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The plasma control system (PCS) in a tokamak device is in charge of controlling the evolution of plasma parameters according to prescribed models against uncertainties and disturbances. The PCS accomplishes its mission by receiving data from the diagnostics and by computing in real-time the commands to the actuators that affect the plasma behavior. The design of the ITER PCS includes many different aspects, which are not limited to the design of control algorithms, including the definition of the verification and validation tests for various components and commissioning procedures. Moreover, contributions come from different parties that adopt heterogeneous sources. To homogenize these contributions and to keep track of the PCS life-cycle throughout various design stages, a specific system-engineering approach has been adopted on top of the standard ITER life-cycle and the requirement management process. Such an approach relies on a database (DB) implemented using Enterprise Architect that allows modeling various aspects of the PCS design using SysML. This article gives an overview of the adopted approach and describes the structure and the current content of the PCS DB. [ABSTRACT FROM AUTHOR]

Tokamaks are the most promising approach for nuclear fusion on earth. They are toroidal machines where the plasma is heated in a ring-shaped vessel and kept away from the vessel by applied magnetic fields. To achieve high performance in tokamaks, plasmas with elongated poloidal cross-section are needed. Such elongated plasmas are vertically unstable, hence position control on a fast time scale is clearly an essential feature for all tokamak devices. In this context the Plasma Control Upgrade project was aimed at increasing the capabilities of the Vertical Stabilization (VS) of the JET tokamak. This paper introduces the new JET VS system and focuses on how the flexibility of this real-time system has been exploited to enlarge its operational limits in terms of maximum controllable disturbance. Eventually, some experimental results achieved during the last experimental campaigns are presented. [ABSTRACT FROM PUBLISHER]

The achievement of the required performances during the operation of large fusion experimental reactors, is strictly related to the flexibility and reliability of the real-time infrastructures. It turns out that, in tokamak reactors, the real-time infrastructure has to be designed so as to meet a number of common requirements. An overview of the real-time infrastructure currently adopted at the JET tokamak is given in this paper, focusing the attention on the solutions that have been developed for addressing these common requirements. Furthermore, three real-time systems recently deployed at JET are described as example. [ABSTRACT FROM PUBLISHER]

The ITER tokamak (Latin for "the way") is the next step toward the realization of electricity-producing fusion power plants, since it has been designed to reach the plasma burning condition. The Central Safety System for Nuclear Risk (CSS) is the control system in charge to assure nuclear safety for the ITER plant, the personnel, and the environment. Since the CSS is a critical safety system, its validation and commissioning play an important role, and the required level of reliability must be demonstrated. In such a scenario, it is strongly recommended to use modeling and simulation tools since the early design phase. Indeed, mathematical models will help in the definition of the control system requirements. These models can be used for the rapid prototyping of the safety system, and hardware-in-the-loop (HIL) simulations can be performed to assess the performance of the control hardware against a plant simulator. This paper introduces the methodology and the software/hardware architecture used to develop both a CSS prototype and a ITER plant model suitable for the test and validation of this prototype. [ABSTRACT FROM AUTHOR]

The International Thermonuclear Experimental Reactor (ITER) is the next step toward the realization of electricity-producing fusion power plants, and it is planned to be in operation in 2016. ITER has been designed so as to reach the plasma burning condition and to operate with high-elongated unstable plasmas. However, due to the constraints that affect the machine realization, these open-loop unstable high-performance plasmas can hardly be stabilized using the poloidal field coils placed outside the tokamak vessel. For this reason, during the ITER design review phase, it has been proposed to investigate the possibility of using in-vessel coils in order to improve the best achievable performance of the vertical stabilization (VS) system. This paper proposes a new approach for the plasma current, position, and shape control design in the presence of in-vessel coils. Two control loops are designed: a first loop that guarantees the VS driving the voltage applied to in-vessel coils, and a second loop that controls the plasma current and up to 32 geometrical shape descriptors as close as possible to the reference values. The performance of the proposed control system is shown by means of simulations of some cases of interest. [ABSTRACT FROM AUTHOR]

Abstract-This paper describes the design and the experimental validation of a multivariable current and shape controller for the Joint European Torus (JET) tokamak. Using this new controller, physicists have been able to carry out experiments where accurate control of the plasma boundary is needed. From a control point of view, the problem has been formulated as an output regulation problem for a linear time-invariant (LTI) plant whose controllable outputs are more than the control inputs. For the case of constant references, we propose a control scheme that minimizes a quadratic cost function. This cost function weights the tracking error at steady state. Our methodology is based on the singular value decomposition of the static gain matrix of the plant. In the controller design, we can also take into account the steady-state control effort. The proposed controller has been implemented and tested on the plant; some experimental results are included in this paper. The activity that has been carried out and that is described in this paper is relevant in the perspective of the new experimental reactor ITER. [ABSTRACT FROM AUTHOR]

The high-energy ionized particles collected by the divertor structures in a tokamak reactor cause a localized thermal load around the strike-points, which are the intersections of the separatrix with the divertor tiles. To spread on a larger region this thermal load, it is convenient to resort to a sweeping, i.e., a periodical movement of the strike-points. This paper introduces the new model-based sweeping algorithm recently implemented on the Joint European Torus tokamak, within the eXtreme Shape Controller architecture. [ABSTRACT FROM AUTHOR]

This paper describes a set of graphic tools for the design and validation of the new plasma shape controller [eXtreme Shape Controller (XSC)], which has been recently implemented at the Joint European Torus tokamak. The XSC enables operation with high elongation and high triangularity plasmas. The software suite, which is called XSC Tools, has been developed to automate the design procedure of the XSC. These tools make use of graphic user interfaces to allow the nonspecialist users to prepare for new operative scenarios, without the help from modeling and control specialists. Once a new controller is generated, all its parameters are saved into a text file, which is used to perform the validation of the scenario via simulations. The same file is then loaded by the real-time code running on the plant, without any further processing. This feature guarantees that the controller, which is running on the plant, is exactly the same one validated through simulations. An additional effort has been made in order to make the XSC Tools machine independent, i.e., to enable their use on the different tokamaks. [ABSTRACT FROM AUTHOR]

Describes the design and experimental validation of a multivariable digital controller for a Tokamak. Linearized model of the plasma confined in a Tokamak; Scope of the controller; Nested loops composing the proposed controller; Improvement of the performance with respect to those obtained with a proportional integral derivative.

Presents an approach for the design of a plasma current and shape feedback control system in a Tokamak. Enlargement of the operating envelope in which the Tokamak can operate without current saturation; Properties of the two feedback loops that comprise the proposed controller structure.

Presents a study which focused on the estimation of the minimum poloidal field coil voltages for the plasma shape control in a tokamak. Problem description; Plasma model; International Thermonuclear Experimental Reactor (ITER) tokamak case.