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Cold atmospheric plasmas (CAPs) within recent years have shown great promise in the field of plasma medicine, encompassing a variety of treatments from wound healing to the treatment of cancerous tumors. For each subsequent treatment, a different application of CAPs has been postulated and attempted to best treat the target for the most effective results. These treatments have varied through the implementation of control parameters such as applied settings, electrode geometries, gas flow, and the duration of the treatment. However, with such an extensive number of variables to consider, scientists and engineers have sought a means to accurately control CAPs for the best-desired effects in medical applications. This paper seeks to investigate and characterize the historical precedent for the use of plasma control mechanisms within the field of plasma medicine. Current control strategies, plasma parameters, and control schemes will be extrapolated through recent developments and successes to gain better insight into the future of the field and the challenges that are still present in the overall implementation of such devices. Proposed approaches, such as data-driven machine learning, and the use of closed-loop feedback controls, will be showcased as the next steps toward application.

Conventionally, neutral beam injection (NBI) in tokamaks is controlled via engineering parameters such as injection voltage and power. Recently, the high-fidelity real-time NBI code RABBIT has been coupled to the discharge control system of ASDEX Upgrade. It allows to calculate the NBI fast-ion distribution and hence the properties of NBI in real-time, making it possible to control them directly. We successfully demonstrate control of driven current, ion heating and stored fast-ion energy by modifying the injected beam power. A combined ECRH and NBI controller is also successfully tested, which is able to adjust the heating mix between ECRH and NBI to match a certain desired ion heating fraction at given total power. Further experiments have been carried out towards control of the ion heat flux (i.e. ion heating plus collisional heat transfer between ions and electrons). They show good initial success, but also leave room for future improvements as the controller runs into instabilities at too high requests. [ABSTRACT FROM AUTHOR]

Reliable diagnostics that measure the detached state of the ITER divertor plasma will be necessary to control heat flux to the divertor targets during steady state, burning plasma operation. This paper conducts an initial exploration into the feasibility of the divertor shunt diagnostic as a lightweight, robust, and real-time detachment sensor. This diagnostic is a set of shunt lead pairs that measure the voltage drop along the divertor cassette body, from which the plasma scrape-off layer (SOL) current is calculated. Using SOLPS-ITER simulations for control-relevant ITER plasma scenarios, the thermoelectric current magnitude along the SOL is shown to decrease significantly with the onset of partial detachment at the outer divertor target. Electromagnetic modelling of a simplified divertor cassette is used to develop a control-oriented inductance-resistance circuit model, from which SOL currents can be calculated from shunt pair voltage measurements. The sensitivity and frequency-response of the resulting system indicates that the diagnostic will accurately measure SOL thermoelectric currents during ITER operation. These currents will be a good measure of the detached state of the divertor plasma, making the divertor shunt diagnostic a potentially extremely valuable and physically robust sensor for real-time detachment control. [ABSTRACT FROM AUTHOR]

The protection of the inner walls of magnetic confinement fusion research reactors is a crucial issue, particularly in this research context where plasma scenarios are explored to reach high power performance, thus leading to high temperature of the plasma facing components (PFCs), possibly close to their technological limitation. The aim is to protect the PFCs from damage during experimental campaigns, whilst enabling the expansion of the operational domain toward long duration and high power performances. With nearly 35 years of operation of Tore Supra and now WEST, CEA’s magnetic fusion research institute, the IRFM, has deployed a system combining thermal instrumentation, modeling of the heat transfer and photonic emission, signal processing and understanding of the physics of plasma-wall interaction to provide an optimized and controlled protection of the PFCs in metallic environment (with tungsten, bore, copper and stainless steel materials). In this context, the WEST Tokamak is a relevant Fusion facility capable of combining steady-state (Vloop ∼ 0) 1000 s long pulse operation with up to 6 MW m ^−2 on its divertor together with an advanced first wall protection system that could be deployed on ITER and future Fusion machines. The paper describes the wall protection system installed on WEST, highlighting its particular features and recent results.

In magnetic fusion devices, error field (EF) sources, spurious magnetic field perturbations, need to be identified and corrected for safe and stable (disruption-free) tokamak operation. Within Work Package Tokamak Exploitation RT04, a series of studies have been carried out to test the portability of the novel non-disruptive method, designed and tested in DIII-D (Paz-Soldan et al 2022 Nucl. Fusion 62 126007), and to perform an assessment of model-based EF control strategies towards their applicability in ITER. In this paper, the lessons learned, the physical mechanism behind the magnetic island healing, which relies on enhanced viscous torque that acts against the static electro-magnetic torque, and the main control achievements are reported, together with the first design of the asynchronous EF correction current/density controller for ITER.

The RFX-mod2 installation is planned to be completed by 2024 and the start of operations is expected in 2025. The high flexibility of the machine (already tested in the previous RFX-mod experiment) allows operation in Reversed Field Pinch and tokamak configuration as well as ultra-low q pulses. In this work we present predictive analysis on transport, performances and plasma control in RFX-mod2 in view of the first experimental campaigns.

