Plasma Magnetic Control Research Articles

Current Limit Avoidance Algorithms for DEMO Operation

Tokamaks are the most promising devices to prove the feasibility of energy production using nuclear fusion on Earth which is foreseen as a possible source of energy for the next centuries. In large tokamaks with superconducting poloidal field (PF) coils, the problem of avoiding saturation of the currents is of paramount importance, especially for a reactor such as the European demonstration fusion power plant DEMO. Indeed, reaching the current limits during plasma operation may cause a loss of control of the plasma shape and/or current, leading to a major disruption. Therefore, a current limit avoidance (CLA) system is essential to assure safe operation. Three different algorithms to be implemented within a CLA system are proposed in this paper: two are based on online solutions of constrained optimization problems, while the third one relies on dynamic allocation. The performance assessment for all the proposed solutions is carried out by considering challenging operation scenarios for the DEMO reactor, such as the case where more than one PF current simultaneously saturates during the discharge. An evaluation of the computational burden needed to solve the allocation problem for the various proposed alternatives is also presented, which shows the compliance of the optimization-based approaches with the envisaged deadlines for real-time implementation of the DEMO plasma magnetic control system.

Modelling, design and simulation of plasma magnetic control for the Spherical Tokamak for Energy Production (STEP)

Magnetic controller performance requirements and design solutions for the Spherical Tokamak for Energy Production (STEP) are driven by the need to produce equilibrium and scenario trajectories, maintaining steady plasma vertical stabilization and shape control for a period of thousands of seconds or longer while avoiding contact with the plasma-facing components (PFCs). Axisymmetric magnetic control schemes including vertical stabilization, plasma current control, plasma shape control and poloidal field coil current control for the STEP Prototype Reactor (SPR) concept are being developed using a suite of control analysis tools (known collectively as TokSys) supporting the integrated plasma control design process. The vertical growth rate based on a linear rigid plasma response model in TokSys is used for assessing the controllability of the vertical instability in SPR. TokSys closed-loop simulations with the axisymmetric non-linear, free-boundary evolution code, GSevolve, are performed for the assessment, identification, and verification of algorithm implementation and controller performance for the various axisymmetric control systems. Dynamic control performance and controllability are shown to be consistent with noise-affected scenario requirements under reasonable power supply and sensor performance assumptions.

Virtualization of accelerators in embedded systems for mixed-criticality: RPU exploitation for fusion diagnostics and control

The need for computing systems featuring high performance, safety, security, and predictability requirements, has become essential for fusion diagnostics and control. In this scenario, the next-generation Multiprocessors System-On-Chip (MPSoCs) characterized by modern processors and hardware accelerators can expand the potential of control infrastructure. To enhance the usability, performance, and dependability of these new technologies and to reduce the deployment cost, the concept of virtualization sees its first application in embedded systems. This work aims to propose three high-level architectural solutions that leverage virtualization to exploit accelerators and Real-Time Processing Units (RPUs) present on MPSoCs to realize Mixed-Criticality Systems (MCSs). That is, systems capable to manage on a single chip real-time tasks for critical automatic control together with other value-added services on the same hardware. The problems to be solved to implement these proposals will be explained in detail to guide future works. To highlight the potential of virtualization for real-time embedded systems in fusion science scenarios we explore some use cases regarding Vertical Stabilization (VS) algorithms for plasma magnetic control that can benefit from the proposed solutions.

Methodology of Plasma Shape Reachability Area Estimation in D-Shaped Tokamaks

This paper suggests and develops a new methodology of estimation for a multivariable reachability region of a plasma separatrix shape on the divertor phase of a plasma discharge in D-shaped tokamaks. The methodology is applied to a spherical Globus-M/M2 tokamak, including the estimation of a controllability region of a vertical unstable plasma position on the basis of the experimental data. An assessment of the controllability region and the reachability region of the plasma is important for the design of tokamak poloidal field coils and the synthesis of a plasma magnetic control system. When designing a D-shaped tokamak, it is necessary to avoid the small controllability region of the vertically unstable plasma, because such cases occur in practice at a restricted voltage on a horizon field coil. To make the estimations mentioned above robust, PID-controllers for vertical and horizontal plasma position control were designed using the Quantitative Feedback Theory approach, which stabilizes the system and provides satisfactory control indexes (stability margins, setting time, overshoot) during plasma discharges. The controllers were tested on a series of plasma models and nonlinear models of current inverters in auto-oscillation mode as actuators for plasma position control. The estimations were made on these models, taking into account limitations on control actions, i.e., voltages on poloidal field coils. This research is the first step in the design of the plasma shape feedback control system for the operation of the Globus-M2 spherical tokamak. The developed methodology may be used in the design of poloidal field coil systems in tokamak projects in order to avoid weak achievability and controllability regions in magnetic plasma control. It was found that there is a strong cross-influence from the PF-coils currents and the CC current on the plasma shape; hence, these coils should be used to control the plasma shape simultaneously.

