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A new model-based plasma current and shape controller has been set up and tested on the JET Tokamak with the existing active circuits and control. The installation has been carried out without causing any interference to the plasma operation and without requiring a long commissioning time. Eventually, the new controller was used on really extremely shaped internal transport barrier experiments at high poloidal beta and in the presence of quite large variations of the plasma current density profile (variation range Δβpol up to 1.5 and Δli up to 0.5). The extreme shape controller (XSC) controller architecture and philosophy also offer new interesting opportunities, e.g., the separatrix sweeping on the divertor plates without significantly affecting the overall plasma shape, and the possibility of improving the overall tokamak performance via combined control of plasma shape, current and profile. The adopted methodology constitutes also an important test bed for feedback control strategies of ITER relevance.

This paper gives an overview of the scope and context of the CODAC high-level real-time applications (Supervision and Plasma Control) and presents the strategy and current state of design of the tools to support the implementation. A real-time framework, which is currently under development with strong support of the worldwide fusion community will not only support the implementation of plasma control strategies with the extensive exception handling and forecasting functionality foreseen for ITER, but also integrated commissioning, orchestration and supervision as well as the real-time needs of ITER plant system developers. A second cornerstone in the implementation strategy is the development of a powerful simulation environment (Plasma Control System Simulation Platform - PCSSP) to design and verify control strategies, event handling and orchestration and automation. The development of PCSSP is currently under contract and this paper will also give an overview of its current state of development. (C) 2015 ITER Organization. Published by Elsevier B.V. All rights reserved.

The present paper deals with concept, prototyping and application of a tensioning system for FRP ties into masonry structures. The proposed system, based on the use of FRP strands instead of traditional steel ties, has the aim to produce a compression stress state on masonry walls where it is applied. Given the objective difficulty in tensioning a FRP strand, it was necessary to both characterize and prototype a suitable connection system between the strand and the pulling system. The experimental phase concerned both the manufacturing of the pulling system and the study of used materials, as well as the characterization of the impregnation technology of FRP ties. The above described system has been produced and used in the framework of the structural retrofitting of the “Real Albergo dei Poveri” building in Naples.

In view of the preparation for the operation of the ITER tokamak it is necessary to develop the plasmascenarios taking into account all engineering constraints coming from the plant and including a realisticcontrol system. It is important to consider that, due to the high energy of ITER plasmas, much morestringent requirements are posed on the control of transients in order to avoid machine damage.Several activities are performed in the EU focusing on one side on the scenario optimization from aphysics point of view and on the other side on the design and modeling of a realistic plasma controlsystem driving the plasma configuration throughout the whole pulse and suitable for implementationon a real machine.The issues related to the computation of the control feed-forward component are addressed. In par-ticular, the possibility to trigger a feed-forward component to solve controllability problems arising inthe transitions from plasma L to H and H to L modes is studied in detail with the support of linear andnon-linear simulations.A control strategy is designed and tested on non-linear simulations of the whole pulse, including linearand non-linear effects due to controller switching, plasma shape reconstruction and power supplies.The paper reports on the results of the studies performed and discuss the proposed design of the plasmacontrol system.© 2014 Published by Elsevier B.V.

A simulation environment known as the Plasma Control System Simulation Platform (PCSSP), specifically designed to support development of the ITER Plasma Control System (PCS), is currently under construction by an international team encompassing a cross-section of expertise in simulation and exception handling for plasma control. The proposed design addresses the challenging requirements of supporting the PCS design. This paper provides an overview of the PCSSP project and a discussion of some of the major features of its design. Plasma control for the ITER tokamak will be significantly more challenging than for existing fusion devices. An order of magnitude greater performance (e.g. [1] , [2] ) is needed for some types of control, which together with limited actuator authority, implies that optimized individual controllers and nonlinear saturation logic are required. At the same time, consequences of control failure are significantly more severe, which implies a conflicting requirement for robust control. It also implies a requirement for comprehensive and robust exception handling. Coordinated control of multiple competing objectives with significant interactions, together with many shared uses of actuators to control multiple variables, implies that highly integrated control logic and shared actuator management will be required. It remains a challenge for the integrated technologies to simultaneously address these multiple and often competing requirements to be demonstrated on existing fusion devices and adapted for ITER in time to support its operational schedule. We describe ways in which the PCSSP will help address these challenges to support design of both the ITER PCS itself and the algorithms that will be implemented therein, and at the same time greatly reduce the cost of that development. We summarize the current status of the PCSSP design task, including system requirements and preliminary design documents already delivered as well as features of the ongoing detailed architectural design. The methods being incorporated in the detailed design are based on prior experience with control simulation environments in fusion and on standard practices prevalent in development of control-intensive industrial product designs.

