Development of SPARC plasma disruption database through electromagnetic predictive MAXFEA Modelling

The evaluation of electromagnetic (EM) behaviour of tokamak components under disruption loads and its correlation with vertical instabilities is critical and characterizing the design considerations. This is particularly crucial for high-field and burning plasma experiments such as the SPARC tokamak, the compact prototype currently under construction by Commonwealth Fusion Systems (CFS). The combination of high toroidal magnetic field (12.2 T) and high plasma current (8.7 MA) makes SPARC less tolerant to disruptions than currently operating research devices. In this context, recent efforts have been made to characterize the disruption loads in SPARC by conducting a series of analyses to estimate the EM response of the SPARC vacuum vessel (VV) during a plasma vertical displacement event (VDE). In this paper, the aforementioned analyses are used as starting point for a comprehensive characterization of VDE scenarios. These analyses are utilized to create a database of predictive disruption simulations, serving as synthetic experiments to generate and collect pertinent data for upcoming research studies. In this context, the paper discusses a preliminary description of an innovative VV synthetic displacements diagnostic, designed starting from disruption database, presented here, and by exploiting the MAXFEA-ANSYS combined use methodology as further step. The disruptive plasma scenarios database will be functional in estimating the displacement and strain fields of SPARC components using a limited set of VV indirect displacement measurements and integral plasma observations, such as plasma current and effective plasma position.

A data-driven Vertical Stabilization system for the ITER tokamak based on Dynamic Mode Decomposition

In this article, we propose an online method to estimate an approximate linear model of a vertically unstable tokamak plasma to consequently adapt the parameters of a Vertical Stabilization controller. The identification procedure is based on the Dynamic Mode Decomposition with control approach, while the tuning procedure takes advantage of linear control theory to impose the desired crossing frequency and gain margins. The proposed technique is aimed at controlling ITER elongated plasmas using the VS3 stabilization coils, located inside the vessel. The effectiveness of the method is proven by means of numerical simulations carried out with the CREATE-NL+ free boundary evolutionary code, also considering the presence of realistic measurement noise levels.

Assessing the Finite-Time Stability of Nonlinear Systems by means of Physics-Informed Neural Networks

In this paper, the problem of assessing the Finite-Time Stability (FTS) property for general nonlinear systems is considered. First, some necessary and sufficient conditions that guarantee the FTS of general nonlinear systems are provided; such conditions are expressed in terms of the existence of a suitable Lyapunov-like function. Connections of the main theoretical result of given in this article with the typical conditions based on Linear Matrix Inequalities (LMI) that are used for Linear Time-Varying (LTV) systems are discussed. An extension to the case of discrete time systems is also provided. Then, we propose a method to verify the obtained conditions for a very broad class of nonlinear systems. The proposed technique leverages the capability of neural networks to serve as universal function approximators to obtain the Lyapunov-like function. The network training data are generated by enforcing the conditions defining such function in a (large) set of collocation points, as in the case of Physics-Informed Neural Networks. To illustrate the effectiveness of the proposed approach, some numerical examples are proposed and discussed. The technique proposed in this paper allows to obtain the required Lyapunov-like function in closed form. This has the twofold advantage of a) providing a practical way to verify the considered FTS property for a very general class of systems, with an unprecedented flexibility in the FTS context, and b) paving the way to control applications based on Lyapunov methods in the framework of Finite-Time Stability and Control.

Implementation of a high-speed multichannel data acquisition system for magnetic diagnostics and plasma centroid position control in ISTTOK

In tokamak and other fusion devices, magnetic control is the main tool that allows to regulate the plasma current, position and shape; it is in charge of actuating the desired plasma current waveform, steering the plasma position to a given set point and maintain the plasma shape close to a prescribed plasma equilibrium. This work describes the application of several physics concepts and computational tools in order to obtain a novel optimal controller for the plasma centroid position, which has been implemented and tested in the real-time plasma control system at the ISTTOK tokamak. A key point for the development of the new control system was the installation of a recently upgraded hardware, that numerically integrates in real-time the magnetic probes signals.

