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This paper presents the main lessons that can be learned from two decades of Plasma Control development experience. It will present the general architectural choices and will provide examples from key systems, and, more importantly, will highlight the plasma operation requirements that have driven the developments. The aim is to present a meaningful set of functional and non- functional requirements that were derived from the JET experience and to discuss their potential applicability to future experimental devices, ITER in particular. ©2009 IEEE.

In this paper we propose a model-free approach to stabilize vertical unstable plasmas in tokamak devices. The approach relies on the extremum seeking algorithm and on averaging theory to bring to zero the motion of the plasma along the unstable mode, hence achieving the stabilization of its vertical dynamic. The effectiveness of the proposed approach is shown by means of simulations. To this aim, the proposed model-free vertical stabilization system is included in the magnetic control system of the ITER tokamak to consider relevant operational scenarios.

The construction and operation of the JT-60SA tokamak is the main project currently carried out jointly by Japan and the European Union under the Broader Approach agreement. Within the Integrated Project Team, Japanese and European scientists are developing and testing a number of tools to support preliminary studies and future operations of JT-60SA. Within this collaborative framework, European scientists are using a set of assessed modeling tools to design and validate possible solutions for the plasma magnetic control system of JT-60SA. This paper introduces these tools and describes a possible control architecture to be used on the JT-60SA tokamak. The effectiveness of the proposed architecture is shown by means of numerical simulations.

Tokamaks are experimental reactors that currently represent the most promising approach to produce electricity by means of the nuclear fusion reactions. In a tokamak a fully ionized gas, called plasma, in which the nuclear reaction occurs, is confined by means of strong magnetic fields. Its performance, physics knowledge and operation safety are significantly affected by several plasma parameters that are controlled by the so called magnetic control system. However, these parameters cannot be directly measured, but they are estimated by the so called magnetic diagnostic, by combining the information from several sensors. More precisely, different combinations of measurements, which may imply also a different quality of the estimation, can be used to infer the same set of plasma parameters. Each acquisition unit should acquire a set of measurements that allows to reconstruct all the plasma parameters that are needed to operate the machine. In this context, it is fundamental to determine an effective assignment of the sensor measurements to the acquisition units, satisfying capacity constraints and, at the same time, maximizing the overall reliability of the magnetic diagnostic system against possible failures of its components. This problem has been tackled in literature as an original variant of the generalized assignment problem. In this work, we provide a description of the problem itself, as well as a presentation of the related formulations. These formulations are then experienced on real-world test cases in order to show their applicability. Such test cases have been derived from the requirements defined for the JET tokamak, which is the world’s largest tokamak currently in operation. We conclude with a discussion on future research perspectives.

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 this study, the problem of input–output finite-time stabilisation of linear time-varying systems is dealt with. The classical definition of input-output finite-time stability (IO-FTS) is extended to that one of structured IO-FTS, which allows to incorporate, in the definition of the stabilisation problem, some amplitude constraints on the control input variables. A sufficient condition and a necessary and sufficient condition for structured IO finite-time stabilisation are provided in the case of L ∞ and L 2 inputs, respectively. Such conditions require the existence of a solution to a certain differential linear matrix inequality. The theory is applied to design the active suspension control system for a two-degree-of-freedom quarter-car model.

In this paper we discuss the problem of plasma vertical stabilization at the EAST tokamak. By exploiting a plasma/circuit linearized model, we show that the plant cannot be strongly stabilized by using the in-vessel coils and a single-input-single-output controller that feeds back only the plasma vertical speed (z)over dot(p) (i.e. without integral action on (z)over dot(p)). Moreover, a stable multi-input-single-output controller that stabilizes the plant without the need of feeding back the plasma vertical position is presented. The proposed solution permits to achieve stabilization of the EAST plant without coupling the vertical stabilization system with the plasma shape controller. Such decoupling is a key requirement to enable advanced design approaches for plasma 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.

The Plasma Control Upgrade (PCU) enhancement project was setup with the aim of effectively increasing the JET operating space by both improving the Vertical Stabilization System Controller and by procuring a more powerful Radial Field Amplifier. The hardware and software technological advancements that have been used allowed to implement a 200 channel, 20 kHz control system with less than 2 μ s jitter and over 2 G o p s of processing capability. The software design, driven by an object oriented analysis, enabled a good compromise between real-time performances and reliability–maintainability objectives. The first operational experience of the system will be presented.