This paper reports on a recently proposed dynamic allocation technique that can be effectively adopted to handle the current saturations of the Poloidal Field coils with the eXtreme Shape Controller. The proposed approach allows to automatically relax the plasma shape regulation when the reference shape requires current levels out of the available ranges, finding in real-time an optimal trade-off between shape control precision and currents saturation avoidance. In this paper the results attained during preliminary analysis are presented, showing the advantage arising from the use of the dynamic allocator, versus the bare use of the eXtreme Shape Controller.

Fusion advanced studies torus (FAST) is a proposal for a satellite facility which can contribute the rapid exploitation of ITER and prepare ITER and DEMO regimes of operation, as well as exploiting innovative DEMO technology. FAST is a compact (R(0) = 1.82 m, a = 0.64 m, triangularity delta = 0.4) machine able to investigate non-linear dynamics effects of alpha particle behaviours in burning plasmas [1,2,5]. The project is based on a dominant 30 MW of ion cyclotron resonance heating (ICRH), 6 MW of lower hybrid (LH) and 4 MW of electron cyclotron resonance heating (ECRH). FAST operates at a wide range [3,4] of parameters, e.g., in high performance H-mode (B(T) up to 8.5T: I(p) up to 8 MA) as well as in advanced Tokamak operation (I(p) = 3 MA), and full non-inductive current scenario (I(p) = 2 MA). Helium gas at 30K is used for cooling the resistive copper magnets [6]. That allows for a pulse duration up to 170s. To limit the TF magnet ripple ferromagnetic insert have been introduced inside the vacuum vessel (VV). Ports have been designed to also accommodate up to 10 MW of negative neutral beam injection (NNBI). Tungsten (W) or liquid lithium (L-Li) have been chosen as the divertor plates material, and argon or neon as the injected impurities to mitigate the thermal loads. (C) 2009 Elsevier B.V. All rights reserved.

The Plasma Control System Simulation Platform (PCSSP) is a highly flexible, modular, time-dependent simulation environment developed primarily to support development of the ITER Plasma Control System (PCS). It has been under development since 2011 and is scheduled for first release to users in the ITER Organization (IO) and at selected additional sites in 2015. Modules presently implemented in PCSSP enable exploration of axisymmetric evolution and control, basic kinetic control, and tearing mode suppression. A basic capability for generation of control-relevant events is included, enabling study of exception handling in the PCS, continuous controllers, and PCS architecture. While the control design focus of PCSSP applications tends to require only a moderate level of accuracy and complexity in modules, more complex codes can be embedded or connected to access higher accuracy if needed. This paper describes the background and motivation for PCSSP, provides an overview of the capabilities, architecture, and features of PCSSP, and discusses details of the PCSSP vision and its intended goals and application. Completed work, including architectural design, prototype implementation, reference documents, and IO demonstration of PCSSP, is summarized and example use of PCSSP is illustrated. Near-term high-level objectives are summarized and include preparation for release of an "alpha" version of PCSSP and preparation for the next development phase. High-level objectives for future work are also discussed. © 2015 Elsevier B.V. All rights reserved.

The JET Shape Controller (SC) uses nine distinct circuits, powering the JET poloidal field coils, to control in real time the coil currents, and the plasma shape, current and position. The control scheme presently used [1] is based on a Multiple Input Multiple Output (MIMO) controller, which is designed to decouple the inductive coupling of the different coils. Achieving such a decoupling, the SC allows the user to tune independently the time response of each circuit. As a matter of fact the intended decoupling algorithm has been incorrectly coded in the JET SC system. This paper describes the modelling and experimental activities performed to correct the code error, and to improve the performance on a subset of the controlled parameters. (C) 2015 Elsevier B.V. All rights reserved.

In a tokamak the magnetic diagnostics are key to the exploitation of the machine. They play a central role in the real-time control of fundamental plasma properties, such as the plasma shape and position, while also contributing with important data to a better understanding of the plasma physics. One of the particular challenges of the ITER magnetics diagnostic is the need to balance high system reliability with sufficient freedom to tune and improve the quality of the diagnostic physics output. This requirement calls for a design pattern where the functions related to plasma control and protection are loosely coupled with the functions related to the plasma science. This work reports on the current status of the magnetics plant system controller design and discusses some possible design solutions that address the aforementioned issue.

