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The treatment of segmental tibial and ankle bone loss after radical surgery for chronic osteomyelitis is one of the most challenging problems encountered by orthopaedic surgeons [...]

Control of a Tokamak requires operating many sophisticated control loops in a dynamic sequence of activities. To take appropriate actions in case technical or physics conditions occur unexpectedly, the continuous control must be backed up by Exception Handling (EH) logic. To mature the Conceptual Design of the ITER Plasma Control System (PCS) with such logic for the Preliminary Design, we studied, in a formal process, how the plasma will be controlled during the 1 st plasma and early operation phases of ITER and analysed the required control and EH functions and dependencies: Three classes of Exception Handling were identified which cover all use cases: modification of the control behaviour of a single control function, modification of the control structure of connected controllers, and change of the control goal which modifies the overall control system. The three classes form a EH hierarchy from low to high impact responses that can be implemented as local EH in the Pulse Continuous Control layer and as central EH in the Pulse Supervision layer of the PCS.

The ITER plasma control system (PCS) has successfully completed its final design for the first plasma (FP) and engineering operations (EOs) phase and plant system commissioning has begun as ITER prepares for this first operation phase. Commissioning of the essential plant systems will continue as each plant system is completed and made ready for operation. Tokamak assembly has begun with the base and lower cylinder of the cryostat and the lower most poloidal field (PF) coil installed in the tokamak pit. The first vacuum vessel (VV) sector and accompanying two toroidal field (TF) coils are being prepared for transfer to the pit. Once the tokamak is assembled, the cryostat top lid is closed and pump down begins, this will start approximately one year of integrated commissioning (IC) to prepare all of the relevant plant systems for FP operation. After a scheduled one month of plasma operation with the goal of achieving a plasma current > 100 kA for at least 100 ms, there will be about six months of EO to complete commissioning of the superconducting central solenoid, PF, and TF magnet systems to full current, without plasma, to complete this initial ITER operation phase. The PCS final design for FP will be described as well as the IC sequence of the main plant systems required for this operations phase. The plans for the FP operation campaign will be described, including specific challenges present on ITER due to large VV eddy currents, issues associated with electron cyclotron heating (ECH) assist, neutral pressure and impurities. This will be followed by the EO phase to commission full current operation of the superconducting magnets, possibly including plasma operation at full TF of 5.3 T to have improved conditions both for Ohmic plasma initiation and ECH absorption.

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.

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.

Position and number of the magnetic sensors affect the accuracy of the plasma shape identification algorithms. These algorithms usually solve an inverse problem based on the solution of the Grad-Shafranov equilibrium equation, and they need as input a certain number of magnetic field and flux measurements in points located around the plasma. The scope of this paper is to investigate on the problem of assessing the minimum number of magnetic sensors that are necessary for the solution of the plasma shape identification in a tokamak machine. In particular, with reference to the ITER inner vessel magnetic diagnostic system, it is shown that, adopting a heuristic procedure, the available measurements are divided in two subsets, where the measurements in the first subset can be reconstructed by means of suitable linear combinations of measurements in the second subset. In this way, it is possible to evaluate the level of the redundancy of the ITER inner vessel diagnostic system and to outline a suitable fault management system.

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.

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.

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.

Plasma current, position and shape control systems and modeling approaches were reviewed. The literature was divided into three categories, namely plasma radial position and current control, vertical stabilization of elongated plasmas, and plasma shape control. The plasma current and radial position control problem can be easily solved using a filament plasma model - the Shafarov lumped parameter equation and two separate PID controllers. At most, a MIMO controller can be used if decoupling is desired. Stabilization of the plasma vertical motion requires new control strategies in which the control algorithm is changed on the basis of an estimate of the measurement accuracy. For controlling the shape during the plasma formation and start-up phase, controllers based on magnetic flux control are useful when control requirements are not stringent.

Describes the design and the experimental validation of a multivariable digital controller for a Tokamak, the Tokamak a configuration variable (TCV). The design of the controller is based on a linearized model of the plasma confined in the Tokamak. The plant is multiple-input-multiple-output (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 /spl Hscr//sub /spl infin// technique, guarantees the control of the plasma current and of the plasma shape. After massive simulations, this controller has been successfully tested on the plant. The experimental results show a significant improvement of the performance with respect to those obtained with a proportional integral derivative (PID) MIMO controller, that was used before on the plant.