To achieve its performance goals, SPARC plans to operate in equilibrium configurations with a strong elongation of $\kappa_\mathrm{areal}\sim1.75$ , which in turn will destabilize the n = 0 vertical instability. However, SPARC also features a relatively thick conducting wall that is designed to withstand disruption forces, leading to lower vertical instability growth rates than usually encountered. In this work, we use the TokSyS framework to survey families of accessible shapes near the SPARC baseline configuration, finding maximum growth rates in the range of $\gamma\lesssim100\,$ s ^−1 . The addition of steel vertical stability plates has only a modest ( ${\sim}25\%$ ) effect on reducing the vertical growth rate and almost no effect on the plasma controllability when the full vertical stability system is taken into account, providing flexibility in the plate conductivity in the SPARC design. Analysis of the maximum controllable displacement on SPARC is used to inform the power supply voltage and current limit requirements needed to control an initial vertical displacement of 5% of the minor radius. From the expected spectra of plasma disturbances and diagnostic noise, requirements for filter latency and vertical stability coil heating tolerances are also obtained. Small modifications to the outboard limiter location are suggested to allow for an unmitigated vertical disturbance as large as 5% of the minor radius without allowing the plasma to become limited. Further, investigations with the 3D COMSOL code reveal that strategic inclusion of insulating structures within the VSC supports are needed to maintain sufficient magnetic response. The workflows presented here help to establish a model for the integrated predictive design for future devices by coupling engineering decisions with physics needs.

In early 2022 one source of ASDEX Upgrade’s (AUG) neutral beam injector 2 was equipped with a first-of-its-kind beam extraction grid system with in-situ variable extraction/acceleration gap that allows one to choose beam energy and beam power independently in a wide operational space, greatly enhancing experimental flexibility. The gap can be changed from one AUG discharge to the next. The extended operational space makes it possible to reduce beam energy and shine through while maintaining high heating power, or to reduce NBI power at high beam energy, e.g. to stay in L mode. Furthermore, the feature opens the door for advanced control of heat and torque deposition, such as to change torque and ion-to-electron heating ratio at constant power. The prototype system was successfully tested in 2022 and already found first applications in AUG’s physics programme. The remaining three sources of the same injector will also be equipped with variable gaps during the 2022–24 opening of AUG and installation of this new system on all sources of NBI 1 is also under discussion in order to exploit the full potential of the new feature.

In 2019 the UK launched the Spherical Tokamak for Energy Production (STEP) programme to design and build a prototype electricity producing nuclear fusion power plant, aiming to start operation around 2040. The plant should lay the foundation for the development of commercial nuclear fusion power plants. The design is based on the spherical tokamak principle, which opens a route to high pressure, steady state, operation. While facilitating steady state operation, the spherical design introduces some specific plasma control challenges: (i) All plasma current during the burn phase should to be generated through non-inductive means, dominated by bootstrap current. This leads to operation at high normalised plasma pressure ${\beta _{\text{N}}}$ with high plasma elongation, which in turn imposes effective active stabilisation of the vertical plasma position. (ii) The tight aspect ratio means very limited space for a central solenoid, imposing that even the current ramp up must be non-inductively generated. (iii) The compact design leads to extreme heat loads on plasma facing components. A double null design has been chosen to spread this load, putting strict demands on the control of the unstable vertical plasma position. (iv) The heat pulses associated with unmitigated ELMs are unlikely to be acceptable imposing ELM free operation or active ELM control. (v) To reduce and spread heat loads, core and divertor radiation and momentum loss has to be controlled, aiming to operate with simultaneously detached upper and lower divertors. (vi) High pressure operation is likely to require active resistive wall mode (RWM) stabilisation. (vii) The conductivity distribution in structures near the plasma must be carefully selected to reduce the growth rates for the vertical instability and the RWM without damping the penetration of the of magnetic fields from active control coils too much. This article describes the initial work carried out to develop a STEP plasma control system.

