, with one iteration strategy, it can satisfy the needs of real-time plasma shape control. Data interface between P-EFIT and PCS is realized and developed by transferring data through RFM. First application of P-EFIT to discharge control in EAST is described.

The discharge scenarios on EAST plasma control system (PCS), characterized by different waveform parameters and different hardware requirements, will need a systematic discharge scenario management system for remote and local operators, in order to optimize storage structure and rationally manage discharge time. The remote management of discharge scenarios will require extending the functionalities of the present PCS “future shot” and “next shot” modules. Taking advantage of database technique, the operators can acquire detail information of all discharge scenarios directly without PCS user interface and search the specified scenarios by key words. In addition, the system can elect optimal scenarios automatically based on discharge schedule and plasma pulse setting for later artificial selection. To this purpose, a new remote discharge scenario management system (RDSMS) basis for Web is being conceived on EAST. The system contains a database with functions of “user management”, “scenario verification”, “prepared scenario management”, “actual discharge scenario management” and “discharge schedule management”. This paper will present the relevant conceptual design and give an account of the test results for implementation on EAST discharges.

During 2016–2018 experimental campaigns, the plasma magnetic control architecture of the EAST tokamak was revised in order to achieve improved performances, with the final aim of feedback control of alternative divertor configurations (i.e. with multiple X-points). This paper reports on the results obtained with the Multi-Input–Multi-Output (MIMO) plasma shape controller tested during the last experimental campaign, which prompted a considerable improvement of the control performances. Simulation results are also discussed.

Both theory and experiment have proved Snowflake configuration could reduce the heat loads on divertor plate. Due to limitation of PF coils, EAST could only operate with quasi-snowflake (QSF). In 2014 EAST campaign, QSF has been achieved by RZIp control. The next important task is the QSF shape control. As tokamak discharge simulation code, Tokamak Simulation Code (TSC), which has been benchmarked by experimental data, is used to simulate EAST QSF discharge. Singular Value Decomposition (SVD) method, a way to decouple the PF current and control parameter, is implemented in TSC code to simulate the course of QSF shape control. The simulation results show SVD method is a good way for EAST QSF shape control. © 2015 Elsevier B.V. All rights reserved.

Small perturbations and strong impurity exhaust capability associated with the small grassy ELMs render the grassy-ELM regime a suitable candidate for achieving steady-state H-mode operation with a radiative divertor, especially in a metal-wall device, such as the Experimental Advanced Superconducting Tokamak (EAST). As the degradation of pedestal performance with excessive divertor impurity seeding or accumulation tends to be accompanied with significantly increased radiation near the divertor X point, feedback control of the absolute extreme ultraviolet (AXUV) radiation near the X point has been employed to maintain the confinement property in EAST. However, the absolute value of the AXUV radiation at the outer target varies with plasma conditions as during the divertor detachment process. Thus, a new feedback-control scheme has been recently developed and applied to grassy-ELM H-mode plasmas in EAST to achieve stationary partial detachment while maintaining good global energy confinement with H 98,y2 >1. In this scheme, electron temperatures (T et) measured by divertor Langmuir probes are used to identify the onset of detachment, and then the plasma control system (PCS) switches to the feedback control of one channel of AXUV radiation near the X point, where a steep gradient in the radiation profile is present. The feedback is performed through pulse-width-modulated duty cycle of a piezo valve to seed impurities with mixed gas (50% Ne and 50% D2) from the outer target plate near the strike point in the upper tungsten monoblock divertor. T et near the strike point is maintained in the range of 5–8 eV, and peak surface temperature on the outer target plate (T IR,peak) is suppressed and maintained at ∼180 °C, based on infrared camera measurements. The plasma stored energy maintains nearly constant over the entire feedback-control period. It thus offers a highly promising plasma control scenario suitable for long-pulse high-performance H-mode operation in EAST, which is potentially applicable to future steady-state fusion reactors as an integrated solution for the control of both ELM-induced transient and steady-state divertor heat loads while maintaining good core confinement.