A novel approach to fault diagnosis of discrete event systems is presented in this paper. The standard approach is based on the offline computation of the set of fault events that may have occurred at each reachable state, providing a fast online diagnosis at a price of excessive memory requirements. A different approach is here adopted, which is based on the online computation of the set of possible fault events required to explain the last observed event. This is efficiently achieved by modelling the plant by Petri nets, since their mathematical representation permits to formulate the fault diagnosis problems in terms of mathematical programming, which is a standard tool. Moreover, the graphical representation of the net allows the diagnoser agent to compute off-line reduced portions of the net in order to improve the efficiency of the online computation, without a big increase in terms of memory requirement.

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.

We describe a software system known as the Plasma Control System Simulation Platform (PCSSP) that is being constructed to support development of the ITER plasma control system. When mature, PCSSP will provide support for industry standard practices such as model-based controller design, simulation testing of controllers, auto-generation of controller code, hardware-and software-in-The-loop testing, use of source code management tools, and open-source methods in development of the software and control algorithms for the ITER Plasma Control System (PCS). It will also contribute to fusion-specific objectives such as validation of ITER pulse schedules prior to their use in experimental operation. We also describe a more mature but less sophisticated software suite known as TokSys, which was developed at General Atomics to support plasma control development and is expected to eventually merge with PCSSP. Plasma control development activities that use or have used these software systems are also described. We will discuss the ongoing ITER PCS control algorithm development for initial plasma operation, which takes advantage of the existing PCSSP infrastructure and library of modules. Plasma control development and initial plasma operation of the EAST and KSTAR tokamaks relied heavily on the TokSys suite, which is now in routine use at EAST, KSTAR, DIII-D, and NSTX-U. © 2015 IEEE.

Simulink is a graphical data flow programming tool for modeling and simulating dynamic systems. A component of Simulink, called Simulink Coder, generates C code from Simulink diagrams. MARTe is a framework for the implementation of real-time systems, currently in use in several fusion experiments. MDSplus is a framework widely used in the fusion community for the management of data. The three systems provide a solution to different facets of the same process, that is, real-time plasma control development. Simulink diagrams will describe the algorithms used in control, which will be implemented as MARTe GAMs and which will use parameters read from and produce results written to MDSplus pulse files. The three systems have been integrated in order to provide a tool suitable to speed up the development of real-time control applications. In particular, it will be shown how from a Simulink diagram describing a given algorithm to be used in a control system, it is possible to generate in an automated way the corresponding MARTe and MDSplus components that can be assembled to implement the target system.

The oPtImal Measurement Probes Allocation (PIMPA) tool has been recently proposed in [8] to maximize the reliability of a tokamak diagnostic system against the failure of one or more of the processing nodes. PIMPA is based on the solution of Integer Linear Programming (ILP) models, and it minimizes the effect of the failure of a data acquisition component. In this paper we focus the attention on the physical parameters needed by the axisymmetric magnetic control system of the ITER tokamak (i.e. the plasma current, position and shape control system). First we introduce the improved ILP model which is then exploited to perform the optimal allocation of the probes needed to reconstruct the desired control variables. The improved model is also used to design some of the parameters of the data acquisition hardware (for example, the number of chassis needed to guarantee a given level of reliability to the system).

Tokamak control systems have to deal with different kinds of instabilities related to the presence of a resistive wall that surrounds the plasma. These instabilities are known as Resistive Wall Modes and are both axisymmetric and non-axisymmetric; they can occur during normal operation of the tokamak, and therefore suitable feedback controllers need to be designed and implemented. In this paper we propose a control architecture able to deal with the two main instabilities: the axisymmetric vertical instability and the non-axisymmetric kink instability. With reference to the case of the ITER tokamak, we design a controller consisting of two separate loops, one for the vertical stabilization and the other one for the stabilization of the kink instability. The two loops are designed in such a way to minimize the control effort and the interaction between them. The effectiveness of the approach is shown in simulation on an appropriate ITER configuration.

