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An overview of recent results obtained at the tokamak ASDEX Upgrade (AUG) is given. A work flow for predictive profile modelling of AUG discharges was established which is able to reproduce experimental H-mode plasma profiles based on engineering parameters only. In the plasma center, theoretical predictions on plasma current redistribution by a dynamo effect were confirmed experimentally. For core transport, the stabilizing effect of fast ion distributions on turbulent transport is shown to be important to explain the core isotope effect and improves the description of hollow low-Z impurity profiles. The L–H power threshold of hydrogen plasmas is not affected by small helium admixtures and it increases continuously from the deuterium to the hydrogen level when the hydrogen concentration is raised from 0 to 100%. One focus of recent campaigns was the search for a fusion relevant integrated plasma scenario without large edge localised modes (ELMs). Results from six different ELM-free confinement regimes are compared with respect to reactor relevance: ELM suppression by magnetic perturbation coils could be attributed to toroidally asymmetric turbulent fluctuations in the vicinity of the separatrix. Stable improved confinement mode plasma phases with a detached inner divertor were obtained using a feedback control of the plasma β. The enhanced Dα H-mode regime was extended to higher heating power by feedback controlled radiative cooling with argon. The quasi-coherent exhaust regime was developed into an integrated scenario at high heating power and energy confinement, with a detached divertor and without large ELMs. Small ELMs close to the separatrix lead to peeling-ballooning stability and quasi continuous power exhaust. Helium beam density fluctuation measurements confirm that transport close to the separatrix is important to achieve the different ELM-free regimes. Based on separatrix plasma parameters and interchange-drift-Alfvén turbulence, an analytic model was derived that reproduces the experimentally found important operational boundaries of the density limit and between L- and H-mode confinement. Feedback control for the X-point radiator (XPR) position was established as an important element for divertor detachment control. Stable and detached ELM-free phases with H-mode confinement quality were obtained when the XPR was moved 10 cm above the X-point. Investigations of the plasma in the future flexible snow-flake divertor of AUG by means of first SOLPS-ITER simulations with drifts activated predict beneficial detachment properties and the activation of an additional strike point by the drifts. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under Grant Agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. Peer Reviewed "Article signat per més de 50 autors/es: U. Stroth, D. Aguiam, E. Alessi, C. Angioni, N. Arden, R. Arredondo Parra, V. Artigues, O. Asunta, M. Balden, V. Bandaru, A. Banon-Navarro, K. Behler, A. Bergmann, M. Bergmann, J. Bernardo, M. Bernert, A. Biancalani, R. Bielajew, R. Bilato, G. Birkenmeier, T. Blanken, V. Bobkov, A. Bock, T. Body, T. Bolzonella, N. Bonanomi, A. Bortolon, B. Böswirth, C. Bottereau, A. Bottino, H. van den Brand, M. Brenzke, S. Brezinsek, D. Brida, F. Brochard, C. Bruhn, J. Buchanan, A. Buhler, A. Burckhart, Y. Camenen, B. Cannas, P. Cano Megias, D. Carlton, M. Carr, P. Carvalho, C. Castaldo, M. Cavedon, C. Cazzaniga, C. Challis, A. Chankin, C. Cianfarani, F. Clairet, S. Coda, R. Coelho, J.W. Coenen, L. Colas, G. Conway, S. Costea, D. Coster, T. Cote, A.J. Creely, G. Croci, D.J. Cruz Zabala, G. Cseh, A. Czarnecka, I. Cziegler, O. D'Arcangelo, A. Dal Molin, P. David, C. Day, M. de Baar, P. de Marné, R. Delogu, S. Denk, P. Denner, A. Di Siena, J.J. Dominguez Palacios