One of the critical challenges for long pulse operation of ITER and future fusion reactors is the excessively high heat and particle fluxes on the divertor targets. The divertor detachment offers an effective way for the control of steady state heat flux and erosion at the divertor target. Dedicated experiments have been systematically performed for both H-mode and L-mode plasmas on EAST with ITER-like W divertor in the last two years to further advance detachment physics understanding. The detachment is identified by the rollover of divertor peak particle flux near the strike point during the density ramping up, which correlates with the reduction of electron temperature down to 5 eV, the increase of Dδ/Dα ratio and radiated power. It has been demonstrated that the more closed divertor has a lower density threshold at the onset of detachment on EAST. Furthermore, the detachment density threshold will be reduced when the strike point moves towards the divertor corner. The onset of detachment exhibits a strong asymmetry between inner and outer targets, with the inner target accessing detachment at lower density due to the lower electron temperature, as expected. The effects of the heating power and plasma current on the detachment onset were also investigated. The experimental results show that the density at detachment onset is lower in L-mode relative to H-mode, and increases with the heating power. The detachment density threshold normalized to the Greenwald density limit is lower at a high plasma current.

For long pulse or steady-state advanced plasma discharges, it is necessary to keep plasma parameters such as plasma shape, loop voltage, plasma pressure in a steady state way. In particular, the heat load on the divertor target must be effectively reduced. By using Low Hybrid Wave (LHW) for the current drive, loop voltage has been feedback controlled while plasma current was controlled by the PF coil current. The control of the plasma pressure has been demonstrated by using LHW. Heat load reduction has been done by radiation and advanced shape configuration. By using impurity seeding with gas puff for the feedforward and Super-Sonic Molecular Beam Injection (SMBI) to feedback the total radiation, radiation can be effectively controlled with slight influence to the core confinement. The quasi-snowflake (QSF) discharge in H-mode has been achieved. It is verified again in H-mode operation, heat load can be effectively reduced under the QSF shape. All these new control algorithms give rise to more assistances to the EAST long pulse operation. © 2018 Elsevier B.V.

This paper deals with the model-based approach that has been adopted to design the plasma magnetic control at the EAST tokamak. Such a design approach, which is based on a linear model for the response of the plasma and of the surrounding coils, has been successfully applied in 2016 to design an alternative solution for the plasma vertical stabilization, the plasma centroid position control, and the controller of the currents in the poloidal field circuits. After an introduction to the linear model for the plasma/circuits response, the proposed control algorithms are presented together with the preliminary results obtained during the 2016 experimental campaign at EAST. A brief description of the ongoing and future design activities is also given.

A ITER-like vertical stabilization (VS) algorithm has been successfully deployed and commissioned at EAST. The proposed algorithm decouples the VS from the plasma shape control, while the algorithms previously implemented to stabilize the EAST plasma exhibit a strong coupling with plasma shape control system. As a consequence, the VS algorithms previously implemented at EAST prevent the deployment of advanced multi-input-multioutput (MIMO) plasma shape control schemes, such as the ones proposed in Albanese et al 2016 (Proc. 2016 IEEE Multi-Conf. System Control (Buenos Aires, Argentina) pp 611-6) and Kolemen et al (2015 J. Nucl. Mater. 463 1186). Indeed, such MIMO controllers rely on the decoupling with the VS system. The proposed ITER-like stabilizes the plasma column (i.e. it controls to zero the plasma vertical speed) on the fastest possible time scale, while leaves the control of the plasma vertical position to the plasma shape controller. Thanks to this frequency separation approach, the plasma shape controller can than be designed starting from the stabilized system, without explicitly taking the VS into account. In this paper we present the implementation details of the adopted solution for the EAST vertical stabilization, together with the results obtained during the 2016 experimental campaign.

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.

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.

