An initial concept for the plasma diagnostic and control (D&C) system has been developed as part of European studies towards the development of a demonstration tokamak fusion reactor (DEMO). The main objective is to develop a feasible, integrated concept design of the DEMO D&C system that can provide reliable plasma control and high performance (electricity output) over extended periods of operation. While the fusion power is maximized when operating near to the operational limits of the tokamak, the reliability of operation typically improves when choosing parameters significantly distant from these limits. In addition to these conflicting requirements, the D&C development has to cope with strong adverse effects acting on all in vessel components on DEMO (harsh neutron environment, particle fluxes, temperatures, electromagnetic forces, etc.). Moreover, space allocation and plasma access are constrained by the needs for first wall integrity and optimization of tritium breeding. Taking into account these boundary conditions, the main DEMO plasma control issues have been formulated, and a list of diagnostic systems and channels needed for plasma control has been developed, which were selected for their robustness and the required coverage of control issues. For a validation and refinement of this concept, simulation tools are being refined and applied for equilibrium, kinetic and mode control studies.

This report summarizes the contributions presented at the IAEA technical meeting on plasma disruptions and their mitigation, held virtually, 20-23 July 2020. The meeting brought together more than 120 experts from nuclear fusion research sites worldwide to discuss experimental, theoretical and modelling work in the field of plasma disruptions with special emphasis on developing a solid basis for possible disruption mitigation strategies in ITER and next generation fusion devices. The main topics of the meeting were: (i) disruption consequences, including electromagnetic loads, heat loads, and runaway electrons; (ii) disruption prediction and avoidance, including machine learning and physics-based approaches, and control aspects; and (iii) disruption mitigation, including shattered pellet injection, alternative techniques and general aspects of disruption mitigation.

RFX-mod2 is an upgrade of RFX-mod that will use a modified shell and mechanical structure in order to enhance plasma-shell proximity and therefore to improve plasma control. The Control and Data Acquisition System for most of the plant systems and diagnostics previously used in RFX-mod will be refurbished, while others will be completely re-built. The most important component that will be completely renewed is the ElectroMagnetic probe (EM) data acquisition system, where a new architecture based on XILINX Zynq FPGA will be used to carry out at the same time both high-speed data acquisition and resampled data streaming for active plasma control. The use of MDSplus will be retained in RFX-mod2, while the MARTe framework used for real-time plasma control will be replaced by MARTe2, a new framework developed under strict software quality standards. Plant control in RFX-mod2 will be supervised by WinCC-OA, replacing the previous FactoryLink SCADA systems. Older plant systems such as vacuum control based on outdated S5 PLCs will be updated and will use OPC-UA for communication with the supervisory control system.

The plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) will have to provide reliable operation near technical and physics limits, while its front-end components will be subject to strong adverse effects within the nuclear and high temperature plasma environment. The ongoing developments for the ITER D&C system represent an important starting point for progressing towards DEMO. Requirements for detailed exploration of physics are however pushing the ITER diagnostic design towards using sophisticated methods and aiming for large spatial coverage and high signal intensities, so that many front-end components have to be mounted in forward positions. In many cases this results in a rapid aging of diagnostic components, so that additional measures like protection shutters, plasma based mirror cleaning or modular approaches for frequent maintenance and exchange are being developed. Under the even stronger fluences of plasma particles, neutron/gamma and radiation loads on DEMO, durable and reliable signals for plasma control can only be obtained by selecting diagnostic methods with regard to their robustness, and retracting vulnerable front-end components into protected locations. Based on this approach, an initial DEMO D&C concept is presented, which covers all major control issues by signals to be derived from at least two different diagnostic methods (risk mitigation).

The system of diagnostics, data acquisition and control foreseen on the Divertor Test Tokamak experiment (DTT) is presented. Conceived in an integrated way, the control system meets the specifications of a fusion experiment devoted to the study of the power exhaust problem in view of DEMO. Diagnostics and feedback control are particularly functional to the need of maintaining the plasma close to equilibrium in situations prone to instabilities where the plasma wall interaction is optimized. Strongly oriented to the exploration of control methods suitable for DEMO, DTT will specifically experiment on physics and engineering model based control systems. Control and data flow schemes are inspired by those of ITER.

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.