A Deep Reinforcement Learning approach for Vertical Stabilization of tokamak plasmas

Reinforcement Learning has emerged as a promising approach to implement efficient data-driven controllers for a variety of applications. In this paper, a Deep Deterministic Policy Gradient (DDPG) algorithm is used to train a Vertical Stabilization agent, to be considered as a possible alternative to the model-based solutions usually adopted in existing machines. The agent is trained and validated considering the ITER tokamak magnetic control as case study environment. The tuning of the DDPG algorithm's hyper-parameters is motivated through a sensitivity analysis.

Vertical stabilization of tokamak plasmas via extremum seeking

In this paper we propose a vertical stabilization (VS) control system for tokamak plasmas based on the extremum seeking (ES) algorithm. The gist of the proposed strategy is to inject an oscillating term in the control action and exploit a modified ES algorithm in order to bring to zero the average motion of the plasma along the unstable mode. In this way, the stabilization of the unstable vertical dynamic of the plasma is achieved. The approach is validated by means of both linear and nonlinear simulations of the overall ITER tokamak magnetic control system, with the aim of demonstrating robust operation throughout the flat-top phase of a discharge and the capability of reacting to a variety of disturbances.

Implementation of a Kalman filter-based eddy current estimator for the P-EFIT magnetic equilibrium reconstruction code

This article discusses the integration of a Kalman filter in the P-EFIT equilibrium reconstruction code, with the aim of estimating the currents induced in the passive structures of a tokamak. The filter is based on a vacuum electromagnetic model of the reactor, and takes advantage of an estimate of the effect of the plasma on the magnetics, provided by the equilibrium reconstruction algorithm. On the other hand, the observer is integrated into the equilibrium reconstruction, which exploits the eddy currents estimates provided by the Kalman filter to refine the obtained solution. To analyze the interplay of the reconstruction code and the proposed observer, the ITER tokamak is considered as a case-study, and the algorithm is tested on a variety of plasma conditions, selected in such a way to maximize the relevance of an accurate knowledge of the passive currents. The code performance is evaluated in terms of convergence metrics, eddy currents estimation accuracy and reconstruction of plasma-related quantities such as plasma-wall gaps, plasma current and plasma profile parameters.

Plasma physics and control studies planned in JT-60SA for ITER and DEMO operations and risk mitigation

A large superconducting machine, JT-60SA has been constructed to provide major contributions to the ITER program and DEMO design. For the success of the ITER project and fusion reactor, understanding and development of plasma controllability in ITER and DEMO relevant higher beta regimes are essential. JT-60SA has focused the program on the plasma controllability for scenario development and risk mitigation in ITER as well as on investigating DEMO relevant regimes. This paper summarizes the high research priorities and strategy for the JT-60SA project. Recent works on simulation studies to prepare the plasma physics and control experiments are presented, such as plasma breakdown and equilibrium controls, hybrid and steady-state scenario development, and risk mitigation techniques. Contributions of JT-60SA to ITER and DEMO have been clarified through those studies.

A deep deterministic policy gradient learning approach to missile autopilot design

In this paper a Deep Reinforcement Learning algorithm, known as Deep Deterministic Policy Gradient (DDPG), is applied to the problem of designing a missile lateral acceleration control system. To this aim, the autopilot control problem is recast in the Reinforcement Learning framework, where the environment consists of a 2-Degrees-of-Freedom nonlinear model of the missile's longitudinal dynamics, while the agent training procedure is carried out on a linearized version of the model. In particular, we show how to account not only for the stabilization of the longitudinal dynamic, but also for the main performance indexes (settling-time, undershoot, steady-state error, etc.) in the DDPG reward function. The effectiveness of the proposed DDPG-based missile autopilot is assessed through extensive numerical simulations, carried out on both the linearized and the fully nonlinear dynamics by considering different flight conditions and uncertainty in the aerodynamic coefficients, and its performance is compared against two model-based control strategies in order to check the capability of the proposed data-driven approach to achieve prescribed closed-loop response in a completely model-free fashion.