Experimental advanced superconducting tokamak (EAST) is an experimental device aiming at steady state plasma operation for fusion research. The values of many discharge parameters, such as plasma shape, position and current must be directly acquired or indirectly evaluated from the magnetic measurements, so the accuracy of magnetic measurements plays an important role in reliable plasma control performance. A method for verifying the key magnetic measurements in real time for each shot is described in this paper. Such magnetics verification will prevent the discharge from a key magnetic signal failure and ensure the quality of a successful discharge. The diagnostics verification algorithm has been implemented in the plasma control system for the EAST. The implementation details and its application in the recent experiment are presented in this paper.

In order to improve the synchronization, flexibility and expansibility of the plasma control on HT-7, a new plasma control system (HT-7 PCS) was constructed. The HT-7 PCS was based on a real-time Linux cluster with a well-defined, robust and flexible software infrastructure which was adapted from DIII-D PCS. In this paper, the hardware structure and system customization details for HT-7 PCS are reported. The plasma position and current control, plasma density control and off-normal event detection, which were realized in separated systems originally, have been integrated and implemented in such HT-7 PCS. All these control algorithms have been successfully validated in the last several HT-7 experiment campaigns. Good control performance has been achieved and the experiment results are discussed in the paper.

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.

The active feedback control of radiated power to prevent divertor target plates overheating during long-pulse operation has been developed and implemented on EAST. The radiation control algorithm, with impurity seeding via a supersonic molecular beam injection (SMBI) system, has shown great success in both reliability and stability. By seeding a sequence of short neon (Ne) impurity pulses with the SMBI from the outer mid-plane, the radiated power of the bulk plasma can be well controlled, and the duration of radiative control (feedforward and feedback) is 4.5 s during a discharge of 10 s. Reliable control of the total radiated power of bulk plasma has been successfully achieved in long-pulse upper single null (USN) discharges with a tungsten divertor. The achieved control range of is 20%–30% in L-mode regimes and 18%–36% in H-mode regimes. The temperature of the divertor target plates was maintained at a low level during the radiative control phase. The peak particle flux on the divertor target was decreased by feedforward Ne injection in the L-mode discharges, while the Ne pulses from the SMBI had no influence on the peak particle flux because of the very small injecting volume. It is shown that although the radiated power increased, no serious reduction of plasma-stored energy or confinement was observed during the control phase. The success of the radiation control algorithm and current experiments in radiated power control represents a significant advance for steady-state divertor radiation and heat flux control on EAST for near-future long-pulse operation.

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 ...

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.

Recent commissioning of two major fully superconducting shaped tokamaks, EAST [Y. Wan, et al., Proc. 21st IAEA Fusion Energy Conf., Chengdu, China, 2006] and KSTAR [Y. K. Oh, et al., Proc. 25th Symp. on Fusion Technology, Rostock, Germany, 2008, O8-3], represents a significant advance in magnetic fusion research. Key to commissioning success in these complex and unique tokamaks was (1) use of a robust, flexible plasma control system (PCS) based on the validated DIII-D design [B. G. Penaflor, et al., Proc. 6th IAEA Tech. Mtg. on Control, Data Acquisition and Remote Participation for Fusion Research, Inuyama, Japan, 2007]; (2) use of the TokSys design and modeling environment, which is tightly coupled with the DIII-D PCS architecture [J. A. Leuer, et al., Fusion Eng. Design, vol. 74, p. 645, 2005], 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 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 superconducting coil current slew rate limits and the presence of non-linear magnetic materials, are discussed, along with commissioning and initial operational results. Application of this same modeling environment to designing startup scenarios and other major control needs for ITER is also discussed.

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.

In view of ITER operations, many experiments are being conducted on the EAST tokamak. In particular, new magnetic control approaches are being developed and need to be tested. In order to design and test a plasma current, position and shape control, an accurate modeling of EAST static equilibria and dynamic plasma evolution is required. Hence, a suite of simulation tools has been developed for the reconstruction of plasma equilibria, the generation of plasma linearized models and for the closed loop testing of magnetic control algorithms. These tools are meant to assess the reliability of such equilibria and models, in order to provide a robust tool for the purposes of control design.