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As part of the ITER design review, a reassessment of the specifications underlying the design of the vertical stabilization system (VS) was performed. Recent results from experiments, aimed at the evaluation of the feasibility of the ITER reference scenarios, have raised several concerns regarding mostly the ramp-up and ramp-down phases of the pulse. The main issue is the value of the internal inductance li which may reach values outside the range 0.7–1, considered as reference for the ITER control system design. Similar concerns apply to the low current L-mode plasmas, needed to the exploitation of the machine towards the development of the 15 MA pulse. The performance of the reference vertical stabilization system, under the revised conditions may be marginal, in particular if the effect of plasma generated noise on the velocity measurement is considered. A reliable and robust VS is mandatory to guarantee the operation of ITER at the reference elongation and plasma current values. To avoid de-scoping of the machine mission, several solutions have been proposed to improve the VS performances, ranging from an upgrade of the maximum voltage available to the present external coils system, to the introduction of in-vessel passive and/or active conductors. The paper presents an overview of the modelling and experimental effort aimed at the assessment of the baseline ITER VS and analyses the proposed solutions to improve the system performance.

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.

Real-time simultaneous control of several radially distributed magnetic and kinetic plasma parameters is being investigated on JET, in view of developing integrated control of advanced tokamak scenarios. This paper describes the new model-based profile controller which has been implemented during the 2006–2007 experimental campaigns. The controller aims to use the combination of heating and current drive (H&CD) systems—and optionally the poloidal field (PF) system—in an optimal way to regulate the evolution of plasma parameter profiles such as the safety factor, q(x), and gyro-normalized temperature gradient, . In the first part of the paper, a technique for the experimental identification of a minimal dynamic plasma model is described, taking into account the physical structure and couplings of the transport equations, but making no quantitative assumptions on the transport coefficients or on their dependences. To cope with the high dimensionality of the state space and the large ratio between the time scales involved, the model identification procedure and the controller design both make use of the theory of singularly perturbed systems by means of a two-time-scale approximation. The second part of the paper provides the theoretical basis for the controller design. The profile controller is articulated around two composite feedback loops operating on the magnetic and kinetic time scales, respectively, and supplemented by a feedforward compensation of density variations. For any chosen set of target profiles, the closest self-consistent state achievable with the available actuators is uniquely defined. It is reached, with no steady state offset, through a near-optimal proportional-integral control algorithm. Conventional optimal control is recovered in the limiting case where the ratio of the plasma confinement time to the resistive diffusion time tends to zero. Closed-loop simulations of the controller response have been performed in preparation for experiments, and typical results are shown. Finally, in the last section of the paper, the first experimental results using this dynamic-model approach to control the plasma current and the safety factor profile on JET, either with the three H&CD systems or also with the PF system as an additional actuator, are presented and discussed.

The International Thermonuclear Experimental Reactor (ITER) is the next step toward the realization of electricity-producing fusion power plants. 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 which affect the machine realization, these open-loop unstable high performance plasmas can be hardly stabilized using the Poloidal Field (PF) 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 system. Because of some technological differences between the in-vessel coils and the PF coils (the former cannot be superconductive), the controller design procedure previously adopted in [1] cannot longer be used. This paper proposes a new approach for the plasma current, position, and shape control design in the presence of in-vessel coils. In particular two control loops are designed: a first loop which guarantees the vertical stabilization by means of a MISO controller which drives the voltage applied to in-vessel coils; a second MIMO loop controls the plasma current and up to 32 geometrical shape descriptors as close as possible to the reference values.

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.

Simultaneous control of the plasma shape, the magnetic and kinetic plasma profiles (such as the safety factor, q(x), and gyro‐normalized temperature gradient, ρTe*;(x), respectively) and the boundary flux is being investigated on JET, and has potential applications in the operation of ITER steady state advanced tokamak discharges. The control of radially distributed parameters was achieved for the first time on JET in 2004 [1–4]. The controller was based on the static plasma response only. The approach newly implemented on JET aims to use a dynamical plasma model, all the available heating and current drive (H&CD) systems, and the poloidal field (PF) system in an optimal way to achieve a set of requested magnetic and kinetic profiles. This paper describes the new model‐based optimal profile controller which has been tested during the last 2007 experimental campaign. The controller aims to use the combination of heating and current drive systems ‐ and optionally the PF system. First experimental results of...

This paper describes the installation of the new JET Shape Controller System [M. Garribba, R. Litunovsky, P. Noll, S. Puppin, The new control scheme for the JET plasma position and current control system, in: Proceedings of the 15th SOFE Conference, Massachusetts, 1993, pp. 33–36; F. Sartori, A. Cenedese, Plasma position and current control management at JET, in: Proceedings of the 42nd IEEE Conference on Decision and Control, Maui, 2003] especially focusing on the addition of the Extreme Shape Controller [G. Ambrosino, et al., A new shape controller for extremely shaped plasmas in JET, Fusion Eng. Des. 66–68 (2003) 797–802]. The activity was performed by the JET Operator in co-operation with the ENEA-CREATE design team, and involved both changes in the hardware and system software of JET and tuning of the proposed Extreme Shape Controller (XSC) design to satisfy the practical requirements of tokamak operation. The application of 10 years experience of controller implementation and commissioning combined with a modern and efficient modelling and design methodology has allowed an unprecedented fast and easy commissioning of the new system.