In this work, model-predictive control (MPC) was combined for the first time with singular perturbation theory, and an original plasma kinetic control method based on extremely simple data-driven models and a two-time-scale MPC algorithm has been developed. A comprehensive review is presented in this paper. Slow and fast semi-empirical models are identified from data, by considering the fast kinetic plasma dynamics as a singular perturbation of a quasi-static equilibrium, which itself is governed, on the slow time scale, by the flux diffusion equation. This control technique takes advantage of the large ratio between the time scales involved in magnetic and kinetic plasma transport. It is applied here to the simultaneous control of the safety factor profile, q (𝑥), and of several kinetic variables, such as the poloidal beta parameter, β _p , and the internal inductance parameter, l _i , on the EAST tokamak. In the experiments, the available control actuators were lower hybrid current drive (LHCD) and co-current neutral beam injection (NBI) from different sources. Ion cyclotron resonant heating (ICRH) and electron cyclotron resonant heating (ECRH) are used as additional actuators in control simulations. In the controller design, an observer provides, in real time, an estimate of the system states and of the mismatch between measured and predicted outputs, which ensures robustness to model errors and offset-free control. Based on the observer information, the controller predicts the behavior of the system over a given time horizon and computes the optimal actuation by solving a quadratic programming optimization problem that takes the actuator constraints into account. A number of control applications are described in the paper, either in nonlinear simulations with EAST-like parameters or in real experiments on EAST. The simulations were performed with a fast plasma simulator (METIS) using either two control actuators (LHCD and ICRH) in a low density scenario, or up to four actuators at higher density: LHCD, ECRH, and two NBI systems driven in a on/off pulse-width-modulation (PWM) mode, with different injection angles. The control models are identified with the prediction-error method, using datasets obtained from open loop simulations in which the actuators are modulated with pseudo-random binary sequences. The simulations with two actuators show that various q (𝑥) profiles and β _p waveforms can be tracked without offset, within times that are consistent with the resistive and thermal diffusion time scales, respectively. In simulations with four actuators, simultaneous tracking of time-dependent targets is shown for q (𝑥) at two normalized radii, 𝑥 = 0 and 𝑥 = 0.4, and for β _p . Due to the inherent mismatch between the optimal NBI power request and the delivered PWM power, the kinetic controller performs with reduced accuracy compared with simulations that do not use the NBI/PWM actuators. The first experimental tests using this new control algorithm were performed on EAST when the only available actuator was the LHCD system at 4.6 GHz. The algorithm was thus used in its simplest single-input-single-output version to track time-dependent targets for the central safety factor, q _0 , or for β _p . In the closed loop control experiments, the q _0 targets were tracked in about one second, consistently with the plasma resistive time constant. Excellent tracking of a piecewise linear β _p target waveform was also achieved. When the NBI system became controllable in real time by the EAST plasma control system, new experiments were dedicated to multiple-input-multiple-output MPC control with three actuators: LHCD and two NBI actuators using the PWM algorithm. Given that the minimum time allowed between NBI on/off switching was 0.1 s, i.e. larger than the characteristic time of the fast plasma dynamics, a reduced version of the MPC controller based only on the slow model was used. Various controller configurations were tested during a single experimental session, with up to three controlled variables chosen among q _0 = q ( 𝑥 = 0), q _1 = q ( 𝑥 = 0.5), β _p and l _i . The main difficulty encountered during this session was the unavailability of the full baseline ICRH and ECRH powers that were used in the reference scenario, and from which the plasma model was identified. This often led to the saturation of one or several actuators, which prevented some targets selected in advance from being accessible. Nevertheless, in cases that were free from actuator saturation, q _0 and q _1 targets were successfully reached, in a time that is consistent with the resistive diffusion time of the model and with small oscillations that are characteristic of the PWM operation of the neutral beams. During the simultaneous control of q _0 and β _p , the ICRH power was too low and, in addition, the plasma density was much larger than the reference one. The q _0 targets were not accessible in this high-density/low-power case, but β _p control was successful. Finally, the simultaneous control of q _0 and l _i was satisfactory and, during the simultaneous control of, q _0 , β _p and l _i , the tracking of β _p and l _i was satisfactory but q _0 was too large due to the lack of ICRH power and to NBI saturation. In conclusion, the extensive nonlinear simulations described in this paper have demonstrated the relevance of combining MPC, data-driven models and singular perturbation methods for plasma kinetic control. This technique was also assessed experimentally on EAST, although some tests were perturbed by undesired parameter changes with respect to the reference scenario.

In the last experimental campaign (OP1.2b) of the stellarator Wendelstein 7-X (W7-X), boronisation as a mean for first wall conditioning was applied for the first time which led to strongly reduced impurity fluxes from plasma-facing components. Thermal detachment at the uncooled target plates of the test divertor unit (TDU) was reached at higher plasma densities and was accompanied by high recycling of neutrals at the target plate [1], [2]. A feedback control system was established in W7-X to actively control the gas injection (actuator) for plasma fuelling and impurity seeding [3] through the divertors. It allowed very successful stabilisation of the detached plasma condition state as well as mitigation of thermal overloads to some baffle tiles. Different routinely available diagnostic signals were used as input parameters (sensors). We describe the setup of the feedback control system, its performance and provide some example results with the main focus on the development of the control scheme which led to the detachment stabilisation over the entire longest (30 s) high-power discharge at W7-X so far. In view of the achieved very successful detachment stabilisation and the necessity to include simultaneous optimisation of the core performance in the future, the feedback control system is being upgraded for the upcoming campaign (OP2.1) in which the water cooled and hereby inherently steady-state capable divertor has been currently installed. The prospects and some experiment ideas for active detachment control are discussed.

We fabricated nanocrystalline Ge films using radio-frequency (RF) magnetron plasma sputtering deposition under a high Ar-gas pressure. The Ge nanograins changed from amorphous to crystalline when the distance between the Ge sputtering target and the substrate was decreased to 5 mm and the RF input power was 11.8 W/cm2 (60 W), where the deposition rate was as high as 660 nm/min. In addition, the size of the nanocrystalline grains increased from 100 to 307 nm when the RF input power for plasma production was increased from 11.8 W/cm2 (60 W) to 17.7 W/cm2 (90 W). In the developed narrow-gap plasma process at sub-Torr pressures, nanocrystalline Ge films were successfully fabricated on Cu substrates at low temperatures, without the substrate being heated. However, when annealing was conducted under an N2 atmosphere, which is the conventional method to induce solid-phase crystallization, the amorphous Ge layer on a Cu substrate changed to a Cu3Ge crystal layer through interdiffusion of Ge and Cu atoms at 400–500 °C.