Finite-time stabilization of linear systems with unknown control direction via extremum seeking

In this article, the finite-time stabilization problem is solved for a linear time-varying system with unknown control direction by exploiting a modified version of the classical extremum-seeking algorithm. We propose to use a suitable oscillatory input to modify the system dynamics, at least in an average sense, so as to satisfy a differential linear matrix inequality condition, which in turn guarantees that the system's state remains inside a prescribed time-varying hyperellipsoid in the state space. The finite-time stability (FTS) of the averaged dynamics implies the FTS of the original system, as the distance between the original and the averaged dynamics can be made arbitrarily small by choosing a sufficiently high value of the dithering frequency used by the extremum-seeking algorithm. The main advantage of the proposed approach resides in its capability of dealing with systems with unknown control direction, and/or with a control direction that changes over time. Being FTS a quantitative approach, this article also gives an estimate of the necessary minimum dithering/mixing frequency provided, and the effectiveness of the proposed finite-time stabilization approach is analyzed by means of numerical examples.

Preliminary evaluation of the LIUQE code reconstruction performance for the DTT device

The Divertor Tokamak Test (DTT) facility is a medium-size tokamak whose construction is starting in the ENEA Site, Frascati. Its principal aim is to investigate alternative solutions for the power exhaust in view of future fusion reactors such as ITER and DEMO, and in this view a palette of several possible magnetic configurations is currently at study for this machine. As in all fusion devices, DTT will be equipped since day-0 with a suitable set of magnetic diagnostics, which is currently in the design phase. In this work, we use the FEM models of several magnetic configurations obtained with the CREATE-L/NL codes and a preliminary set of magnetic sensors to evaluate the accuracy achievable by the LIUQE equilibrium reconstruction code over a variety of magnetic configurations.

On the numerical solution of DLMIs

This paper presents a novel approach for the numerical solution of differential linear matrix inequalities. The solutions are searched in the class of piecewise-quadratic functions with symmetric matrix coefficients to be determined. To limit the numbers of unknowns, congruence constraints are considered to guarantee continuity of the solution and of its derivative. In Example section, some control problems involving differential linear matrix inequalities are considered and solved in order to compare the proposed approach with alternative approximation methods adopted in the literature.

GPU-optimized fast plasma equilibrium reconstruction in fine grids for real-time control and data analysis

P-EFIT, a GPU parallel equilibrium reconstruction code, is based on the EFIT framework, but built with the CUDA (Compute Unified Device Architecture) to take advantage of massively parallel Graphical Processing Unit (GPU) cores to significantly accelerate the computation. With the parallelized Grad-Shafranov solver and middle-scale matrix calculation modules, P-EFIT can accurately reproduce the EFIT reconstruction algorithms at a fraction of the EFIT computational time. Integrated into EAST plasma control system, P-EFIT not only provides control signal results but also enhance EAST plasma control capacity with unique designed function modules. Using the synthetic magnetic diagnostic signals from ITER plasma equilibria obtained by the CREATE-L and CREATE-NL codes in standalone and streaming mode, P-EFIT has good enough accuracy and time latency performance in most of the considered cases. P-EFIT achieves full kinetic equilibrium reconstruction algorithms and repeats the EFIT results with the DIII-D internal plasma current and kinetic profile measurements in one-percent cost time of EFIT. All these works suggest that P-EFIT can provide quality magnetic equilibrium reconstruction in real-time, offer full kinetic equilibrium reconstruction with high spatial resolution and high

A reduced basis approach to plasma equilibrium reconstruction in tokamaks

In this paper, the Proper Orthogonal Decomposition approach is exploited to obtain a Reduced Order Model for the plasma magnetic equilibrium problem, described by the Grad-Shafranov equation. The proposed method is applied over a dataset consisting of a few hundred ITER plasma equilibria, and is applied to the problem of plasma equilibrium reconstruction.

Advances in the physics studies for the JT-60SA tokamak exploitation and research plan

JT-60SA, the largest tokamak that will operate before ITER, has been designed and built jointly by Japan and Europe, and is due to start operation in 2020. Its main missions are to support ITER exploitation and to contribute to the demonstration fusion reactor machine and scenario design. Peculiar properties of JT-60SA are its capability to produce long-pulse, high-?, and highly shaped plasmas. The preparation of the JT-60SA Research Plan, plasma scenarios, and exploitation are producing physics results that are not only relevant to future JT-60SA experiments, but often constitute original contributions to plasma physics and fusion research. Results of this kind are presented in this paper, in particular in the areas of fast ion physics, high-beta plasma properties and control, and non-linear edge localised mode stability studies.