A new model-based shape controller (XSC, i.e., eXtreme Shape Controller) able to operate with high elongation and triangularity plasmas has been designed and implemented at JET in 2003. The use of the XSC needs a number of steps, which at present are not automated and therefore imply the involvement of several experts. To help the session leader in preparing an experiment, a number of software tools are needed. The paper describes the SW tools that are currently in the developing phase, and describes the new framework for the preparation of a JET experiment.

ITER is targeting Q = 10 with 500 MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination. Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios. Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design. Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the plasma boundary), and the strong component of eddy current at the start-up (resulting in a poor S/N ratio for plasma reconstruction for Ip < 2 MA requiring a robust plasma control) make the ITER magnetic diagnostic system a demanding part of the magnetic control and investment protection systems. Finally the paper will illustrate the identified roles of magnetic control in the PCS (plasma control system) as formally defined in the recent first step of the design and development of the system.

The plasma control system simulation platform (PCSSP) for ITER shall support the analysis and development of methods to be used by the ITER plasma control system (PCS) for handling exceptions to optimize pulses and assist in machine protection. PCSSP will permit to investigate physical and technical events, such as component failures, control degradation, operation domain excess, plasma state bifurcation or instabilities, and interlock activity. Serving that purpose, the plasma, actuator, diagnostics and PCS simulation modules in PCSSP will be enhanced to compute nominal and off-normal data. Configured by an event schedule, an event generator will orchestrate the activation and manipulate the characteristics of such off-normal computation. In the simulated PCS exceptions will be handled in a pulse supervision layer operating on top of the pulse continuous control (PCC) feedback loops. It will monitor events, decide on which exceptions to respond, and compute new control references to modify PCC behavior. We discuss basic concepts for the event generation in PCSSP, and a preliminary architecture for exception handling in PCS, and show how these will be configured with event and pulse schedules.

This paper proposes an approach aimed at avoiding or mitigating the consequences of a plasma disruption in a tokamak. The control strategy is based on a Disruption Predictive Control (DPC) where the use of a set of disruption predictors could allow actively modification of some plasma parameters when a disruption warning is issued. The DPC approach requires the continuous monitoring of various plasma performance indicators, which play the role of disruption precursors. The active control is based on the availability of magnetic and nonmagnetic plasma parameter response models as well as multivariable control algorithms and architectures. © 2012 - IOS Press and the authors.

A project dedicated to the enhancement of the JET vertical stabilization system was launched in 2006, including an upgrade of the Power Supply of the Radial Field Amplifier, of hardware and software of the vertical stabilization control system. The main aim was to double the JET capability in stabilising high current plasmas when subject to perturbations, in particular large Edge Localised Modes. We present here the results of first plasma operation with the new Enhanced Radial Field Amplifier and its data acquisition and control system, focussing on the benefits of an approach based on phased commissioning, modelling and offline algorithm validation. (C) 2011 EURATOM. Published by Elsevier B.V. All rights reserved.

The JET enhancement project Plasma Control Upgrade (PCU) aimed at increasing the capabilities of the plasma vertical stabilization (VS) system. One of the activities of this project was devoted to the development of simple but sufficiently accurate models of the VS system so as to address the main design choices, use the simulation tools as reliable test-beds, and provide an adequate support to the engineering design and commissioning of the new Enhanced Radial Field Amplifier (ERFA). This paper illustrates some of the main achievements of the modelling activity, which gave rise to a closed loop model of the VS system, including plasma, PF coils and passive structures. In particular the paper deals with the selection of the set of turns to be used in the control coils and with the estimation of the eddy current effects on the VS system. The latter analysis addressed an upgrade of the converter units of ERFA, successfully implemented during its commissioning on plasma in August 2009.

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.

In this paper we present a dynamic model for the analysis of the vibrations of the vibrations of a planar articulated flexible boom to be used for tokamak in-vessel maintenance operations. The peculiarity of the mechanical structure of the boom enables us to consider separately the oscillations in the horizontal and vertical planes so that two separate models can be constructed for describing these phenomena. The results of simulations based on booms like that proposed for NET in-vessel operations are presented.

Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with Ip^4. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current Ip of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7–0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.