• Empirical error field correction currents have been deduced by the execution of the non-disruptive compass scan method. • Empirical error field control has been tested at JET, allowing locked mode spin up and exploration of lower density regime. • The design of the empirical error field controller for ITER is presented. This work reports on error field identification and control studies carried out at JET and insights on the development of the empirical EF controller for ITER. The empirical EF controller has been included in the JET real-time central controller following the execution of the non-disruptive compass scan method (Paz-Soldan C. et al., Nuclear Fusion 62 (2022) 126007), which allowed the identification of the EF source and the currents for error field compensation. When testing the empirical EF controller, born locked n = 1 modes have been observed to spin-up and a lower density regime has been explored in the 1.8 MA plasma current, 2.1 T toroidal magnetic field scenario than otherwise achievable. These experimental results demonstrate the benefits of EF correction. In preparation of EF correction studies in ITER, the empirical EF controller for ITER has been developed and integrated in the plasma control system database. [ABSTRACT FROM AUTHOR]

A first-ever demonstration of controlling power and torque injection through time evolution of neutral beam energy has been achieved in recent experiments at the DIII-D tokamak (Luxon 2002 Nucl. Fusion 42 614). Pre-programmed waveforms for the neutral beam energy produce power and torque inputs that can be separately and continuously controlled. Previously, these inputs were tailored using on/off modulation of neutral beams resulting in large perturbations (e.g. power swings of over 1 MW). The new method includes, importantly for experiments, the ability to maintain a fixed injected power while varying the torque. In another case, different beam energy waveforms (in the same plasma conditions) produce significant changes in the observed spectrum of beam ion-driven instabilities. Measurements of beam ion loss show that one energy waveform results in the complete avoidance of coherent losses due to Alfvénic instabilities. This new method of neutral beam operation is intended for further application in a variety of DIII-D experiments including those concerned with high-performance steady state scenarios, fast particle effects, and transport in the low torque regime. Developing this capability would provide similar benefits and improved plasma control for other magnetic confinement fusion facilities.

A first-ever demonstration of controlling power and torque injection through time evolution of neutral beam energy has been achieved in recent experiments at the DIII-D tokamak (Luxon 2002 Nucl. Fusion 42 614). Pre-programmed waveforms for the neutral beam energy produce power and torque inputs that can be separately and continuously controlled. Previously, these inputs were tailored using on/off modulation of neutral beams resulting in large perturbations (e.g. power swings of over 1 MW). The new method includes, importantly for experiments, the ability to maintain a fixed injected power while varying the torque. In another case, different beam energy waveforms (in the same plasma conditions) produce significant changes in the observed spectrum of beam ion-driven instabilities. Measurements of beam ion loss show that one energy waveform results in the complete avoidance of coherent losses due to Alfvénic instabilities. This new method of neutral beam operation is intended for further application in a variety of DIII-D experiments including those concerned with high-performance steady state scenarios, fast particle effects, and transport in the low torque regime. Developing this capability would provide similar benefits and improved plasma control for other magnetic confinement fusion facilities.

In order to realize advanced plasma control based on visible cameras, a high-speed image acquisition and processing system has been developed recently on Experimental Advanced Superconducting Tokamak (EAST). This system is optimized in many ways to achieve high-speed, real-time, low-latency performance, and to load multiple acquisition cards simultaneously. The acquisition rate of this system can be close to 10 000 FPS when the frame size is set as 320 $\times240$ and the pixel depth is set as 8 bits. DMA is used for high-speed data transmission; the memory copy function is optimized for reducing the time cost on data memory reading and writing. Besides, the visualization subsystem based on the Python web can communicate with the acquisition machines and also can synthesize data of multiple acquisition machines in real time to perform image fusion and access display. In addition, a thermal event recognition function based on visible imaging is also included in this system. The convolutional neural network (CNN) model of hot spots and multifaceted asymmetric radiation from the edge (MARFE) detection for EAST plasma discharge has been developed so far. The detection results can be visualized in quasi-real time. In terms of data storage, a new data storage format is designed; a GUI for off-line data analysis and processing based on MATLAB is provided. [ABSTRACT FROM AUTHOR]

For the first time, a real-time capable NBI code, which has a comparable fidelity to the much more computationally expensive Monte Carlo codes such as NUBEAM, has been coupled to the discharge control system of a tokamak. This implementation has been done at ASDEX Upgrade and is presented in this paper. Modifications to the numerical scheme of RABBIT for the time-dependent solution of the Fokker–Planck equation have been carried out to make it compatible with the non-equidistant time-steps, as they occur in real-time simulations. We demonstrate that this allows RABBIT to run in real-time both in a steady-state and time-dependent fashion and show and discuss an actual real-time simulation. Its accuracy is identified by comparing to offline RABBIT and TRANSP-NUBEAM runs (where more diagnostics are available for preciser inputs).

During the last decade a new probe design for active plasma resonance spectroscopy, the multipole resonance probe (MRP), was proposed, analyzed, developed, and characterized in two different designs: the spherical MRP (sMRP) and the planar MRP (pMRP). The advantage of the latter is that it can be integrated into the chamber wall and can minimize the perturbation of the plasma. Both designs can be applied for monitoring and control purposes of plasma processes for industrial applications. As usual for this measurement technique, a mathematical model is required to determine plasma parameter (electron density, electron temperature, and collision frequency of electrons with neutral atoms) from the measured resonances. Based on the cold plasma model a simple relationship between the resonance frequency and the electron density can be derived and leads to excellent measurement results. However, a simultaneous measurement of the electron temperature in low-pressure plasmas requires a kinetic model, because the half-width of the resonance peak is broadened by kinetic effects. Such a model has been derived and first results show the broadening of the spectra as expected. Deriving a relation between the half-width and the electron temperature will allow the simultaneous measurement and an improvement of monitoring and control concepts. [ABSTRACT FROM AUTHOR]