Tokamak plasma models development for plasma magnetic control systems design by first principle equations and identification approach

To design plasma magnetic control systems in modern tokamaks one needs models of the plasma. These models are linear parameter varying (LPV) because of relatively small variations of outputs of the feedback loops around scenarios and may be obtained by first-principle equations, identification approach or by their combinations. The challenge of obtaining models of the plasma in the tokamak becomes more complicated by the existence of plasma position and poloidal field control loops with diagnostics and actuators which are needed as inner cascades for plasma current and shape control in particular on the Globus-M2 tokamak (Ioffe Inst., St. Petersburg, RF). The plasma equilibrium is reconstructed on the base of magnetic measurements outside the plasma by Picard iterations, moving filaments or neural networks, and linear plasma models are developed around the equilibrium with the help of the Kirchhoff’s law and force balance. In order to ensure the operability of plasma current and shape feedback control systems, the identification approach (controlled plant model design on the base of experimental input-output signals) is planned to be used. The basic methods of the identification are supposed to be applied as follows: subspaces, wavelets, linear matrix inequalities (LMIs), adaptive state observers, and dynamical neural networks which are able to automatically adjust their states to an unknown plant (elements of artificial intelligence). The solutions of the identification problem will be compared with the models obtained by the first principles with the aim to get the sufficient accuracy of coincidence. The approaches to be developed of getting tokamak plasma models may be applied to any vertically elongated (D-shaped) operating tokamak such as Globus-M2 (RF), D-IIID, NSTX (US), JET, ST40 (GB), ASDEX Upgrade (Germany), TCV (Switzerland), EAST (China), Damavand (Iran) etc.

Plasma Control in Tokamaks

The plasma magnetic control systems for ITER (International Thermonuclear Experimental Reactor) are presented. The systems comprise original engineering solutions for plasma position, current and shape control for the two versions of ITER: ITER-1 and ITER-2, including those designed in V. A. Trapeznikov Institute of Control Sciences of the RAS. It is noted that in ITER-1 the plasma position and shape were controlled by all PF-coils and robust H∞-controllers, while to decrease peaks of control power for suppressing minor disruptions the additional nonlinear circuit was used without significant changes in displacements of the gaps between the plasma separatrix and the first wall. In ITER-2 the special circuit with a fast voltage rectifier was used for plasma vertical speed stabilization about zero, while for plasma current and shape control the special cascade control systems were designed with and without the control channels decoupling, with robust H∞-controllers and predictive model, and with adaptive stabilization of the plasma vertical position as well. To increase the plasma controllability region in the vertical direction the additional horizontal field coils were introduced into the ITER-2 vacuum vessel and the capabilities of the system with the new circuits to control the plasma vertical position at the noise presence were investigated.

Plasma magnetic time-varying nonlinear robust control system for the Globus-M/M2 tokamak

This paper is devoted to the development of the magnetic plasma control system for the Globus-M/M2 spherical tokamak (ST) when minor disruptions occur. The novelty of the research is the design of robust cascade multivariable control systems for time-varying plasma models in the presence of nonlinear thyristor current inverters as actuators for plasma position control. A new architecture of coordinated control of plasma position and shape is proposed. Quantitative Feedback Theory is used for the synthesis of robust Single-Input Single-Output (SISO) controllers for plasma position and PF-currents. Multiple-Input Multiple-Output (MIMO) H∞ control is applied to control the plasma shape. The simulation results of the gaps-based MIMO plasma shape control system with the plasma equilibrium reconstruction code in the feedback are shown. The robust stability of the control system is estimated.

Robust Cascade LMI Design of MIMO Control System for Plasma Position, Current, and Shape Model with Time-Varying Parameters in a Tokamak

A new robust hierarchical plasma magnetic control system with time-varying parameters and cascade circuits for the Globus-M2 tokamak (Ioffe Institute) has been synthesized by means of the linear matrix inequalities (LMI) method. Each control cascade has a separate objective: placing poles of the closed-loop system in the D-region to guarantee robust performance, restriction of the H∞-norm of the transfer function between external disturbances and the plant outputs of the closed-loop system, tracking PF-currents, CS-current, plasma current, plasma position, gaps between the plasma separatrix and the first wall. Control cascades were synthesized for an array of plant linear models corresponding to reconstructed plasma equilibria at different time points of the tokamak discharges. The LMIs allow the synthesis of one LTI controller, providing the required performance and system stability margin for each model from a given array. Tracking cascades use new MIMO PID controllers adjusted by means of the LMIs.

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.

Plasma Control in Tokamaks. Part 1.

Various concepts of tokamaks as leaders in solving the controlled thermonuclear fusion problem are presented. The evolution of tokamaks from round in vertical cross-section tokamaks with a large aspect ratio up to tokamaks with a small aspect ratio including spherical ones is characterized. The classification of modern tokamaks according to their poloidal systems with the location of the poloidal field coils inside and outside of the toroidal field coil is given, taking into account the presence of coils inside a vacuum vessel to stabilize the plasma position. The methods of plasma diagnostics by magnetic measurements outside the plasma, actuators for both plasma magnetic and kinetic control, associated with plasma additional heating, plasma magnetic and kinetic models, instabilities and disruptions of a plasma discharge are described.

