Search

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.

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.

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.

The Ignitor Program maintains the objective of approaching D-T ignition conditions by incorporating systematical advances made with relevant high field magnet technology and with experiments on high density well confined plasmas in the present machine design. An additional objective is that of charting the development of the high field line of experiments that goes from the Alcator machine to the ignitor device. The rationale for this class of experiments, aimed at producing poloidal fields with the highest possible values (compatible with proven safety factors of known plasma instabilities) is given. On the basis of the favourable properties of high density plasmas produced systematically by this line of machines, the envisioned future for the line, based on novel high field superconducting magnets, includes the possibility of investigating more advanced fusion burn conditions than those of the D-T plasmas for which Ignitor is designed. Considering that a detailed machine design has been carried out (Coppi et al 2013 Nucl. Fusion 53 104013), the advances made in different areas of the physics and technology that are relevant to the Ignitor project are reported. These are included within the following sections of the present paper: main components issues, assembly and welding procedures; robotics criteria; non-linear feedback control; simulations with three-dimensional structures and disruption studies; ICRH and dedicated diagnostics systems; anomalous transport processes including self-organization for fusion burning regimes and the zero-dimensional model; tridimensional structures of the thermonuclear instability and control provisions; superconducting components of the present machine; envisioned experiments with high field superconducting magnets. © 2015 IAEA.

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.

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.

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 has been designed and implemented. This paper presents the experimental results achieved at JET during the 2012 campaigns using the XSC. © 2013 Euratom: EURATOM-ENEA-CREATE/ Universita' di Napoli Federico II; European Atomic Energy Community.

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.

Plasma shape identification is a crucial problem that needs to be tackled for controlling the plasma configuration in a tokamak machine. The usual way to cope with this issue is to solve an inverse problem based on the solution of the Grad-Shafranov equilibrium equation; the inputs to this algorithm are a certain number of magnetic field and flux measurements in points located around the plasma. The number and the location of the magnetic sensors providing the measurements affect the accuracy of the identification algorithm. In this paper we deal with 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 we consider the ITER inner vessel magnetic diagnostic system and show that we can partition the available measurements 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. This allow us to evaluate the redundancy level of the ITER inner vessel diagnostic system and to outline a suitable fault management system.

A current limit avoidance (CLA) system has been proposed to avoid current saturation in the poloidal field (PF) coils of the Joint European Torus tokamak when the eXtreme Shape Controller is used to control the plasma shape. CLA uses the redundancy of the PF coil system to automatically obtain almost the same plasma shape with a different combination of control currents. This paper describes the set of graphic tools that has been recently developed to aid the design of the CLA parameters by nonexpert users.

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.

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.

Fusion Advanced Studies Torus (FAST) conceptual study has been proposed as possible European ITER Satellite. This facility is aimed at exploring and preparing ITER operation scenarios as well as helping DEMO design and R&D. This paper presents the plasma position and shape control studies, focusing on the machine flexibility in terms of both operational space and physics that can be investigated. The optimization of a copper shell position inside the vacuum vessel provides a slowing down of the growth rate of the vertical instability around 13s−1. A sensitivity analysis of each poloidal field (PF) circuit and their effects on plasma shape is carried out. A SVD based PID control scheme has been used for controlling the plasma current, shape and position.

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.

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 propose a solution to the input saturation avoidance problem in the JET tokamak shape control by illustrating its capabilities to enforce coil currents selections that tend to move away from the saturation limits within the allowable degrees of freedom. The proposed solution revisits the dynamic allocation scheme first proposed in [10] for input redundant plants and generalized in [6] and [9] for not input redundant ones. In particular we propose a different scheme for the input allocation, more suitable within a set-point regulation setting, and prove its convergence properties. We also present simulation results to assess the steady-state and transient performances obtained with the novel scheme, as compared to the previous approaches.

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.

FAST is a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at studying, under burning plasma relevant conditions, fast particle (FP) physics, plasma operations and plasma wall interaction in an integrated way. FAST has the capability to approach all the ITER scenarios significantly closer than the present day experiments using deuterium plasmas. The necessity of achieving ITER relevant performance with a moderate cost has led to conceiving a compact tokamak (R = 1.82 m, a = 0.64 m) with high toroidal field (B(T) up to 8.5 T) and plasma current (I(p) up to 8 MA). In order to study FP behaviours under conditions similar to those of ITER, the project has been provided with a dominant ion cyclotron resonance heating system (ICRH; 30 MW on the plasma). Moreover, the experiment foresees the use of 6 MW of lower hybrid (LHCD), essentially for plasma control and for non-inductive current drive, and of electron cyclotron resonance heating (ECRH, 4 MW) for localized electron heating and plasma control. The ports have been designed to accommodate up to 10 MW of negative neutral beams (NNBI) in the energy range 0.5-1 MeV. The total power input will be in the 30-40 MW range under different plasma scenarios with a wall power load comparable to that of ITER (P/R similar to 22 MW m(-1)). All the ITER scenarios will be studied: from the reference H mode, with plasma edge and ELMs characteristics similar to the ITER ones (Q up to approximate to 1.5), to a full current drive scenario, lasting around 170s. The first wall (FW) as well as the divertor plates will be of tungsten in order to ensure reactor relevant operation regimes. The divertor itself is designed to be completely removable by remote handling. This will allow us to study (in view of DEMO) the behaviour of innovative divertor concepts, such as those based on liquid lithium. FAST is capable of operating with very long pulses, up to 170s, despite being a copper machine. The magnets initial operation temperature is 30 K, with cooling provided by helium gas. The in vessel components, namely FW and divertor, are actively cooled by pressurized water above 80 degrees C. The same water is also used to bake the vacuum vessel. FAST is equipped with ferromagnetic inserts to keep the toroidal field magnet ripple down to 0.3%.

The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300–500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode and a late H-mode onset. Static equilibrium analyses for this scenario, which determine PF coil currents to produce a given plasma configuration, indicate that the original PF coil limitations do not allow low li(

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.

In the recent years Plasma Control has always increased his importance in any advanced experiment. It is now clear that ITER will not be able to operate without a quite advanced and sophisticated control apparatus. Necessarily this system will have to integrate several different aspects of the Plasma behavior. One of the most important parts of a closed loop control system is the quality of the measurement of the plasma parameters that should be controlled. Eventually, this aspect involves sophisticated and complex diagnostic apparatus. This paper presents an overview of the present status, and further studies and developments needed, in the next future, for the design and realization of an integrated plasma control system aimed at both stabilizing the plasma non-axisymmetric instabilities and controlling the most important internal plasma parameters. In particular the Edge Localized Modes (ELMs), the Neo-Classical Tearing Modes (NTM), the Resistive Wall Mode (RWM) and the Plasma Profiles control system necessities will be shortly illustrated. Keywords: Plasma, Control, Diagnostics PACS: 52.55.Fa, 52.70.Ds, 07.05.Dz.

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.

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 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

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 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.

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.

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.

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. >

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.

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.

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.

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.