We present a new approach for the design of a plasma current and shape feedback control system in a Tokamak. This approach takes explicitly into account the problem of current saturation in the coils which are used for the control action. The proposed design technique is essentially aimed at enlarging 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 linear time invariant controller designed with the H/sub /spl infin// technique, which operates in the absence of current saturation; the second gracefully degrades the system performance when the coil currents approach their saturation limits. Some stability considerations on the overall control system are given, and a case study is presented.

A high-order, multivariable, modern plasma controller has been designed using H-infinity optimal control techniques and tested oil the Tokamak Configuration Variable (TCV) tokamak. An initial design for the control of the plasma current, position, and shape parameters is described. The design process was based on the CREATE-L linearized model of TCV, and the controller was implemented oil a digital processor. The results demonstrated that the required performance was delivered and the controller response was in good agreement with predictions using the model.

A new approach to the integrated modelling of a mechanical structure and piezoelectric plates is presented. Tn particular a beam-like structure is considered, i.e. a structure which may be modelled by means of the elastic line equations. It is shown how to include into the beam model the interaction with the piezoelectric sensors and actuators dynamics and with the dynamics of the electric network at which the piezoelectric devices are connected. The resulting model is particularly useful for the design of control systems.

The scientific goal of the Ignitor experiment is to approach, for the first time, the ignition conditions of a magnetically confined D-T plasma. The IGNIR collaboration between Italy and Russia is centred on the construction of the core of the Ignitor machine in Italy and its installation and operation within the Triniti site (Troitsk). A parallel initiative has developed that integrates this programme, involving the study of plasmas in which high-energy populations are present, with ongoing research in high-energy astrophysics, with a theory effort involving the National Institute for High Mathematics, and with INFN and the University of Pisa for the development of relevant nuclear and optical diagnostics. The construction of the main components of the machine core has been fully funded by the Italian Government. Therefore, considerable attention has been devoted towards identifying the industrial groups having the facilities necessary to build these components. An important step for the Ignitor programme is the adoption of the superconducting MgB2 material for the largest poloidal field coils (P14) that is compatible with the He-gas cooling system designed for the entire machine. The progress made in the construction of these coils is described. An important advance has been made in the reconfiguration of the cooling channels of the toroidal magnet that can double the machine duty cycle. A facility has been constructed to test the most important components of the ICRH system at full scale, and the main results of the tests carried out are presented. The main physics issues that the Ignitor experiment is expected to face are analysed considering the most recent developments in both experimental observations and theory for weakly collisional plasma regimes. Of special interest is the I-regime that has been investigated in depth only recently and combines advanced confinement properties with a high degree of plasma purity. This is a promising alternative to the high-density L-regime that had been observed by the Alcator experiment and whose features motivated the Ignitor project. The provisions that are incorporated in the machine design, and in that of the plasma chamber in particular, in order to withstand or prevent the development of macroscopic instabilities with deleterious amplitudes are presented together with relevant analyses. © 2013 IAEA, Vienna.

The eXtreme Shape Controller (XSC) has been originally designed to control the plasma shape at JET during the flat-top phase, when the plasma current has a constant value. During the JET 2012 experimental campaigns, the XSC has been used to improve the shape control during the transient phases of plasma current ramp-up and ramp-down. In order to avoid the saturation of the actuators with these transient phases, a Current Limit Avoidance system (CLA ) has been designed and implemented. This paper discusses the CLA algorithm and presents the experimental results achieved at JET during the 2012 campaigns using the XSc. © 2013 IEEE.

In magnetically confined fusion plasmas, feedback control of plasma parameters is assuming an increasingly important role. The complexity of the phenomena that occur in the plasmas and the limited number of actuators available require the implementation of sophisticated control systems to achieve adequate quality of plasma confinement. The range of requirements is wide and the relevant timescales can range from a few to hundreds of milliseconds. In addition, the high degree of coupling between control parameters increases the level of complexity that the control systems have to address.