In magnetic fusion devices, undesired non-axisymmetric magnetic field perturbations, typically called error fields (EF), have been observed to have a detrimental effect on plasma stability and confinement and can lead to brute plasma terminations, i.e. plasma disruption events. The main strategies that can be adopted to minimize the effect of EFs on the plasma dynamics consist in a careful alignment of the coils, when assembling the fusion device, and, most commonly, in the use of EF correction coils which counteract the non-axisymmetric fields by prescribing properly designed correction currents. In this work, an assessment of the n = 1 EF source in MAST-U is presented. When constructing the MAST-U device, an optimization of poloidal and divertor coil positions has been adopted to reduce the n = 1 EF source. This optimization consisted in the application of coil shifts and tilts, of the order of mms and mrads, respectively. To investigate the presence of a residual n = 1 EF dedicated studies have been performed during the first MAST-U campaign. The compass scan method was employed to identify the n = 1 EF, relying on the detection of the locked mode (LM) onset, which proved to be a challenging task. Therefore, a methodology based on Transfer Functions (TFs) among each coil and the n = 1 radial magnetic field has been developed which allows the detection of LM formation. Such method is described here, complemented with the experimental results achieved, which suggest that the intrinsic n = 1 EF source on MAST-U is relatively small with respect to MAST. Indeed, the empirical correction currents for n = 1 EF minimization are smaller, about 0.2 kA, than the ones used in MAST, 1 kA range. This proves that the optimal coil alignment for n = 1 EF minimization has been a successful strategy in MAST-U. • Error field detection studies have been performed in MAST-U suggesting a small amplitude of n=1 error field. • To this purpose a new method based on transfer function identification has been developed and successfully tested to retrieve the plasma response, the key quantity for detecting locked mode onset times. [ABSTRACT FROM AUTHOR]

The magnetized nuclear fusion plasma is a non-linear dynamic system with limits and constraints. It requires a sophisticated plasma control system with a wide variety of functions and components to ensure optimal and safe performance. A graph-based modelling framework is proposed for the integrated development of physics models, plasma scenarios and control systems. The framework contains actuators and sensors, continuous plasma processes and variables, discrete plasma states and events, and requirements. Most importantly, it defines the couplings between these elements. A Dependency Structure Matrix (DSM), a technique to represent and organize complex graphs, analyses these couplings to reveal a potential global system layout. The framework is demonstrated for ITER, resulting in a fully traceable graph model. The DSM suggests that the system can be organized into five distinct groups: Heating and current drive, magnetic configuration, burn dynamics, transport and exhaust, and plasma–wall interaction. Each group consists of actuators, sensors and physics. All couplings between groups are made apparent in the DSM. Although ITER features specific actuators and sensors, these groups appear common for magnetically confined fusion devices. • A graph model defines the interactions between tokamak physics and control systems. • Dependency Structure Matrix techniques divide the model into five coupled subsystems. • The model supports allocation of actuators and sensors to control functions. [ABSTRACT FROM AUTHOR]

Experiments at DIII-D and EAST have demonstrated simultaneous high confinement, divertor detachment, and active control of detachment level, all of which are required for ITER. Comparing detachment control via Teand Jsat, it appears that Teis the most straightforward sensor to use for accessing detachment onset, while Jsatoffers more precise control of degree of detachment. Based on these results, control using nitrogen seeding has so far shown the best ability to follow a target value with the low disruptivity and little to no degradation of performance when an Internal Transport Barrier (ITB) is present, but not all facilities allow its use. Neon seeding also can be paired with feedback control with low impact on core performance as long as there is no disruption, however shots with neon seeding commonly disrupted during these experiments. Argon is effective in EAST, but tends to degrade performance (by ≈10%βp) when detachment is achieved. With ideal conditions and strike point position control, data from a single Langmuir probe are an acceptable input to the control algorithm, but this simple system is easily defeated by strike point displacement comparable to the Teor Jsatscale lengths. The presence of an ITB seems to be critical to retaining core performance in detachment in these parameter ranges, as the pedestal pressure tends to decrease as a result of impurity seeding.

Turbulence in the plasma sheath around reentry vehicles is known to contribute to radio-communications blackout, but a practical laboratory model of that extreme environment remains elusive. Herein, we present a table-top plasma system with sustained, chaotic convection for that purpose. Strong sound waves exert acoustic radiation pressure on gradients within the plasma and are shown to drive sufficient convection to cause abrupt and chaotic variation in the plasma properties. The volume-averaged plasma conductivity and collision time are determined in real time by phase-sensitive detection of a microwave probe signal. The experiment provides unique opportunities to study transmission into plasma conditions that can inform detailed models of high-temperature turbulent flows. [ABSTRACT FROM AUTHOR]

The roadmap to the realization of fusion energy describes a path towards the development of a DEMO tokamak reactor, which is expected to provide electricity into the grid by the mid of the century (Romanelli, 2013). The DEMO diagnostic and control (D&C) system must provide measurements with high reliability and accuracy, not only constrained by space restrictions in the blanket, but also by adverse effects induced by neutron, gamma radiation and particle fluxes. In view of the concept development for DEMO control, an initial selection of suitable diagnostics has been obtained (Biel et al., 2019). This initial group of diagnostic consists of 6 methods: Microwave diagnostics, thermo-current measurements, magnetic diagnostics, neutron/gamma diagnostics, IR interferometry/polarimetry, and a variety of spectroscopic and radiation measurement systems. A key aspect for the implementation, performance and lifetime assessment of these systems on DEMO, is mainly attributable to their location, that must be well protected, and meet their own set of specific requirements. With this in mind, sightline analysis, space consumption and the evaluation of optical systems are the main assessment tools to obtain a high level of integration, reliability and robustness of all this instrumentation; essential features in future commercial fusion power nuclear plants. In this paper we concentrate on spectroscopic and radiation measurement systems that require sightlines over a large range of plasma regions and inner reactor surfaces. Moreover, this paper outlines the main results and strategies adopted in this early stage of DEMO conceptual design to assess the feasibility of this initial set of diagnostic methods based on sightlines and the integration of these needed for DEMO D&C. [ABSTRACT FROM AUTHOR]