Durán, D. Dunai, A. Drenik, M. Dreval, R. Drube, M. Dunne, B.P. Duval, R. Dux, T. Eich, S. Elgeti, A. Encheva, K. Engelhardt, B. Erdös, I. Erofeev, B. Esposito, E. Fable, M. Faitsch, U. Fantz, M. Farnik, H. Faugel, F. Felici, O. Ficker, S. Fietz, A. Figueredo, R. Fischer, O. Ford, L. Frassinetti, M. Fröschle, G. Fuchert, J.C. Fuchs, H. Fünfgelder, S. Futatani, K. Galazka, J. Galdon-Quiroga, D. Gallart Escolà, A. Gallo10, Y. Gao11, S. Garavaglia3, M. Garcia Muñoz16, B. Geiger21, L. Giannone1, S. Gibson32, L. Gil2, E. Giovannozzi18, S. Glöggler1, M. Gobbin8, J. Gonzalez Martin, T. Goodman, G. Gorini, T. Görler, D. Gradic, G. Granucci, A. Gräter, H. Greuner, M. Griener, M. Groth, A. Gude, L. Guimarais, S. Günter, G. Haas, A.H. Hakola, C. Ham, T. Happel, N. den Harder, G. Harrer, J. Harrison, V. Hauer, T. Hayward-Schneider, B. Heinemann, T. Hellsten, S. Henderson, P. Hennequin, A. Herrmann, E. Heyn, F. Hitzler, J. Hobirk, K. Höfler, J.H. Holm, M. Hölzl, C. Hopf, L. Horvath, T. Höschen, A. Houben, A. Hubbard, A. Huber, K. Hunger, V. Igochine, M. Iliasova, T. Ilkei, K. Insulander Björk, C. Ionita-Schrittwieser, I. Ivanova-Stanik, W. Jacob, N. Jaksic, F. Janky, A. Jansen van Vuuren, A. Jardin, F. Jaulmes, F. Jenko, T. Jensen, E. Joffrin, A. Kallenbach, S. Kálvin, M. Kantor, A. Kappatou, O. Kardaun, J. Karhunen4, C.-P. Käsemann, S. Kasilov, A. Kendl, W. Kernbichler, E. Khilkevitch, A. Kirk, S. Kjer Hansen, V. Klevarova, G. Kocsis, M. Koleva, M. Komm, M. Kong, A. Krämer-Flecken, K. Krieger, A. Krivska, O. Kudlacek, T. Kurki-Suonio, B. Kurzan, B. Labit, K. Lackner, F. Laggner, A. Lahtinen, P.T. Lang, P. Lauber, N. Leuthold, L. Li, J. Likonen, O. Linder, B. Lipschultz, Y. Liu, A. Lohs, Z. Lu, T. Luda di Cortemiglia, N.C. Luhmann, T. Lunt, A. Lyssoivan, T. Maceina, J. Madsen, A. Magnanimo, H. Maier, J. Mailloux, R. Maingi, O. Maj, E. Maljaars, P. Manas, A. Mancini, A. Manhard, P. Mantica, M. Mantsinen, P. Manz, M. Maraschek, C. Marchetto, L. Marrelli, P. Martin, A. Martitsch, F. Matos, M. Mayer, M.-L. Mayoral, D. Mazon, P.J. McCarthy, R. McDermott, R. Merkel, A. Merle, D. Meshcheriakov, H. Meyer, D. Milanesio, P. Molina Cabrera, F. Monaco, M. Muraca, F. Nabais, V. Naulin, R. Nazikian, R.D. Nem, A. Nemes-Czopf, G. Neu, R. Neu, A.H. Nielsen, S.K. Nielsen, T. Nishizawa, M. Nocente, J.-M. Noterdaeme, I. Novikau, S. Nowak, M. Oberkofler, R. Ochoukov, J. Olsen, F. Orain, F. Palermo, O. Pan, G. Papp, I. Paradela Perez, A. Pau, G. Pautasso, C. Paz-Soldan, P. Petersson, P. Piovesan, C. Piron, U. Plank, B. Plaum, B. Plöck, V. Plyusnin, G. Pokol, E. Poli, L. Porte, T. Pütterich, M. Ramisch, J. Rasmussen, G. Ratta, S. Ratynskaia, G. Raupp, D. Réfy, M. Reich1, F. Reimold, D. Reiser, M. Reisner, D. Reiter, T. Ribeiro, R. Riedl, J. Riesch, D. Rittich, J.F. Rivero Rodriguez, G. Rocchi, P. Rodriguez-Fernandez, M. Rodriguez-Ramos, V. Rohde, G. Ronchi, A. Ross, M. Rott, M. Rubel, D.A. Ryan, F. Ryter, S. Saarelma, M. Salewski, A. Salmi, O. Samoylov, L. Sanchis Sanchez, J. Santos, O. Sauter, G. Schall, K. Schlüter, K. Schmid, O. Schmitz, P.A. Schneider, R. Schrittwieser, M. Schubert, C. Schuster, T. Schwarz-Selinger, J. Schweinzer, E. Seliunin, A. Shabbir, A. Shalpegin, S. Sharapov, U. Sheikh, A. Shevelev, G. Sias, M. Siccinio, B. Sieglin, A. Sigalov, A. Silva, C. Silva, D. Silvagni, J. Simpson, S. Sipilä, E. Smigelskis, A. Snicker, E. Solano, C. Sommariva, C. Sozzi, G. Spizzo, M. Spolaore, A. Stegmeir, M. Stejner, J. Stober, E. Strumberge1, G. Suarez Lopez, H.-J. Sun, W. Suttrop, E. Sytova, T. Szepesi, B. Tál, T. Tala, G. Tardini, M. Tardocchi, D. Terranova, M. Teschke, E. Thorén, W. Tierens, D. Told, W. Treutterer, G. Trevisan, E. Trier, M. Tripský, M. Usoltceva, M. Valisa, M. Valovic, M. van Zeeland, F. Vannini, B. Vanovac, P. Varela, S. Varoutis, N. Vianello, J. Vicente, G. Verdoolaege, T. Vierle, E. Viezzer, I. Voitsekhovitch, U. von Toussaint, D. Wagner, X. Wang, M. Weiland, A.E. White, M. Willensdorfer, B. Wiringer, M. Wischmeier, R. Wolf, E. Wolfrum, Q. Yang, Q. Yu, R. Zagórski, I. Zammuto, T. Zehetbauer, W. Zhang, W. Zholobenko, M. Zilker, A. Zito, H. Zohm, S. Zoletnik and the EUROfusion MST1 Team "