The plasma control system is a key instrument for successfully investigating the physics of burning plasma at ITER. It has the task to execute an experimental plan, known as pulse schedule, in the presence of complex relationships between plasma parameters like temperature, pressure, confinement and shape. The biggest challenge in the design of the control system is to find an adequate breakdown of this task in a hierarchy of feedback control functions. But it is also important to foresee structures that allow handling unplanned exceptional situations to protect the machine. Also the management of the limited number of actuator systems for multiple targets is an aspect with a strong impact on system architecture. Finally, the control system must be flexible and reconfigurable to cover the manifold facets of plasma behaviour and investigation goals. In order to prepare the development of a control system for ITER plasma operation, a conceptual design has been proposed by a group of worldwide experts and reviewed by an ITER panel in 2012. In this paper we describe the fundamental principles of the proposed control system architecture and how they were derived from a systematic collection and analysis of use cases and requirements. The experience and best practices from many fusion devices and research laboratories, augmented by the envisaged ITER specific tasks, build the foundation of this collection. In the next step control functions were distilled from this input. An analysis of the relationships between the functions allowed sequential and parallel structures, alternate branches and conflicting requirements to be identified. Finally, a concept of selectable control layers consisting of nested “compact controllers” was synthesised. Each control layer represents a cascaded scheme from high-level to elementary controllers and implements a control hierarchy. The compact controllers are used to resolve conflicts when several control functions would use the same command signals as their outputs. They consist of a collection of potentially conflicting control functions from which one at a time is exclusively activated by a mode selector signal. It can be shown that this architectural design is capable of implementing all of the presently known functional control requirements. Furthermore, this design takes already into account that the result of future experiments at ITER will create additional requirements on the functions or performance of ITER plasma control.

Several alternative plasma control schemes using the same plasma current and shape controller (SC) with different plasma vertical stabilisation (VS) controllers are explored, assessing their efficiency in the suppression of plasma shape transients after vertical displacement event (VDE) disturbances and their robustness to changes of the local dynamics. We attempt to decrease the settling time and the overshoot peaks while maintaining robustness to changes of local dynamics, by speeding up the controller response to disturbances. For VS, static output feedback (SOF) and linear-quadratic-Gaussian (LQG) control are considered. Both SOF and LQG are further augmented with an additional intermediate-level control loop that attempts to bring the system back to the origin after a VDE event faster than the SC normally does.

Model-driven design is recently gaining a wide spreading in different fields, such as design of mechatronics and embedded systems. Approaches based on either UML or SysML permit to efficiently manage the design of complex systems in different areas. In this paper a SysML modeling approach is proposed for the design of real-time applications based on the MARTe framework. The MARTe framework has been recently adopted for the development of real-time systems in several European experimental fusion reactors. The proposed approach allows to achieve a better standardization of the development cycle and a better documentation of the developed systems. Furthermore, by using model-2-text tools it is possible to automatically generate part of the real-time executable code and the application configuration file. The proposed approach has been applied for the modeling of the control system for the Frascati Tokamak Upgrade. © 2005-2012 IEEE.

Tokamak control systems have to deal with different kinds of instabilities related to the presence of a resistive wall that surrounds the plasma. These instabilities are known as Resistive Wall Modes and are both axisymmetric and non-axisymmetric; they can occur during normal operation of the tokamak, and therefore suitable feedback controllers need to be designed and implemented. In this paper we propose a control architecture able to deal with the two main instabilities: the axisymmetric vertical instability and the non-axisymmetric kink instability. With reference to the case of the international thermonuclear experimental reactor ITER, we design a controller consisting of two separate loops, one for the vertical stabilization and the other one for the stabilization of the kink instability. The two loops are designed in such a way to minimize the control effort and the interaction between them. The effectiveness of the approach is shown in simulation on an appropriate ITER configuration.

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.

Validation by means of simulation is a crucial step when developing real-time control systems. Modeling and simulation are an essential tool since the early design phase, when the control algorithms are designed and tested. This phase is commonly carried out in off-line environments such as Matlab® and Simulink®. A MARTe-based simulator has been recently developed to validate the new JET Vertical Stabilization (VS) system. MARTe is the multi-thread framework used at JET to deploy hard real-time control systems. This paper presents the software architecture of the MARTe-based simulator and it shows how this tool has been effectively used to evaluate the effects of Edge Localized Modes (ELMs) on the VS system. By using the simulator it is possible to analyze different plasma configurations, extrapolating the limit of the new vertical amplifier in terms of the energy of the largest rejectable ELM.