In future tokamak devices, the control system will have to handle several control goals simultaneously with a limited number of actuators in long and high performance discharges. One of the critical roles of the future control systems will be the management of actuators, which would assign the most convenient available actuators primarily to the control goals of the highest importance at the time. Such a system would consist of a discharge program defining the experiment, a discharge supervisor making automatic high level decisions in real time and a component handling the actuators at a lower level: The virtual actuator, which is a software object responsible for distributing the controller commands to a set of selected actuators. This paper describes the implementation of a virtual actuator for all 8 gyrotrons of ASDEX-Upgrade. We also describe the intended use of the virtual actuator for three experiments: β control using ECRH, a disruption avoidance strategy, and electron temperature profile control. The paper also gives an overview of future actuator management developments at ASDEX-Upgrade: extension to all heating sources, inclusion of mirrors for ECRH, and intelligent real time distribution of the actuators between the control goals. [ABSTRACT FROM AUTHOR]

• The flight simulator provides rapid simulation of a simplified ASDEX Upgrade physics model combined with a DCS control system model. • It is targeted for rapid discharge program validation prior to each plasma pulse and for model-oriented control system development. • The physics model can be extended enabling more detailed investigations at longer simulation time. • ITER PCSSP, Matlab/Simulink, as well as ASTRA-SPIDER are used as enabling technologies. Discharge scenarios and control schemes in ASDEX Upgrade (AUG) are evolving more and more complex. Especially in physics investigations for ITER and DEMO sophisticated scenarios exploit the operational space. This increases the probability of design flaws or human errors in the pulse configuration, but also aggravates the potential damage in the failure case. The ASDEX Upgrade Flight Simulator Fenix, which is currently under construction, will provide a fast and efficient simulation tool for testing and validating discharge scenarios, as well as control and monitoring functions, during their development and immediately prior to experimental pulse execution. This ensures, that the scenarios and settings are adequate to reach the experimental goals and that the margins to operational limits are sufficiently large also during the dynamic evolution of the discharge. Simplified physics and plant system "control" models combined with a representation of the ASDEX Upgrade Discharge Control System (DCS) allow for fast simulation runs with reasonable prediction quality. In the simulation an event generator can trigger plasma instabilities, technical failures and external events to test the resilience of the designed pulse against unplanned incidents. The granularity of modelling shall be customizable, such that the simulator can also be used for detail investigations with elaborate physics at the cost of longer simulation time. As a basis for implementation the ITER Plasma Control System Simulation Platform (PCSSP) has been chosen. The flight simulator, extends PCSSP with an ASTRA co-simulator for the ASDEX Upgrade tokamak model and with custom modules for its actuators, diagnostics and control system. Plugins will enable reading original AUG discharge programs and configuration files, as well as storing the results in the AUG shot file database. [ABSTRACT FROM AUTHOR]

• The magnetic-flux diffusion equation is combined with a plasma-energy balance equation to obtain a control-oriented response model for control design. • The electron temperature, plasma resistivity, and lower-hybrid current drive are modeled by following an uncertainty-based approach. • The problem of designing a model-based controller for simultaneous q-profile and plasma-energy regulation is formulated as an optimization problem. • The tradeoff between two competing objectives, namely the tracking error and the control effort, is optimally solved during the control design process. • The tracking performance of the proposed controller, which is robust against the model uncertainties, is successfully tested in disturbance-rejection nonlinear simulations. Integrated control of the toroidal current density profile, or alternatively the q -profile, and plasma stored energy is essential to achieve advanced plasma scenarios characterized by high plasma confinement, magnetohydrodynamics stability, and noninductively driven plasma current. The q -profile evolution is closely related to the evolution of the poloidal magnetic flux profile, whose dynamics is modeled by a nonlinear partial differential equation (PDE) referred to as the magnetic-flux diffusion equation (MDE). The MDE prediction depends heavily on the chosen models for the electron temperature, plasma resistivity, and non-inductive current drives. To aid control synthesis, control-oriented models for these plasma quantities are necessary to make the problem tractable. However, a relatively large deviation between the predictions by these control-oriented models and experimental data is not uncommon. For this reason, the electron temperature, plasma resistivity, and non-inductive current drives are modeled for control synthesis in this work as the product of an "uncertain" reference profile and a nonlinear function of the different auxiliary heating and current-drive (H&CD) source powers and the total plasma current. The uncertainties are quantified in such a way that the family of models arising from the modeling process is able to capture the q -profile and plasma stored energy dynamics from a typical EAST shot. A control-oriented nonlinear PDE model is developed by combining the MDE with the "uncertain" models for the electron temperature, plasma resistivity, and non-inductive current drives. This model is then rewritten into a control framework to design a controller that is robust against the modeled uncertainties. The resulting controller utilizes EAST's H&CD powers and total plasma current to regulate the q profile and plasma stored energy even when mismatches between modeled and actual dynamics are present. The effectiveness of the controller is demonstrated through nonlinear simulations. [ABSTRACT FROM AUTHOR]