The plasma design point during stationary phase operation of EU-DEMO has been described extensively in the past, and the assumptions employed for its definition have been supported by various investigations. However, a stationary snapshot is by far not sufficient to ensure the actual operability of the plasma scenario. On the con-trary, both normal and off-normal transients have a strong impact on the machine design, providing the most challenging requirements to the actuators. This has to be accounted for in the design from the early phases on. In this paper, the main requirements originating from plasma operation, studied during the Pre-Concept Design Phase, are briefly reviewed. These encompass the so-called planned transients, e.g. plasma current ramps, access and termination of the burn phase or plasma fueling with pellets, but also the mitigation of off-normal events which may lead to a disruption, e.g. the divertor reattachment or an uncontrolled increase of radiation after a sudden W influx from the wall.

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.

Future fusion reactors require a safe, steady state divertor operation. The required detached operation is, in tokamaks with metal walls, usually achieved by seeding of impurities, such as nitrogen. With strong seeding levels, the dominant radiation is emitted from a small, poloidally localized volume inside the confined region, in the vicinity of the X-point. The location of the radiating volume is observed to vary relative to the X-point depending on seeding and power levels, i.e. depending on the degree of detachment. At the ASDEX Upgrade tokamak, the position of the radiator relative to the X-point can be controlled in real time by a modulation of the nitrogen puff level. At a certain height of the radiator above the X-point, an ELM-suppressed regime is observed with minimal reduction of confinement. While the control of the X-point radiator already allows operation in full detachment at a dissipated power fraction of around 95 %, which is required for a future reactor and was previously never achieved in a controlled way, such an ELM-suppressed regime additionally eliminates the challenge of the transient, intolerably high heat fluxes by ELMs. Both requirements are met in the presented regime while maintaining a high energy confinement at high density.

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.

Fenix, the ASDEX Upgrade (AUG) flight simulator under development, is based on the Plasma Control System Simulation Platform (PCSSP) developed for ITER, the ASTRA transport code and the SPIDER equilibrium code. Fenix will give a session leader the possibility to check whether the discharge will meet experimental goals prior to execution. It is also designed to facilitate the development of control system features and the validation of physical models. It reads the AUG discharge program and checks if all the parameters and reference waveforms are reasonable during the discharge simulation. ASTRA serves as a plant model (physical model of AUG tokamak) which outputs idealised diagnostic signals (temperature, density, etc.). ASTRA calculates these data from the particle and energy transport in the plasma core, plasma edge, Scrape of Layer (SOL) and divertor. It also includes particle balance, L-H transition and sawtooth models, and it is equipped with the 2D equilibrium reconstruction code SPIDER. The second component of Fenix is a model of the Discharge Control System (DCS) and AUG actuators. The DCS model processes diagnostic signals from ASTRA, computes commands for actuators and sends them back to ASTRA closing the feedback loop. Controllers for coil currents to control plasma current, position and shape are implemented in the model as well as gas puff valves (divertor and midplane) and the pellet injector to control electron density. The DCS model also simulates external heating actuators such as NBI, ECRH and ICRH. Fenix has methods for reading configuration files and archiving. This article presents modelling of the ASTRA and DCS components, and the first results from the simulations of the flattop phase.