In view of the preparation for the operation of the ITER tokamak it is necessary to develop the plasmascenarios taking into account all engineering constraints coming from the plant and including a realisticcontrol system. It is important to consider that, due to the high energy of ITER plasmas, much morestringent requirements are posed on the control of transients in order to avoid machine damage.Several activities are performed in the EU focusing on one side on the scenario optimization from aphysics point of view and on the other side on the design and modeling of a realistic plasma controlsystem driving the plasma configuration throughout the whole pulse and suitable for implementationon a real machine.The issues related to the computation of the control feed-forward component are addressed. In par-ticular, the possibility to trigger a feed-forward component to solve controllability problems arising inthe transitions from plasma L to H and H to L modes is studied in detail with the support of linear andnon-linear simulations.A control strategy is designed and tested on non-linear simulations of the whole pulse, including linearand non-linear effects due to controller switching, plasma shape reconstruction and power supplies.The paper reports on the results of the studies performed and discuss the proposed design of the plasmacontrol system.© 2014 Published by Elsevier B.V.

Abstract The Plasma Control System (PCS) of the Frascati Tokamak Upgrade (FTU) is not equipped with any runaway electron (RE) beam control or suppression tool. In this paper we propose an upgraded {PCS} including an architecture for the control of disruption-generated {REs} that, making use of filtering techniques to estimate the onsets of the current quench (CQ) and of the {RE} beam current plateau, provides a controlled plasma current shut-down and a simultaneous {RE} position control. The control strategy is based on a nonlinear technique, called Input Allocation, that allows to re-configure the current in the poloidal field (PF) coils and improve the {PCS} responsiveness needed for {RE} position control. Preliminary results on the implementation of the Input Allocation and an experimental proposal to test the control scheme architecture are discussed.

Input-output datasets from two magnetic confinement fusion (MCF) experiments of the reversed-field pinch (RFP) type are examined. The RFP datasets, which are samples of the distributed magnetic field dynamics, are naturally divided into many smaller batches due to the pulsed-plasma operation of the experiments. The two RFP experiments considered are (i) EXTRAP T2R (T2R) with 64 inputs and 64 outputs and (ii) RFX-mod (RFX) with 192 inputs and 192 outputs. Both T2R and RFX are magnetohydrodynamically unstable and operates under magnetic feedback with optional dither injection. Using subspace system identification techniques and randomised cross-validation (CV) methods to minimise the generalisation error, state-space orders of the empirical systems are suggested. These system orders are compared to "stabilisation diagrams" commonly used in experimental modal analysis practice. The relation of the CV system order to the decay of the singular values from the subspace method is observed. Both (i) stable vacuum diffusion and (ii) unstable plasma response datasets are analysed. Apparent simulation and prediction errors are quantified for both cases using a deviation-accounted-for index. These results are purely data-driven. A simple approach towards exploitation of the subspace techniques for finite-element model refinement and data confrontation is presented.

This paper describes the new JET enhancement project "Plasma Control Upgrade" (PCU). Initially aimed at an overhaul of JET plasma control capabilities it was eventually focused on improving the vertical stabilisation (VS) system ability to recover from large ELM (edge localised mode) perturbations. The paper describes the results of the first two years where the activity was aimed principally at researching a solution that could be implemented within the timing and budget constraints. A very important task was that of improving the modelling of JET plasma, iron core and passive structures. Using dedicated experiments, the models were progressively refined until it was possible not just to explain the experimental data but predict the VS system behaviour. At the same time the project team studied the best options for power supply (PS) and control system upgrades and evaluated whether a change of turns in the stabilisation coil was desirable and possible. A new fast radial field power supply is now being ordered and the VS control system is being upgraded.

Suppression of plasma instabilities is a key issue to improve the confinement time of controlled thermonuclear fusion with tokamaks. Radiofrequency sources are currently used to stabilize the sawteeth and Neoclassic Tearing Modes instabilities that degrade the plasma performance. In particular, Electron Cyclotron Radiofrequency Heating (ECRH) is a suitable way to control the growth of instabilities. With this technique, in fact, it is possible to drive a localized current exactly in the narrow region of the plasma where the instabilities occur. The theory has shown the feasibility of a control strategy by means of a real-time system based on ECRH, that reduces the growth of non-linear tearing magnetic surfaces that develop into rotating island topology. This paper will show that a real-time instability control system based on the application of Electron Cyclotron Current Drive has been installed on Frascati Tokamak Upgrade (FTU) (8 T magnetic field, 1.6 MA plasma current), and the first tests will be presented. A typical architecture of an ECRH system devoted to the control of instabilities includes: the real-time detection of the instabilities by proper plasma diagnostics and an actively steered actuator that injects EC beams towards their target during the plasma discharge. To solve the hard real-time requirements of the actuator system, a homemade system based on a PXI embedded controller has been realized to be completely integrated into the main control system of the FTU tokamak. The system acts via the Digital Waveform Reference of the Electron Cyclotron gyrotrons and it is hardware driven by an external clock, triggers, and gates coming from a controller process of a feedback loop. (c) 2006 Elsevier B.V. All rights reserved.