• Toroidal current density profile control is critical for the NSTX-U missions. • A control-oriented plasma response model has been developed earlier for the NSTX-U. • Based on this model, a current profile controller is designed for the NSTX-U. • The controller implements a linear-quadratic-integral optimal control strategy. • The performance of the proposed controller is validated in TRANSP simulations. Active control of the toroidal current density profile is critical for the upgraded National Spherical Torus eXperiment device (NSTX-U) to maintain operation at the desired high-performance, MHD-stable, plasma regime. Initial efforts towards current density profile control have led to the development of a control-oriented, physics-based, plasma-response model, which combines the magnetic diffusion equation with empirical correlations for the kinetic profiles and the non-inductive current sources. The developed control-oriented model has been successfully tailored to the NSTX-U geometry and actuators. Moreover, a series of efforts have been made towards the design of model-based controllers, including a linear-quadratic-integral optimal control strategy that can regulate the current density profile around a prescribed target profile while rejecting disturbances. In this work, the tracking performance of the proposed current-profile optimal controller is tested in numerical simulations based on the physics-oriented code TRANSP. These high-fidelity closed-loop simulations, which are a critical step before experimental implementation and testing, are enabled by a flexible framework recently developed to perform feedback control design and simulation in TRANSP. [ABSTRACT FROM AUTHOR]

• An integrated architecture has been tested in DIII-D by using ONFR as a supervisor. • The proposed architecture enables simultaneous q-profile, beta, and NTM control. • Authority over ECH&CD is shared by the q-profile + beta and NTM control algorithms. • Preliminary experimental testing shows how competing control objectives are handled. There is an increasing need for integrating individual plasma-control algorithms with the ultimate goal of simultaneously regulating more than one plasma property. Some of these integrated-control solutions should have the capability of arbitrating the authority of the individual plasma-control algorithms over the available actuators within the tokamak. Such decision-making process must run in real time since its outcome depends on the plasma state. Therefore, control architectures including supervisory and/or exception-handling algorithms will play an essential role in future fusion reactors like ITER. However, most plasma-control experiments in present devices have focused so far on demonstrating control solutions for isolated objectives. In this work, initial experimental results are reported for simultaneous current-profile control, normalized-beta control, and Neoclassical Tearing Mode (NTM) suppression in DIII-D. Neutral beam injection (NBI), electron-cyclotron (EC) heating & current drive (H&CD), and plasma current modulation are the actuation methods. The NBI power and plasma current are always modulated by the Profile Control category within the DIII-D Plasma Control System (PCS) in order to control both the current profile and the normalized beta. EC H&CD is utilized by either the Profile Control or the Gyrotron categories within the DIII-D PCS as dictated by the Off-Normal and Fault Response (ONFR) system, which monitors the occurrence of an NTM and regulates the authority over the gyrotrons. The total EC power and poloidal mirror angles are the gyrotron-related actuation variables. When no NTM suppression is required, the gyrotrons are used by the Profile Control category, but when NTM suppression is required, the ONFR transfers the authority over the gyrotrons to the NTM stabilization algorithm located in the Gyrotron category. Initial experimental results show that simultaneous control of different aspects of the plasma dynamics may improve the overall control and plasma performances. Also, the potential of the ONFR system to successfully integrate competing control algorithms is demonstrated. [ABSTRACT FROM AUTHOR]

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). [ABSTRACT FROM AUTHOR]

Due to the discrete nature of the saddle coils for the active feedback control of the plasma instabilities, the emergence of the sideband modes is inevitable. In the Keda Torus eXperiment (KTX), the edge Mirnov coils are unevenly distributed on the inner surface of the vacuum vessel to suppress the sideband modes. These arrays of the Mirnov coils are used for the feedback control system of KTX. They are also used for the magnetohydrodynamics (MHD) mode identification, which is a fundamental and significant method to distinguish and describe the plasma instabilities. A set of suitable MHD mode analysis methods has been utilized to complete the mode detection, including the spatial discrete Fourier transform (SDFT) method, the singular value decomposition (SVD) method and the Lomb periodogram method. These methods can obtain comprehensive mode information of the plasma instabilities and their results can be used as the feedback of the feedback control system. These methods have been successfully applied to detect and characterize an impulsive mode of ($m = 1$ , $n = 0$) and rotating modes of ($m = 2$ , 3, 4, $n = 1$) in the tokamak plasma of KTX, while the matrix decomposition technique is not applicable for KTX. [ABSTRACT FROM AUTHOR]

Preliminary implementation of plasma GAP shape control in EAST is presented. Compared with ISOFLUX method, GAP control could work together with not only real-time equilibrium reconstruction but also real-time boundary reconstruction. Based on the plasma response matrix, the GAP controller is designed and the PID parameter has been optimized for EAST plasma shape control. P-EFIT provides real-time gap calculation and GAP controller is integrated into EAST PCS. Successful plasma shape with GAP control scheme is achieved in EAST 2018 summer campaign. Experimental results prove that GAP control algorithm and controller could achieve similar control ability as ISOFLUX and provide an option for shape control with optical-based boundary reconstruction in EAST and future device CFETR. [ABSTRACT FROM AUTHOR]