In the present work, the role of plasma facing components protection in driving the EU-DEMO design will be reviewed, focusing on steady-state and, especially, on transients. This work encompasses both the first wall (FW) as well as the divertor. In fact, while the ITER divertor heat removal technology has been adopted, the ITER FW concept has been shown in the past years to be inadequate for EU-DEMO. This is due to the higher foreseen irradiation damage level, which requires structural materials (like Eurofer) able to withstand more than 5 dpa of neutron damage. This solution, however, limits the tolerable steady-state heat flux to ~1 MW/m2, i.e. a factor 3–4 below the ITER specifications. For this reason, poloidally and toroidally discontinuous protection limiters are implemented in EU-DEMO. Their role consists in reducing the heat load on the FW due to charged particles, during steady state and, more importantly, during planned and off-normal plasma transients. Concerning the divertor configuration, EU-DEMO currently assumes an ITER-like, lower single null (LSN) divertor, with seeded impurities for the dissipation of the power. However, this concept has been shown by numerous simulations in the past years to be marginal during steady-state (where a detached divertor is necessary to maintain the heat flux below the technological limit and to avoid excessive erosion) and unable to withstand some relevant transients, such as large ELMs and accidental loss of detachment. Various concepts, deviating from the ITER design, are currently under investigation to mitigate such risks, for example in-vessel coils for strike point sweeping in case of reattachment, as well as alternative divertor configurations. Finally, a broader discussion on the impact of divertor protection on the overall machine design is presented.

openaire: EC/H2020/633053/EU//EUROfusion An Electron Cyclotron Wave (ECW) system will be used in the European DEMO for the stabilization of Neoclassical Tearing Modes (NTMs). In order to avoid movable mirrors in the harsh environment close to the plasma and to simplify the NTM launcher integration, the tuning of the ECW deposition location can be achieved by launching frequency-tunable ECWs from fixed mirrors while the frequency is tuned in discrete steps of 2–3GHz. An overview of the frequency step-tunable ECW system for NTM stabilization is presented. The design considerations are discussed based on the current DEMO baseline parameters and the status of technologies. A simulation of NTM stabilization with an idealized frequency tunable ECW system on an analytical NTM model is shown. The simulation takes into account a realistic tuning speed based on the present technology and considers the current NTM launcher configurations in DEMO. A simple sweeping strategy is adapted for the control of frequency. Various uncertainties, which will affect the feasibility, need to be further investigated.

Actuator management is an essential part of a modern tokamak plasma control system. It has to deal with a large number of control task simultaneously, needs to be able to operate close to stability limits and to avoid disruptions. In the ASDEX Upgrade tokamak experiment, the process of actuator management development is ongoing. As a first step, we have removed direct assignement of physical actuators to controllers responsible for the control task execution. Instead, the controller is communicating to the so-called virtual actuator, which groups an arbitrary number of actuators of the same type (so far ECRH only) and distributes the controller command between them. The virtual actuator has been implemented and used on ASDEX Upgrade in the 2019 experimental campaign. It gave valuable contributions to several physics experiments. First, the implementation of the virtual actuator enabled β p control by ECRH power, which is of interest especially for I-mode and ELM-free H-mode discharges. Second, it was used in the electron temperature control experiments, which is a useful control tool for transport studies. Third, the concept was used for the density limit disruption avoidance experiments where central ECRH power is applied to recover the plasma from a disruptive zone. The current paper describes the use of the virtual actuator in these cases and based on the gained experience, it suggests future developments and improvements of the virtual actuator on ASDEX Upgrade discharge control system.

Efficient fuelling will be an essential task in the EU-DEMO. The basic requirement here is to establish the target plasma core density with a minimum particle flux by injecting mm-size solid fuel pellets. Modelling showed this requires a pellet launch from the vessel inboard. Optimization can be achieved by the pellet parameters and the injection geometry; the latter however taking into account boundary conditions resulting from system integration needs. Design activities integrating the pellet transfer system into the vessel and the breeding blanket unveiled several possible variants requiring different levels of technical efforts. Basically, all extra efforts bear the benefit of a deeper and hence more favourable particle deposition. To quantify the potential gain, a full closed loop modelling was performed calculating the required pellet particle flux for any solution considered. Results allow now to balance potential advantages against related efforts required. Furthermore, the analysis tools developed can be employed for even more refined optimization of the pellet actuator tool by e.g. taking into account the interplay of pellet fuelling with burn control requirements.