Simulation suite for plasma magnetic control at EAST tokamak

In view of ITER operations, many experiments are being conducted on the EAST tokamak. In particular, new magnetic control approaches are being developed and need to be tested. In order to design and test a plasma current, position and shape control, an accurate modeling of EAST static equilibria and dynamic plasma evolution is required. Hence, a suite of simulation tools has been developed for the reconstruction of plasma equilibria, the generation of plasma linearized models and for the closed loop testing of magnetic control algorithms. These tools are meant to assess the reliability of such equilibria and models, in order to provide a robust tool for the purposes of control design.

Plasma Magnetic Control in Tokamak Devices

In tokamak experimental reactors, the magnetic control system is one of the main plasma control systems that is required, together with the density control, since the very beginning, even before first operations. Indeed, the magnetic control drives the current in the external poloidal circuits in order to first achieve the breakdown conditions and, after plasma formation, to track the desired plasma current, shape and position. Furthermore, when the plasma poloidal cross-section is vertically elongated, the magnetic control takes also care of the vertical stabilization of the plasma column, and therefore it is an essential system for operation. This chapter introduces a reference architecture for plasma magnetic control in tokamaks. Given the proposed architecture, the techniques to design all the required control algorithms is also presented. Experimental results obtained on the JET and EAST tokamaks and simulations for machines currently under construction are shown to prove the effectiveness of the proposed architecture and control algorithms.

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.

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.

Synthesizing a tokamak plasma magnetic control circuit with allowance for the characteristics of the power supply system

The problem of controlling the position and shape of the plasma in a tokamak is considered. Although a great many works have been devoted to this problem, most of them use relatively simple models of a power system that acts as an actuator. Full power system models are rarely used, even during simulations. A mathematical model of a thyristor converter, one of the main elements of a power system, is proposed here in the form of a switching system. An algorithm based on a predictive control model is described. A numerical simulation of the model system’s operation is performed.

Plasma vertical stabilisation in ITER

This paper describes the progress in analysis of the ITER plasma vertical stabilisation (VS) system since its design review in 2007–2008. Two indices characterising plasma VS were studied. These are (1) the maximum value of plasma vertical displacement due to free drift that can be stopped by the VS system and (2) the maximum root mean square value of low frequency noise in the dZ/dt measurement signal used in the VS feedback loop. The first VS index was calculated using the PET code for 15 MA plasmas with the nominal position and shape. The second VS index was studied with the DINA code in the most demanding simulations for plasma magnetic control of 15 MA scenarios with the fastest plasma current ramp-up and early X-point formation, the fastest plasma current ramp-down in a divertor configuration, and an H to L mode transition at the current flattop. The studies performed demonstrate that the VS in-vessel coils, adopted recently in the baseline design, significantly increase the range of plasma controllability in comparison with the stabilising systems VS1 and VS2, providing operating margins sufficient to achieve ITER's goals specified in the project requirements. Additionally two sets of the DINA code simulations were performed with the goal of assessment of the capability of the PF system with the VS in-vessel coils: (i) to control the position of runaway electrons generated during disruptions in 15 MA scenarios and (ii) to trigger ELMs in H-mode plasmas of 7.5 MA/2.65 T scenarios planned for the early phase of ITER operation. It was also shown that ferromagnetic structures of the vacuum vessel (ferromagnetic inserts) and test blanket modules insignificantly affect the plasma VS.

The Initial Design of Process Control System Based on EPICS for HL-2A & HL-2M

In addition to the real time plasma magnetic and kinetic control, process control is also very important for the operation of a Tokamak. Process control aims at the normal control of the plant systems including information sharing, schema of the Tokamak operation, and alarm handling. A new tokamak named “HL-2M” will be built during the next four years and many conceptual designs for the operation of the new device might be validated on HL-2A, which is currently running at SWIP Chengdu, China. For those purposes, we select the EPICS as the basis for the process control system of HL-2A. Some user interface, data management, and development tools are adopted such as control system studio (CSS), MS SQL Server, Matlab, LabVIEW, VC++, and Java. In this study, we will describe the initial design of the whole HL-2A/HL-2M process control system and its related achievements in HL-2A. c © 2012 The Japan Society of Plasma Science and Nuclear Fusion Research

Current, Position, and Shape Control in Tokamaks

The need to achieve increasingly better performance in present and future tokamak devices has made plasma control increasingly important in tokamak engineering. When high performance and robustness are required, it is essential to adopt a model-based approach to design a control system. We introduce the basics of plasma current, position, and shape control in tokamaks. As an example, the plasma magnetic control systems of the JET tokamak is presented, together with an approach proposed for plasma axisymmetric magnetic control at the ITER tokamak.

Coupling plasmas and 3D passive structures in the JET tokamak

The impact of the 3D effects on the dynamic response of the plasma in the JET tokamak is assessed on the time scale of interest for plasma magnetic control and stabiliza tion. These effects are due to the limbs of the magnetic circuit and to the eddy currents in the 3D metallic structures. The plasma behaviour is taken into account by a specific matrix of modifi ed inductances interacting with the remaining part of the syst em via a virtual coupling surface.