Abstract Fast vertical control of shaped plasmas is essential for the successful realization of plasma operation near the limit of vertical control to achieve maximum confinement and stable, disruption free operation. In KSTAR, the normal conducting in-vessel vertical control coil (IVC coil) is employed to respond to vertical transients much faster than the superconducting coils are capable. The power supply for the IVC coil is capable of responding in a time commensurate with the expected vertical growth rates on KSTAR. The diagnostics used for the fast vertical control since the first operation with shaped plasmas in KSTAR, however, have a low signal-to-noise ratio which limits the gains that can be used successfully in the proportional and derivative control loops and thus the speed of the control loop. The successful use of relative flux for the Z-position estimate and of the loop voltage difference from a pair of up-down symmetric loops to provide sensitive and less noisy vertical estimate will be discussed. In addition, the control loop for the IVC coil is decoupled from the slow motion controlled by the superconducting coils using a high-pass filter in the control software. Finally the noticeable improvement in plasma control for plasmas near the vertical stability limit will be discussed. [ABSTRACT FROM AUTHOR]

The plasma configuration in tokamaks is created and controlled by magnetic fields generated by the external coils. Similarly, plasma macroscopic parameters are determined using sensors measuring magnetic field or magnetic flux. Some algorithms require separation of the plasma contribution from the contribution of poloidal field coils and currents induced in passive structures. In this paper, an efficient, real-time applicable method is introduced. The method is based on the identification of a state-space model in vacuum shots using both experimental and finite-element model data. The tests were performed for ISTTOK and RFX-mod. In , a simple method for plasma position determination for ISTTOK is proposed. We show that the centroid position measurement by magnetic diagnostics is feasible using our algorithm for separation of the plasma signal. The correctness of the resulting method is proven by comparison with the results of the heavy-ion beam. [ABSTRACT FROM AUTHOR]

In order to reconstruct the plasma boundary directly and independently, a new method based on visible imaging diagnostics is developed for experimental advanced superconducting tokamak (EAST); its effectiveness has been verified by conducting offline experiments based on EAST’s historical data. The offline boundary reconstruction results, including X-points, are compared to offline equilibrium fitting code (offline EFIT), which is a commonly used reconstruction method. With an average error of 1.5 cm, the total processing time for one frame is less than 2 ms by current software and hardware. When the camera sensor is not saturated, the algorithm is robust for different image intensities of the plasma discharge. The results of the boundary reconstruction by visible imaging diagnostics also show potential for real-time plasma control in a tokamak. [ABSTRACT FROM AUTHOR]

Currently, the fast evolution of cameras in recent years has made them promising tools for diagnostics of Tokamak. The solution presented in this paper consists of a prototype of high-speed visible image acquisition and processing system (HVIAPs) dedicated for experimental advanced superconducting tokamak shape and position control. Graphics processing unit (GPU) and field-programmable gate array (FPGA) are typically used as accelerators or co-processors in addition to a CPU. Such a heterogeneous computing system can combine the advantages of its individual components. The main imaging equipment components used for this system are four high-resolution fast cameras equipped with fiber interface. Image data from the cameras are received by the frame grabber card based on the FPGA and transmitted to the GPU via the peripheral component interconnect express interface. The software support for the system includes low-level drivers and application programming interface libraries for all components and the algorithm developed for image processing. The offline image process results are compared to equilibrium fitting code, which is a commonly used reconstruction method, with an average error of 1.5 cm. The total processing time for one frame is less than 0.3 ms. [ABSTRACT FROM AUTHOR]

The authors constructed a dynamic simulation model of a nuclear fusion power plant on Modelica language to obtain fundamental knowledge on the plasma control requirements for the future commercial fusion power plants. The fusion power plant model was designed with a 1500-MW thermal output tokamak reactor with He-cooled Li2TiO3 solid breeder blanket (coolant outlet conditions: 8 MPa and 515.8 °C). A superheated Rankine cycle was designed to achieve the electrical output of 485.38 MW with the operating pressure of 20.5 MPa. Two plasma output patterns, a step decrease of power and a single pulse decrease of power, were simulated to assess the response of the power plant. A sudden step decrease in fusion neutron led to an immediate decrease in the blanket temperature and the first coolant temperature. In order to avoid the sharp temperature drop, a need for a turbine bypass mechanism or a He coolant boiler bypass mechanism was indicated. On the other hand, because of the delay in the plant responses, the deviation of the electrical output from steady state could be minimized by recovering the plasma output in few tens of seconds. Based on the findings, a new diagram was presented that illustrates an important plasma control requirements for future commercial fusion power plants. [ABSTRACT FROM AUTHOR]

Results on the integration and the operation of the KSTAR plasma control system (PCS) upgrade are given. Real-time hardware, new realtime-capable operating system, and a brand-new data acquisition are assembled in order to extend the performance and compatibility with modern computer systems. The first full commissioning and eventual routine use performed in 2016 so that the system can now acquire more than 400 channels for more than 100 seconds of data with 5 kHz sampling. The performance test results are summarized, featuring In/Out streaming echo tests and the synchronization verifications. Examples of general performance improvements are demonstrated, and additional features added to the software are also described. [ABSTRACT FROM AUTHOR]

Method for automatic detection of L-H transition using Support Vector Machine (SVM), a popular tool of supervised machine learning tools, has been evaluated in order to improve plasma density control in KSTAR. Through the SVM, a nonlinear classifier is trained to distinguish L-mode and H-mode using two kinds of diagnostic data measured in KSTAR. The trained classifier has been analyzed for possible usage on the real-time detection through the truncation of the training samples. Study on the optimization of the training samples, and corresponding accuracy change is made for evaluating feasibility for real-time implementations. [ABSTRACT FROM AUTHOR]