For electricity producing tokamak fusion reactors like EU-DEMO, it is prudent to choose a plasma scenario close to the ITER baseline, where the largest amount of experimental evidence is available. Nevertheless, there are some aspects in which ITER and EU-DEMO have to differ, as the simple exercise of up-scaling from ITER to a larger device is constrained both by physical nonlinearities and by technological limits. In this work, relevant differences between ITER and the current EU-DEMO baseline in terms of plasma scenario are discussed. Firstly, EU-DEMO is assumed to operate with a very large amount of radiative power originating both from the scrape-off layer and, markedly, from the core. This radiation level is obtained by means of seeded impurities, whose presence significantly affects many aspects of the scenario itself, especially in terms of transient control. Secondly, because of the need of breeding tritium, the EU-DEMO wall is less robust than the ITER one. This implies that every off-normal interruption of the plasma discharge, for example in presence of a divertor reattachment, cannot rely on fast-shutdown procedures finally triggering a loss of plasma control at high current, but other strategies need to be developed. Thirdly, the ITER method for the control of the so-called sawteeth (ST) has been shown to be too expensive in terms of auxiliary power requirements, thus other solutions have to be explored. Finally, the problem of actively mitigating, or suppressing, the Edge Localised Modes (ELMs) has recently increased the interest on naturally ELM-free regimes (like QH-mode, I-mode, and also negative triangularity) for EU-DEMO, thus increasing the needs for ELM mitigation or suppression with respect to the approach adopted in ITER.

DEMO is a proposed demonstration fusion power plant which is currently being designed. Fusion power, Pfus, has to be controlled to stay above a minimum level to produce sufficient net electricity. However, this has an impact on the power through the separatrix, Psep, and thus can produce excessive heat flux to the divertor which can lead to damage. This gives a limit for the maximum allowable Psep. The aim of this work is to develop a framework which will give the possibility to design and to test controllers by simulating DEMO plasma transport. Therefore, we coupled the ASTRA transport modelling code for fusion devices with the Simulink simulation framework. ASTRA is equipped with equilibrium, transport, fuelling, heating and current drive modules. Simulink is a powerful tool to model and to simulate different dynamic systems. It allows fast and simple development of controllers using its built-in blocks. Therefore, coupling of ASTRA with Simulink gives the advantage of fast development of controllers for the power plant modelled with sophisticated physics based on the transport codes. This coupling was used to run the first basic simulations, where we simulated feedback control of fusion power, power through separatrix and pedestal top Greenwald density fraction, n GW ped top . n GW ped top is controlled via the pellet frequency to keep it below one. Fusion power is controlled with external heating (NBI). To prevent increasing Psep and heat load on the divertor while increasing Pfus, xenon gas can be puffed into the vessel. However, Psep has to be kept above a threshold to stay in H-mode. Therefore, feedback control of the xenon gas puff is modelled. These controllers were used to control fusion power, power through separatrix and pedestal top Greenwald density fraction with two different external disturbances. In the first one, we mimicked changes of the H-factor and in the second one we mimicked the fall of a tungsten flake. Results show that we can control these parameters already with a combination of simple controllers.

Recent I-mode investigations from the ASDEX Upgrade tokamak are reported. It is shown that neutral-beam-injection heated I-modes can be stationary, which is important in terms of extrapolability towards future fusion devices. Furthermore, detailed studies on the weakly coherent mode are reported. In particular, experimental observations point towards its existence in L-mode, before I-mode starts. Moreover, its impact on density and temperature fluctuations is evaluated. Studies of stationary divertor heat fluxes show that in I-mode, the upstream power fall-off length is between those observed in L-mode and H-mode, and it is connected to the scrape-off layer temperature fall-off length. Moreover, analysis of transient divertor heat loads shows that intermittent turbulent events, observed in the confinement region and linked to the weakly coherent mode, are responsible for a significant part of divertor heat loads. EUROfusion Consortium grant agreement No 633053 EUROfusion Enabling Research work-package AWP15-ENR-09/IPP-02

Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened. Peer Reviewed Article escrit per 365 autors/autores: Labit, B.; Eich, T.; Harrer, G. F.; Wolfrum, E.; Bernert, M.; Dunne, M. G.; Frassinetti, L.; Hennequin, P.; Maurizio, R.; Merle, A.; Meyer, H.; Saarelma, S.; Sheikh, U.; Adamek, J.; Agostini, M.; Aguiam, D.; Akers, R.; Albanese, R.; Albert, C.; Alessi, E.; Ambrosino, R.; Andr be, Y.; Angioni, C.; Apruzzese, G.; Aradi, M.; Arnichand, H.; Auriemma, F.; Avdeeva, G.; Ayllon-Guerola, J. M.; Bagnato, F.; Bandaru, V. K.; Barnes, M.; Barrera-Orte, L.; Bettini, P.; Bilato, R.; Biletskyi, O.; Bilkova, P.; Bin, W.; Blanchard, P.; Blanken, T.; Bobkov, V.; Bock, A.; Boeyaert, D.; Bogar, K.; Bogar, O.; Bohm, P.; Bolzonella, T.; Bombarda, F.; Boncagni, L.; Bouquey, F.; Bowman, C.; Brezinsek, S.; Brida, D.; Brunetti, D.; Bucalossi, J.; Buchanan, J.; Buermans, J.; Bufferand, H.; Buller, S.; Buratti, P.; Burckhart, A.; Calabr, G.; Calacci, L.; Camenen, Y.; Cannas, B.; Cano Megías, P.; Carnevale, D.; Carpanese, F.; Carr, M.; Carralero, D.; Carraro, L.; Casolari, A.; Cathey, A.; Causa, F.; Cavedon, M.; Cecconello, M.; Ceccuzzi, S.; Cerovsky, J.; Chapman, S.; Chmielewski, P.; Choi, D.; Cianfarani, C.; Ciraolo, G.; Coda, S.; Coelho, R.; Colas, L.; Colette, D.; Cordaro, L.; Cordella, F.; Costea, S.; Coster, D.; Cruz Zabala, D. J.; Cseh, G.; Czarnecka, A.; Cziegler, I.; D’Arcangelo, O.; Dal Molin, A.; David, P.; De Carolis, G.; De Oliveira, H.; Decker, J.; Dejarnac, R.; Delogu, R.; Den Harder, N.; Dimitrova, M.; Dolizy, F.; Domínguez-Palacios Durán, J. J.; Douai, D.; Drenik, A.; Dreval, M.; Dudson, B.; Dunai, D.; Duval, B. P.; Dux, R.; Elmore, S.; Embréus, O.; Erds, B.; Fable, E.; Faitsch, M.; Fanni, A.; Farnik, M.; Faust, I.; Faustin, J.; Fedorczak, N.; Felici, F.; Feng, S.; Feng, X.; Ferreira, J.; Ferr, G.; Février, O.; Ficker, O.; Figini, L.; Figueiredo, A.; Fil, A.; Fontana, M.; Francesco, M.; Fuchs, C.; Futatani, S.; Gabellieri, L.; Gadariya, D.; Gahle, D.; Galassi, D.; Gałązka, K.; Galdon-Quiroga, J.; Galeani, S.; Gallart, D.; Gallo, A.; Galperti, C.; Garavaglia, S.; Garcia, J.; Garcia-Lopez, J.; Garcia-Mu oz, M.; Garzotti, L.; Gath, J.; Geiger, B.; Giacomelli, L.; Giannone, L.; Gibson, S.; Gil, L.; Giovannozzi, E.; Giruzzi, G.; Gobbin, M.; Gonzalez-Martin, J.; Goodman, T. P.; Gorini, G.; Gospodarczyk, M.; Granucci, G.; Grekov, D. 1; Grenfell, G.; Griener, M.; Groth, M.; Grover, O.; Gruca, M.; Gude, A.; Guimarais, L.; Gyergyek, T.; Hacek, P.; Hakola, A.; Ham, C.; Happel, T.; Harrison, J.; Havranek, A.; Hawke, J.; Henderson, S.; Hesslow, L.; Hitzler, F.; Hnat, B.; Hobirk, J.; Hoelzl, M.; Hogeweij, D.; Hopf, C.; Hoppe, M.; Horacek, J.; Hron, M.; Huang, Z.; Iantchenko, A.; Iglesias, D.; Igochine, V.; Innocente, P.; Ionita-Schrittwieser, C.; Isliker, H.; Ivanova-Stanik, I.; Jacobsen, A.; Jakubowski, M.; Janky, F.; Jardin, A.; Jaulmes, F.; Jensen, T.; Jonsson, T.; Kallenbach, A.; Kappatou, A.; Karpushov, A.; Kasilov, S.; Kazakov, Y.; Kazantzidis, P. V.; Keeling, D.; Kelemen, M.; Kendl, A.; Kernbichler, W.; Kirk, A.; Kocsis, G.; Komm, M.; Kong, M.; Korovin, V.; Koubiti, M.; Kovacic, J.; Krawczyk, N.; Krieger, K.; Kripner, L.; Křivská, A.; Kudlacek, O.; Kulyk, Y.; Kurki-Suonio, T.; Kwiatkowski, R.; Laggner, F.; Laguardia, L.; Lahtinen, A.; Lang, P.; Likonen, J.; Lipschultz, B.; Liu, F.; Lombroni, R.; Lorenzini, R.; Loschiavo, V. P.; Lunt, T.; MacUsova, E.; Madsen, J.; Maggiora, R.; Maljaars, B.; Manas, P.; Mantica, P.; Mantsinen, M. J.; Manz, P.; Maraschek, M.; Marchenko, V.; Marchetto, C.; Mariani, A.; Marini, C.; Markovic, T.; Marrelli, L.; Martin, P.; Martín