This paper deals with the problem of designing a plasma current, position, and shape controller for the ITER-FEAT tokamak. Exploiting the features of the plant, we show that it is possible to design separately two control loops: a first loop which stabilizes the vertical position by means of a simple derivative action, and a second loop which drives the plasma current and the plasma geometrical shape descriptors as close as possible to the reference values by means of a multi-variable PI controller. The reduced-order model for the design of the external loop has been obtained via a singular perturbation decomposition.

This paper deals with the design of a new plasma current and shape controller for the JET tokamak. This new controller is aimed at improving the performance of the present controller so as to allow the control of extremely shaped plasmas with higher values of elongation and triangularity. The design approach makes use of a linearized mathematical model linking the voltages applied to the active coils with a number of geometrical descriptors characterizing the plasma shape. To describe accurately the plasma shape a number of geometrical descriptors, greater than the number of available active circuits, is needed. Hence the control problem under investigation is characterized by the fact that the number of parameters to be controlled is larger than the number of control inputs. To overcome this problem a singular value analysis is carried out to identify the principal directions of the algebraic mapping between coil currents and geometrical descriptors. These principal directions are then assumed as controller outputs so as the original multivariable (rectangular) control problem is transformed in a standard square problem. Moreover the singular value decomposition approach, makes it possible to solve this modified problem by means of a number of SISO problems. Each of this SISO control problem has been solved designing a suitable PID controller.

In this paper a mixed PID/H/sub /spl infin// discrete-time controller is proposed for the control of the plasma shape and position in a tokamak. This controller is designed to be tested experimentally on the TCV tokamak, using high performance digital hardware. The challenge of the problem lies in the instability of the vertical position, the non-minimum phase characteristics and the nonlinearity of the plant. The inner PID loop stabilises the plant, providing adequate stability margins, while the outer-loop H/sub /spl infin// controller provides the decoupling between the five controlled parameters. Simulations carried out using a validated linear model demonstrate a significant improvement of the performance compared with that obtained with an experimentally tuned MIMO controller.

In this paper, a short review of the work done in the framework of a nation-wide research programme on 'Models and Methods for Plasma Control in Magnetically Confined Fusion Experiments' is presented. The broad aim of the overall programme is to develop and propose a new effective and reliable approach to the on-line plasma control for future fusion experiments, starting from the today's theoretical background, validated by experimental evidence from a number of tests performed on existing experiments. The proposed formulation to approach the control problem is a linearized model in terms of suitable state variables and input/output relationships. The basic project has been subdivided into four major areas of investigation: the linearized response plasma model, the three-dimensional electromagnetic model, the identification techniques and finally the plasma control requirements. The most remarkable results, achieved so far in each area above, are presented in the paper. (C) 2001 Elsevier Science B.V. All rights reserved.

This paper describes the design and the experimental validation of a multivariable digital controller for a tokamak, the TCV. The design of the controller is based on a linearized, validated model of the plasma confined in the tokamak. The plant is MIMO and the various outputs are strongly coupled. Moreover the plant is open-loop unstable. The scope of the controller is to stabilize the plasma and to guarantee some closed-loop performance in terms of decoupling among the plant outputs. The proposed controller is composed of two nested loops: one is devoted to the vertical stabilization, the other, designed using the H/sup /spl infin// technique, guarantees the control of the plasma current and of the plasma shape. After massive simulations, this controller was successfully tested on the plant. The experimental results show a significant improvement of the performance with respect to those obtained with a PID MIMO controller, that was used before on the plant.

In this paper we present an approach for the design of a plasma current and shape feedback control system in a tokamak which takes explicitly into account the problem of current saturation in the coils designated for the control action. The proposed design technique is essentially aimed to enlarge the operating envelope in which the tokamak can operate without current saturation. The proposed controller structure consists of two feedback loops: one is a standard LTI controller designed to operate in the absence of current saturation; the other one gracefully degrades the system performance when the coil currents approach their saturation limits. A case study is presented.

In this paper we deal with the output regulation problem for an LTI plant whose controllable outputs are more than the control inputs. For the case of constant references, we present a two-degree-of-freedom control scheme which 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. Then we show how, introducing some stronger assumptions, it is possible to modify the controller so as to reduce the steady-state control effort.

This paper deals with the problem of designing a plasma current, position, and shape controller for the ITER-FEAT tokamak. Exploiting the features of the plant, we show that it is possible to design separately two control loops: a first loop which stabilizes the vertical position by means of a simple derivative action, and a second loop which drives the plasma current and the geometrical shape descriptors as close as possible to the reference values. The second loop is designed by assigning the dominant closed loop poles.