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 Multithreaded Application Real-Time executor (MARTe) is a data driven framework environment for the development and deployment of real-time control algorithms. The main ideas which led to the present version of the framework were to standardise the development of real-time control systems, while providing a set of strictly bounded standard interfaces to the outside world and also accommodating a collection of facilities which promote the speed and ease of development, commissioning and deployment of such systems. At the core of every MARTe based application, is a set of independent inter-communicating software blocks, named Generic Application Modules (GAM), orchestrated by a real-time scheduler. The platform independence of its core library provides MARTe the necessary robustness and flexibility for conveniently testing applications in different environments including non-real-time operating systems. MARTe is already being used in several machines, each with its own peculiarities regarding hardware interfacing, supervisory control configuration, operating system and target control application. This paper presents and compares the most recent results of systems using MARTe: the JET Vertical Stabilisation system, which uses the Real Time Application Interface (RTAI) operating system on Intel multi-core processors; the COMPASS plasma control system, driven by Linux RT also on Intel multi-core processors; ISTTOK real-time tomography equilibrium reconstruction which shares the same support configuration of COMPASS; JET error field correction coils based on VME, PowerPC and VxWorks; FTU LH reflected power system running on VME, Intel with RTAI.

EPICS is used worldwide mostly for controlling accelerators and large experimental physics facilities. Although EPICS is well fit for the design and development of automation systems, which are typically VME or PLC-based systems, and for soft real-time systems, it may present several drawbacks when used to develop hard real-time systems/applications especially when general purpose operating systems as plain Linux are chosen. This is in particular true in fusion research devices typically employing several hard real-time systems, such as the magnetic control systems, that may require strict determinism, and high performance in terms of jitter and latency. Serious deterioration of important plasma parameters may happen otherwise, possibly leading to an abrupt termination of the plasma discharge. The MARTe framework has been recently developed to fulfill the demanding requirements for such real-time systems that are alike to run on general purpose operating systems, possibly integrated with the low-latency real-time preemption patches. MARTe has been adopted to develop a number of real-time systems in different Tokamaks. In this paper, we first summarize differences and similarities between EPICS IOC and MARTe. Then we report on a set of performance measurements executed on an x86 64 bit multicore machine running Linux with an IO control algorithm implemented in an EPICS IOC and in MARTe.

When developing critical control system it is appropriate to resort to modeling, simulation and rapid prototyping tools since the early design phases. A design approach based on the rapid prototyping of the plant model can be effectively adopted when non-conventional or new control systems have to be developed, or when the required performance are very demanding. Furthermore, such an approach is the only one viable when the plant is not available yet and/or the testing on-site is very risky, which is usually the case for fusion experimental reactors (tokamaks). This paper presents a flexible architecture for the model-based design of control systems, which permits to perform rapid prototyping and hardware-in-the-loop validation. The proposed architecture has been recently adopted for the rapid prototyping of the Central Safety System for Nuclear Risk of the ITER tokamak.

The need to improve the performance of modern tokamak operations has led to a further development of the plasma shape and position control systems. In particular, extremely elongated plasmas, with high vertical-instability growth rate, are envisaged to reach the required performance for ignition. This request for better performance from the experimentalists' side has motivated the development of the new vertical-stabilization (VS) system at the JET tokamak, which has been proposed within the Plasma Control Upgrade project. The main aim of the project is to enhance the capabilities of the plasma vertical position control system in order to operate with very highly elongated plasmas in the presence of large perturbations. This paper focuses on the new software architecture of the VS system, which relies on a highly configurable real-time framework. Due to its flexibility, the new VS system executes different control algorithms, and it schedules the one which maximizes the performance in each plasma phase.

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

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

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

UniSim is a graphical tool which can be used to avoid the choice of a specific commercial platform when teaching the IEC 61131-3 standard for programmable logic controllers. UniSim allows to develop automation software which is fully compliant with the standard, and to validate it via simulation. It can also be used to fast prototyping automation systems using a desktop equipped with low-cost commercial I/O boards. This prototype can be interfaced with a plant simulator to perform hardware-in-the-loop validation of the control system.

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.

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.