It is now well known that in order to reach plasma burning conditions it is needed to operate with high elongated unstable plasmas. These plasmas can be hardly stabilized using Poloidal Field coils placed outside the tokamak vessel, due to the shielding effects of the conducting structures. Therefore for the ITER tokamak it has been proposed to install in-vessel coils for the plasma vertical stabilization. In this paper we design a plasma vertical controller which makes use of both in-vessel and ex-vessel coils for the stabilization of ITER plasmas in various operating conditions. Besides guaranteeing robust stabilization of ITER plasmas, the controller also takes into account constraints on the current flowing in the in-vessel coils; such constraints are required to operate these coils safely. Furthermore the proposed controller is designed as a static-output feedback, where only three gains have to be specified. A sufficient condition is provided to design the controller by solving a Bilinear Matrix Inequalities feasibility problem. The effectiveness of the proposed approach is shown in simulation on a number of ITER plasma configurations.

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.

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.

The plasma in a tokamak is magnetically confined through electromagnetic fields generated by a set of poloidal field coils, with feedforward nominal voltages or currents based on an approximate plasma model. Therefore, feedback control is necessary. A good magnetic control is essential for achieving good performance in present devices and expected to be of paramount importance in future tokamaks with a burning plasma due to the more stringent requirements in terms of plasma-wall clearance, stability margins, limited number of tolerable disruptions and so on. Magnetic fields are able to control plasma current, position and shape as well as three dimensional instabilities on the resistive time scales (resistive wall modes). They also have an influence on the control of the plasma current density profile. This paper focuses on the axisymmetric control of plasma current, position and shape, illustrating modelling assumptions and techniques, controller design procedures, assessment of controller performance and magnetic diagnostic requirements.

This paper deals with the off-line optimization of poloidal field coil currents and voltages to obtain desired plasma scenarios including ramp up, flat top, and ramp down phases. The open loop control problem is solved via a quadratic optimization technique which requires the solution of a certain number of nonlinear plasma equilibrium problems. An alternative approach is based on the use of linearized models. This approach may fail during shape transitions where the effect of control on shape presents significant variations (e.g. limiter-diverted transition or plasma wobbling across a double null equilibrium configuration). To face these situations a novel methodology based on the SVD (singular value decomposition) is proposed, which also opens the way to a linear closed loop control of plasma shape during limiter-diverted configuration transitions. The proposed methodologies are applied to ITER scenarios. This paper appears in: Decision and Control, 2006 Publication Date: 13-15 Dec. 2006 On page(s): 2220-2225 Location: ISBN: INSPEC Accession Number: 9419788 Digital Object Identifier: 10.1109/CDC.2006.377412 Posted online: 2007-05-07 11:27:07.0

The stability robustness problem for a class of nominally linear systems subject to time-varying parameter uncertainty is dealt with. The allowable range of parameter variation is assumed to be a hyperrectangle in parameter space. The proposed robustness test, which is based on a Lyapunov approach, requires a finite number of positivity tests of a quadratic form in correspondence to some salient points in parameter space. >

A flight control system which is capable of reproducing a microgravity environment within a commercial aircraft by forcing it to follow the dynamic behavior of a mass launched in the vacuum is studied. The control problem is formulated as a variational problem and an efficient numerical procedure for the online solution of the problem is given. The solution technique is based on the so-called receding horizon technique. The performance of the proposed control system is illustrated by numerical simulations. >

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.

This paper is focused on the plasma shape and current control in a large, next generation tokamak characterized by long-duration plasma discharges. The new features of these fusion devices are described and their impact on the feedback control requirements are investigated. Many of the controller design issues are addressed in the H/sub /spl infin////spl mu/ synthesis framework, including the model reduction, the power constraints, the voltage limits and plant uncertainties. The application to the case of the International Thermonuclear Experimental Reactor (ITER) shows the effectiveness of this approach for the design of plasma controllers.

In this paper the mathematical model of a modern arc heated Plasma Wind Tunnel used to test parts of space vehicles during the re-entry phase is presented. Due to the complexity of the physical processes arising during the operating of the plant, a number of simplification is proposed to obtain a control design oriented fast and efficient simulation software. Such a software was successfully used during the design phase to become confident with the dynamic behavior of the facility and to achieve a process-control integrated design. During the operating life of the wind tunnel it will be used to support the test engineers in the design of the control laws for the tracking of the desired trajectories in temperature and pressure on the test article. A short description of the test control algorithm which is based on the above model is also provided.

The paper deals with some electromagnetic aspects of interest for their impact on the ITER project. In particular the following topics are illustrated: • feedback control of the distance between plasma separatrix and first wall; • identification of the plasma separatrix from signals detected by magnetic sensors; • ohmic losses in the cold structures, during normal operation scenario and plasma disruptions. The numerical results, which mainly refer to the ITER TAC-8 configuration, are discussed in the paper with the aim of contributing to the still open ITER design options.