Plasma shape control assessment for JT-60SA using the CREATE tools

This paper deals with the plasma shape control problem in JT-60SA. An assessment of the plasma shape control performance is presented, aimed at the definition of an optimal set of gaps to be controlled. Indeed, JT-60SA represents a relevant benchmark to further validate this control approach given the high beta regimes that are envisaged during its operation. Moreover, such regimes represent a challenge from the plasma magnetic control perspective.
The control approach considered for the assessment is based on the eXtreme Shape Controller (XSC), since such an approach permits to control, in a least mean square sense, a number of shape descriptors that is larger than the number of poloidal field coils. Considering that the design of the XSC is model-based, the CREATE linear model for the plasma-circuit response has been used for the design.
In the presented analysis, the capability of tracking different plasma shapes, as well as the one of rejecting disturbances has been considered. The result of this analysis suggests that a set of about 20 gaps equally spaced along the plasma boundary permits to control the shape with a steady-state root-mean square error of less than 1 cm during the flattop of JT-60SA Scenario 2, in the presence of a set of relevant disturbances.

MIMO Shape Control at the EAST tokamak: simulations and experiments

During 2016-2018 experimental campaigns, the plasma magnetic control architecture of the EAST tokamak was revised in order to achieve improved performances, with the final aim of feedback control of alternative divertor configurations (i.e. with multiple X-points). This paper reports on the results obtained with the Multi-Input-Multi-Output (MIMO) plasma shape controller tested during the last experimental campaign, which prompted a considerable improvement of the control performances. Simulation results are also discussed.

Model-based plasma vertical stabilization and position control at EAST

This paper deals with the model-based approach that has been adopted to design the plasma magnetic control at the EAST tokamak. Such a design approach, which is based on a linear model for the response of the plasma and of the surrounding coils, has been successfully applied in 2016 to design an alternative solution for the plasma vertical stabilization, the plasma centroid position control, and the controller of the currents in the poloidal field circuits. After an introduction to the linear model for the plasma/circuits response, the proposed control algorithms are presented together with the preliminary results obtained during the 2016 experimental campaign at EAST. A brief description of the ongoing and future design activities is also given.

Status of the ITER remote experimentation centre

The ITER Remote Experimentation Centre (REC) project (one of the three sub-projects of the International Fusion Energy Research Centre (IFERC)) is progressing under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach (BA) activities in the field of fusion energy research. The objectives of the REC activity are to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, and to develop the remote experiment system and verify the functions required for remote experimentation by using the Satellite Tokamak (JT-60SA) facilities to facilitate the future exploitation of ITER and JT-60SA. The functions of REC will be tested, and the total system will be demonstrated using JT-60SA and existing facilities in the EU, such as JET and WEST.
The hardware of the REC has been prepared in Rokkasho Japan, which has the remote experiment room with a large video wall to show the plasma and operation status, IT equipment and a storage system by the reuse of the Helios supercomputer tape library. A broadband network infrastructure of 10Gbps has been installed connected to SINET5. Using this network system, fast data transfer from ITER to REC was examined in 2016, and the transfer of the data volumes expected for the initial ITER experiments has been demonstrated. A secure remote experimentation system has been developed, using JT-60SA, that has functions for preparing and setting of shot parameters, viewing the status of control data, streaming of the plasma status, data-exchange function of shot events, and monitoring of the facility operation. Remote data analysis techniques, data visualisation software, a documentation management and experiment planning system and numerical simulation codes for the preparation and performance estimation of discharges have also been developed.

ITER-like vertical stabilization system for the east Tokamak

A ITER-like vertical stabilization (VS) algorithm has been successfully deployed and commissioned at EAST. The proposed algorithm decouples the VS from the plasma shape control, while the algorithms previously implemented to stabilize the EAST plasma exhibit a strong coupling with plasma shape control system. As a consequence, the VS algorithms previously implemented at EAST prevent the deployment of advanced multi-input-multi-output (MIMO) plasma shape control schemes, such as the ones proposed in Albanese et al 2016 (Proc. 2016 IEEE Multi-Conf. System Control (Buenos Aires, Argentina) pp 611-6) and Kolemen et al (2015 J. Nucl. Mater. 463 1186). Indeed, such MIMO controllers rely on the decoupling with the VS system.
The proposed ITER-like stabilizes the plasma column (i.e. it controls to zero the plasma vertical speed) on the fastest possible time scale, while leaves the control of the plasma vertical position to the plasma shape controller. Thanks to this frequency separation approach, the plasma shape controller can than be designed starting from the stabilized system, without explicitly taking the VS into account. In this paper we present the implementation details of the adopted solution for the EAST vertical stabilization, together with the results obtained during the 2016 experimental campaign.