Solís, J. R.; Martitsch, A.; Mastrostefano, S.; Matos, F.; Matthews, G.; Mayoral, M.-L.; Mazon, D.; Mazzotta, C.; Mc Carthy, P.; McClements, K.; McDermott, R.; McMillan, B.; Meineri, C.; Menkovski, V.; Meshcheriakov, D.; Messmer, M.; Micheletti, D.; Milanesio, D.; Militello, F.; Miron, I. G.; Mlynar, J.; Moiseenko, V.; Molina Cabrera, P. A.; Morales, J.; Moret, J.-M.; Moro, A.; Moulton, D.; Nabais, F.; Naulin, V.; Naydenkova, D.; Nem, R. D.; Nespoli, F.; Newton, S.; Nielsen, A. H.; Nielsen, S. K.; Nikolaeva, V.; Nocente, M.; Nowak, S.; Oberkofler, M.; Ochoukov, R.; Ollus, P.; Olsen, J.; Omotani, J.; Ongena, J.; Orain, F.; Orsitto, F. P.; Paccagnella, R.; Palha, A.; Panaccione, L.; Panek, R.; Panjan, M.; Papp, G.; Paradela Perez, I.; Parra, F.; Passeri, M.; Pau, A.; Pautasso, G.; Pavlichenko, R.; Perek, A.; Pericoli Radolfini, V.; Pesamosca, F.; Peterka, M.; Petrzilka, V.; Piergotti, V.; Pigatto, L.; Piovesan, P.; Piron, C.; Piron, L.; Plyusnin, V.; Pokol, G.; Poli, E.; Pölöskei, P.; Popov, T.; Popovic, Z.; Pór, G.; Porte, L.; Pucella, G.; Puiatti, M. E.; Pütterich, T.; Rabinski, M.; Juul Rasmussen, J.; Rasmussen, J.; Rattá, G. A.; Ratynskaia, S.; Ravensbergen, T.; Réfy, D.; Reich, M.; Reimerdes, H.; Reimold, F.; Reiser, D.; Reux, C.; Reznik, S.; Ricci, D.; Rispoli, N.; Rivero-Rodriguez, J. F.; Rocchi, G.; Rodriguez-Ramos, M.; Romano, A.; Rosato, J.; Rubinacci, G.; Rubino, G.; Ryan, D. A.; Salewski, M.; Salmi, A.; Samaddar, D.; Sanchis-Sanchez, L.; Santos, J.; Särkimäki, K.; Sassano, M.; Sauter, O.; Scannell, R.; Scheffer, M.; Schneider, B. S.; Schneider, P.; Schrittwieser, R.; Schubert, M.; Seidl, J.; Seliunin, E.; Sharapov, S.; Sheeba, R. R.; Sias, G.; Sieglin, B.; Silva, C.; Sipilä, S.; Smith, S.; Snicker, A.; Solano, E. R.; Hansen, S. K.; Soria-Hoyo, C.; Sorokovoy, E.; Sozzi, C.; Sperduti, A.; Spizzo, G.; Spolaore, M.; Stejner, M.; Stipani, L.; Stober, J.; Strand, P.; Sun, H.; Suttrop, W.; Sytnykov, D.; Szepesi, T.; Tál, B.; Tala, T.; Tardini, G.; Tardocchi, M.; Teplukhina, A.; Terranova, D.; Testa, D.; Theiler, C.; Thorén, E.; Thornton, A.; Tilia, B.; Tolias, P.; Tomes, M.; Toscano-Jimenez, M.; Tsironis, C.; Tsui, C.; Tudisco, O.; Urban, J.; Valisa, M.; Vallar, M.; Vallejos Olivares, P.; Valovic, M.; Van Vugt, D.; Vanovac, B.; Varje, J.; Varju, J.; Varoutis, S. 1; Vartanian, S.; Vasilovici, O.; Vega, J.; Verdoolaege, G.; Verhaegh, K.; Vermare, L.; Vianello, N.; Vicente, J.; Viezzer, E.; Villone, F.; Voitsekhovitch, I.; Voltolina, D.; Vondracek, P.; Vu, N. M. T.; Walkden, N.; Wauters, T.; Weiland, M.; Weinzettl, V.; Wensing, M.; Wiesen, S.; Wiesenberger, M.; Wilkie, G.; Willensdorfer, M.; Wischmeier, M.; Wu, K.; Xiang, L.; Zagorski, R.; Zaloga, D.; Zanca, P.; Zaplotnik, R.; Zebrowski, J.; Zhang, W.; Zisis, A.; Zoletnik, S.; Zuin, M.