The basic features of a procedure for the optimization of the plasma scenario in an air core tokamak are presented. The method takes into account the eddy currents in the passive conducting structures. The problem is reduced to the synthesis of a time-varying magnetic field. The solution of this inverse electromagnetic problem is carried out by means of an optimization procedure based on the receding horizon approach. The paper includes an example of application to the ITER tokamak. >

It is pointed out that plasma position and shape control calls for fast algorithms for plasma boundary identification, i.e. methods capable of determining the shape in a time interval not longer than the sampling time of the controller. The authors present an online identification algorithm which realizes a trade-off between accurate plasma shape reconstruction and computational burden. The proposed identification has been tested with reference to some types of ITER (International Thermonuclear Experimental Reactor) plasmas. >

A design of a dynamic positioning system for a moored lay-vessel in station keeping phase is presented. The design of the position controller is obtained by using a simplified model of the vessel-mooring system dynamics. The position control algorithm results to be made up with a Kalman filter as estimator of the low frequency motion of the vessel, and an optimal state feedback as position regulator; moreover a recursive stochastic algorithm is used for updating the parameters of the high frequency model of the vessel dynamics. The performance of the controller is shown by means of computer simulations. These have been obtained with the designed controller working against a numerical simulator of the complex vessel dynamics explicitly taking into account nonlinear interactions between waves and hull.

The tracking problem for a class of nonlinear systems in the presence of parameter uncertainties is considered. The uncertainties are assumed bounded and their bounds ‘a priori’ known. A feedback control based on a bank of decentralized generalized PD (Proportional + Derivative) controllers is proposed; it is shown that it guarantees ultimate boundedness of the tracking error between plant output and reference trajectory. The stability analysis of the tracking error equation is carried out using some results of VSS (Variable Structure Systems) theory. Moreover the design procedure permits one to arbitrarily prescribe the rate of convergence and the size of the set of ultimate boundedness of the tracking error.

In this paper the attitude control problem for the small launch vehicle VEGA is discussed. In particular, a direct optimisation method is applied to the determination of the parameters of a prescribed mission trajectory, in order to compute a reference pitch attitude angle time history. The attitude and engine-thrusters control system, which allows to track the obtained reference attitude, are designed by using a fuzzy logic approach. The effectiveness of the proposed method is illustrated via simulation performed into the MATLAB/spl trade//SIMULINK/spl trade/ environment.

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.

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.

The need to achieve increasingly better performance in present and future tokamak devices has made plasma control increasingly important in tokamak engineering. When high performance and robustness...

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 control of present day tokamaks usually relies upon primitive modelling and TCV is used to illustrate this. A counter example is provided by the successful implementation of high order SISO controllers on COMPASS-D. Suitable models of tokamaks are required to exploit the potential of modern control techniques. A physics based MIMO model of TCV is presented and validated with experimental closed loop responses. A system identified open loop model is also presented. An enhanced controller based on these models is designed and the performance improvements discussed.

A procedure is proposed to synthesize a feedback controller for the vertical position stabilization of the plasma in an elongated tokamak. The procedure is based on a computer aided design method allowing satisfactory compromises between power requirements and the dynamic behaviour of the overall control system; such an approach permits the takinginto account of the physical constraints on the feasibility of the power amplifiers.

A linear model for feedback control of the plasma position and shape in the International Thermonuclear Experimental Reactor (ITER) is discussed. A model of the poloidal field (PF) system and of the disturbances is first derived. The main task of the control system is to avoid any contact of the hot plasma with the wall during the long duration of the burn phase. For this purpose, the control variables are specified as six gaps between the plasma separatrix and the first wall, including divertor channels. The structure model includes PF coils, vacuum vessel, first wall, backplate, and divertor fins, and it refers to the TAC-4 outline design ITER geometry. A multivariable controller is designed using the optimal linear quadratic approach. The simulation of the closed-loop system shows how the plasma shape is recovered: Step gap variations of 15 cm and poloidal beta drops of 0.2 are considered as disturbances. The performance parameters are voltages and currents in the PF coils and gap recovery time; voltage saturation of the actuators is also taken into account.

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.