Physics and operation oriented activities in preparation of the JT-60SA tokamak exploitation

The JT-60SA tokamak, being built under the Broader Approach agreement jointly by Europe and Japan, is due to start operation in 2020 and is expected to give substantial contributions to both ITER and DEMO scenario optimisation. A broad set of preparation activities for an efficient start of the experiments on JT-60SA is being carried out, involving elaboration of the Research Plan, advanced modelling in various domains, feasibility and conception studies of diagnostics and other sub-systems in connection with the priorities of the scientific programme, development and validation of operation tools. The logic and coherence of this approach, as well as the most significant results of the main activities undertaken are presented and summarised.

Control-oriented tools for the design and validation of the JT-60SA magnetic control system

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.

  Conferences

A Model Predictive Control System for Different Phases of Plasma Confinement in DEMO Tokamak

This paper deals with the use of a model predictive control technique to perform the magnetic confinement of a plasma in the DEMO tokamak. We show how to adapt the proposed strategy for the different phases, and hence different goals, of the plasma discharge and how to take into account the constraints that characterize the normal operations of a nuclear fusion reactor. We validate the performance of the proposed control system by using a nonlinear evolution code describing the plasma in a tokamak.

Plasma magnetic control for DEMO tokamak using MPC

This paper proposes a model predictive controller (MPC) designed to tackle the magnetic control of the DEMO plasma. DEMO is an ambitious EU project aimed at the construction by 2050 of a tokamak demonstrating that energy from nuclear fusion can be conveniently commercially used. Magnetic control is one of the problems to be solved to operate the machine. Using MPC, in this paper it is shown how it is possible to take into account all the various constraints on the input and state variables during the controller design. The effectiveness of the proposed approach is demonstrated by means of simulations, using a validated nonlinear evolution code describing the interactions between the plasma and the surrounding metallic strictures.

Event-driven adaptive Vertical Stabilization in tokamaks based on a bounded Extremum Seeking algorithm

This paper presents an alternative architecture for a plasma Vertical Stabilization system, which is based on Extremum Seeking (ES) algorithm. The proposed solution relies on a bounded version of this algorithm, which is extended with an event-based gain adaption mechanism, in order to improve the overall performance when a switching power supply feeds the control circuit. To show the effectiveness of the proposed architecture, the case study of the ITER tokamak is considered.

A RL-based Vertical Stabilization System for the EAST tokamak

The Vertical Stabilization (VS) is one of the most challenging control problems in tokamak fusion reactors. Existing solutions are usually tailored for the specific device. In this work, a model-free agent trained with the Q-learning algorithm is proposed as a possible alternative solution to the VS problem. The EAST tokamak is then considered as case-study to show the effectiveness of the proposed approach. The robustness of the VS agent is assessed by taking into account different plasma conditions and by including it into simulations of the overall EAST plasma control system.

Stabilizing elongated plasmas using extremum seeking: the ITER tokamak case study

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.

Model-Based MIMO Isoflux Plasma Shape Control at the EAST Tokamak: Experimental Results

This paper reports on the experimental validation of a MIMO plasma shape controller at the EAST tokamak, carried out during the 2019-2020 experimental campaign. In particular the decoupled SISO controller that is routinely used during operations at EAST, was temporarily replaced by a MIMO isoflux controller, whose design is based on the eXtreme Shape Controller approach, which permits to control a larger number of plasma shape parameters. The proposed plasma shape controller has proven to be able of withstanding relevant disturbances, i.e. to the additional power injections required to reach an almost fully non-inductive current driven regime.