The capability of the poloidal field coil system, as presently designed, to provide an effective vertical stabilization of the plasma in the IGNITOR machine has been investigated using the CREATE-L response model. An optimization of the vertical position control strategy has been carried out and the most effective coil combination has been selected to stabilize the plasma while fulfilling engineering constraints on the coils and minimizing the required power and voltage. The growth rate of the vertical instability and the power required by the active stabilization system has been estimated with this model. The possible failure of the relevant electromagnetic diagnostics has been taken into account, evaluating the robustness of the plasma position reconstruction strategy. A realistic description of the power supply system has permitted to carry out the optimization of the proportional-integrative-derivative (PID) controller, both with a voltage and a current loop control scheme. An assessment of the requirements for the plasma cross section shape control has been carried out considering perturbations of the plasma global parameters independent of each other and showing that the undesired shape modification rejection is possible with the present PFC and power supply system. The PF coils have been rated relative to their capability to restore shape modifications due to different plasma disturbances. The most effective coil combination, that minimizes recovery time and voltage required, has been identified. (C) 2007 Elsevier B.V. All rights reserved.

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 describes the development and experimental validation of a simulation model for the design of FTU plasma radial position and current controllers. These controllers have been designed to be tested experimentally on the FTU tokamak. The results predicted in simulation were actually delivered during FTU operation, in two different discharges.

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.

As part of the ITER Engineering Design Activity, a number of questions related to plasma control has been addressed, using linearised and non-linear simulation codes to assess the control of the plasma shape given the particular design restrictions of ITER.

The ITER Central Interlock System (CIS) executes manual and automatic interlocks in order to protect tokamak’s investments. When a new control system, such as the CIS, has to be deployed, it is strongly recommended to use modeling and simulation tools, in order to test and validate the control logics. This paper presents the software tools that have been recently developed to perform the modeling of the ITER CIS. In particular, the architectures of both a CIS oriented plant simulator and of a CIS logical prototype are presented, together with the simulation of two test cases.

The Central Safety System for Nuclear Risk (CSS-N) coordinates the safety control systems to ensure nuclear safety for the ITER complex. Since the CSS-N is a safety critical system, its validation and commissioning play a very important role; in particular 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, the modeling tools will help in the definition of the control system requirements. Furthermore the models can than be used for the rapid prototyping of the safety system. Hardware-in-the-loop simulations can also be performed in order to assess the performance of the control hardware against a plant simulator. The proposed approach relies on the availability of a plant simulator to develop the prototype of the control system. This paper introduces the methodology used to design and develop both the CSS-N Oriented Plant Simulator and the CSS-N Prototype.

The development of linear model of the present JET (Joint European Torus) electromagnetic system has allowed the design of an optimized Plasma Shape Controller. Quite extreme plasma boundaries (elongation k = 1.9 and triangularity = 0.6) are shown to be controllable also in the presence of large variations of the values of poloidal beta (0.1 < beta(p) < 0.6) and/or internal magnetic inductance (0.8 < l(i) < 1.4). The developed model gives also an excellent simulation of the growth rate of the experimental vertical disruptions. (C) 2003 Elsevier Science B.V. All rights reserved.

A design procedure for the synthesis of an adaptive controller for a robotic manipulator is presented. The design involves the evaluation of a nominal input signal by using a model of the manipulator with nominal values of the loads, and of an adaptive component by means of the sole knowledge of the maximum possible size of the uncertain parameters of the model. The proposed control scheme also guarantees an accurate tracking of a planned path when an approximate model of the manipulator is used and/or load variations occur during operation. Moreover, the resulting control signal is smooth.

The aim of this paper is to present some new tools used to review the capability of the ITER Poloidal Field (PF) system in controlling the broad range of plasma configurations presently forecasted during ITER operation. The attention is focused on the axi-symmetric aspects of plasma magnetic configuration control since they pose the greatest challenges in terms of control power and they have the largest impact on machine capital cost. Some preliminary results obtained during ongoing activities in collaboration between ENEA/CREATE and EFDA are presented. The paper is divided in two main parts devoted, respectively, to the presentation of a procedure for the PF current optimisation during the scenario, and of a software environment for the study of the PF system capabilities using the plasma linearized response. The proposed PF current optimisation procedure is then used to assess Scenario 2 design, also taking into account the presence of axisymmetric eddy currents and possible variations of poloidal beta and internal inductance. The numerical linear model based tool derived from the JET oriented eXtreme Shape Controller (XSC) tools is finally used to obtain results on the strike point sweeping in ITER. © 2007 Elsevier B.V. All rights reserved.