Conceptual design of DTT magnetic diagnostics

The Divertor Tokamak Test (DTT) is a new tokamak device whose main mission is to explore innovative divertor concepts for DEMO and test them in heat loads conditions on plasma facing components relevant to a fusion reactor. The device is presently going through a detailed design process, and the tendering process has started for some of the components. Magnetic diagnostics for plasma current and shape control, and vertical position stabilization are essential diagnostics for tokamak operation, and are the first diagnostic components that will need to be ready for installation and commissioning when DTT vacuum vessel and magnets are assembled. The DTT operational conditions for in-vessel and ex-vessel magnetic diagnostics resemble those that will be encountered in ITER (high heat flux, long pulses, large Electromagnetic stresses), except for the 14MeV neutron effects. Furthermore the compact machine assembly implies a very tight space for High Field Side sensors, both in-vessel and ex-vessel, therefore a dedicated thin sensors design is mandatory.
In this work the conceptual design of DTT magnetic diagnostics will be presented, including Mirnov Coils, LTCC coils, saddle loops, flux loops and diamagnetic loops, Hall probes and optic fibre plasma current measurements. The constraints that have been taken into account in choosing the sensors technology will be outlined, motivating the different design choices. Sensors number and sensitivity have been determined with a model-based optimization procedure: the error in the reconstruction of plasma current and current centroid position has been iteratively estimated by varying the placement of a current filament on a grid spanning the vacuum vessel volume. The maximum reconstruction error has been evaluated for different sensors number, position and effective area (NA). The optimal sensors setup will be discussed, in order to keep the error in plasma current less than 1%, and the error on the current centroid position less than 1cm. Finally the difficult task of integrating the sensors design with the machine first wall and vacuum vessel design will be discussed.

Developing steady state ELM-absent H-mode scenarios with advanced divertor configuration in EAST tokamak

2D and 3D modelling of JT-60SA for disruptions and plasma start-up

Verification tests for remote participation at ITER REC

Shape reconstruction and eddy current estimation via Kalman filter at the EAST tokamak

Plasma shape control is a core issue for thermonuclear fusion reactors. In order to achieve high control performances, the magnetic control system of a tokamak device must rely on a fast and accurate shape reconstruction algorithm, which is capable of precisely estimating the poloidal flux map that best fits the available experimental measurements at every control cycle. In this work, a possible solution to this problem is discussed, based on the well known Kalman filtering theory. The CREATE equilibrium codes [1], [2] have been used to generate linearized models of the plasma response, which can be embedded in an optimal state observer in order to achieve a fast and accurate reconstruction of the plasma shape, plus an estimate of the eddy currents induced in the passive structures. This reconstruction is also suitable for a real-time implementation, as it entails only matrix multiplications and a single matrix inversion. As a testbed, the proposed solution has been applied to experimental data coming from the EAST tokamak.

Robust plasma vertical stabilization in tokamak devices via multi-objective optimization

In this paper we present a robust design procedure for plasma vertical stabilization systems in tokamak fusion devices. The proposed approach is based on the solution of a multi-objective optimization problem, whose solution is aimed at obtaining the desired stability margins under different plasma operative scenarios. The effectiveness of the proposed approach is shown by applying it to the ITER-like vertical stabilization system recently tested on the EAST tokamak.

On plasma vertical stabilization at EAST tokamak

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 p (i.e. without integral action on z 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.

Model-based plasma vertical stabilization and position control at EAST

A model-based approach for the design of plasma position and shape control has been recently proposed in [1], aiming at the development of a multi-input-multi-output architecture for integrated plasma shape and flux expansion control at the EAST tokamak. Such an approach exploits the linear models for the response of the plasma and of the surrounding coils derived from the CREATE magnetic equilibrium codes [2, 3]. These linear models have been used for the first time at EAST to design advanced magnetic configuration [4]. During the EAST 2016 experimental campaign the CREATE models have been further exploited to design an alternative approach for plasma vertical stabilization (VS), which is based on the algorithm proposed for ITER [5], and which assures the decoupling between VS and plasma shape control, as required in [1]. Furthermore, the same modeling tools have been used to design the plasma centroid position control, by taking explicitly into account the new VS loop.
In this work we first briefly introduce the adopted model-based design approach, as well as the proposed algorithm for VS and plasma position control. Eventually the results obtained during the 2016 experimental campaign at EAST are discussed. In particular, it is shown that, although the proposed ITER-like VS does not control the vertical position, it is capable to stabilize the plasma column using only the in-vessel coils, without the need of the slow stabilizing action from the superconductive circuits, as routinely done by the EAST PCS [6].

Status of ITER Remote Experimentation Centre

The project of the ITER Remote Experimentation Centre (REC) (one of the three sub-projects of the International Fusion Energy Research Centre (IFERC)) is progressing under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach (BA) activities in the field of fusion energy research. The objectives of the REC activity are to identify the functions and solve the technical issues for the construction of the REC for ITER at Rokkasho, and to develop the remote experiment system and verify the functions required for the remote experiment by using the Satellite Tokamak (JT-60SA) facilities in order to facilitate the future exploitation of the future experiments of ITER and JT-60SA [1, 2]. In the end, the functions of REC will be tested, and the total system is demonstrated by using JT-60SA and existing facilities in the EU, such as JET and WEST.
The hardware of the REC has been prepared in Rokkasho Japan, which has the remote experiment room with a large video wall to show the plasma and operation status, IT equipment and the storage system by the reuse of the supercomputer of Helios. The REC facility will be widely used for the remote collaboration. Also, the broadband network infrastructure of 10Gbps has been produced. Using this network system, the fast data transfer from ITER to REC has been examined in 2016, and the transfer of the full data in the initial ITER experiment has been demonstrated [3]. Secure remote experiment system has been developed, using JT-60SA, that has functions for preparing and setting of shot parameters, viewing the status of control data, streaming of the plasma status, data-exchange function of shot events, and monitoring of the facility operation. The software of the remote data analysis has been progressed significantly. The unified software infrastructure to access data stored both at on- and remote- sites and the documentation management system are also being developed. Numerical simulation for preparation and performance estimation of the shots by the appropriate implementation of the simulation code, such as the integrated simulation code, and the analysis codes for the plasma shots.

New conditions for Annular Finite-time Stability of Linear Systems

In this paper we investigate the annular finite-time stability (AFTS) problem for linear systems. A system is said to be annular finite-time stable if the norm of the system state remains within an upper and lower treshold for a given finite interval of time. Two necessary and sufficient conditions are provided for AFTS, the former requiring the solution of a differential Lyapunov equation (DLE), the latter involving an optimization feasibility problem constrained by differential linear matrix inequalities (DLMIs). We show that the DLE-based condition is more efficient from the computational point of view; however the DLMI-based condition is the starting point to investigate the design problem. To this regard a necessary and sufficient condition for the existence of a state feedback controller which renders the closed loop annular finite-time stable is provided. A numerical example illustrates the improvement of the proposed approach with respect to those existing literature.

A MIMO architecture for integrated control of plasma shape and flux expansion for the EAST tokamak

This paper deals with the magnetic control system of the EAST tokamak, i.e. with the system in charge of controlling the position and the shape of the plasma column into the vacuum vessel, as well as the current that flows into it. A multi-input multi-output architecture is proposed in order to achieve integrated control of both plasma boundary and flux expansion in the scrape-off layer. The proposed architecture is based on an XSC-like control approach [1]. In order to show the effectiveness of the proposed approach, a set of Matlab/Simulink � tools that allow to reproduce in simulation the EAST experiments is used. This simulation environment is validated against the experiments, and is then exploited to design and validate the proposed MIMO controller.

  PhD Thesis

A flexible architecture for plasma magnetic control in tokamak reactors

Plasma magnetic control is one of the core engineering issues to be tackled in a fusion device. Over the last years, model based approaches have been proposed to face this issue, proving their effectiveness and allowing to reduce the time span needed for control testing and validation. The first part of this work is intended to give an overview of the subject, from the historical milestones to the underlying physics; the most common techniques for tokamak plasmas electromagnetic modeling and control are also introduced and discussed. After this introduction, a general architecture for plasma magnetic control in tokamaks is proposed. Finally, the proposed solution is applied to the Experimental Advanced Superconducting Tokamak (EAST) tokamak, where a new plasma magnetic control architecture was developed and implemented during the 2016-2018 phd experimental campaigns, and to the Japan Torus-60 Super Advanced (JT-60SA) device, which is currently under construction in Japan.