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).

Integrated edge-localized-mode (ELM)-free or small ELM scenarios for the demonstration fusion power plant (DEMO) are investigated in ASDEX Upgrade using argon seeding for radiative power removal mainly in the pedestal region. An important aspect is the modification of the electron pressure in the pedestal by the additional radiative power losses. Full ELM suppression could be achieved in a no-ELM H-mode scenario featuring an edge electromagnetic quasicoherent mode up to a heating power of 5 MW, where argon radiation allowed the extension of the heating power operational space. At higher powers up to 12 MW (reaching the beta limit), ELMs of reduced size prevail and detachment is obtained by the argon seeding. Control of the position of a radiating zone localized inside the X-point was found to be favorable compared to the control of the separatrix power for low P s e p / P h e a t . Integration of pedestal cooling with Ar and ELM suppression by resonant magnetic perturbations (RMP) allowed an increase of the core radiation and a partial recovery of normalized confinement to H98 = 1. This favorable behavior is finally limited by the loss of the RMP density pump-out effect, followed by an ELM-free H-mode phase and re-occurrence of ELMs. For an active tailoring of the pedestal pressure profile, precise knowledge of the radiation profile is required. Modeling of the argon radiation profile with the STRAHL code showed good agreement with bolometry only if charge exchange of neutral deuterium with argon ions was taken into account, highlighting the importance of the neutral density in the pedestal region. In view of best integration of a no-ELM scenario with divertor detachment and high heating power, the divertor compression and enrichment of argon and neon were compared using dynamic gas puff experiments. Argon shows more than a factor of 3 higher divertor enrichment compared to neon, but the absolute values decrease with higher neutral deuterium pressure.

One main programme topic at the ASDEX Upgrade all-metal-wall tokamak is development of a high-density regime with central densities at reactor grade level while retaining high-confinement properties. This required development of appropriate control techniques capable of coping with the pellet tool, a powerful means of fuelling but one which presented challenges to the control system for handling of related perturbations. Real-time density profile control was demonstrated, raising the core density well above the Greenwald density while retaining the edge density in order to avoid confinement losses. Recently, a new model-based approach was implemented that allows direct control of the central density. Investigations focussed first on the N-seeding scenario owing to its proven potential to yield confinement enhancements. Combining pellets and N seeding was found to improve the divertor buffering further and enhance the operational range accessible. For core densities up to about the Greenwald density, a clear improvement with respect to the non-seeding reference was achieved; however, at higher densities this benefit is reduced. This behaviour is attributed to recurrence of an outward shift of the edge density profile, resulting in a reduced peeling-ballooning stability. This is similar to the shift seen during strong gas puffing, which is required to prevent impurity influx in ASDEX Upgrade. First tests indicate that highly-shaped plasma configurations like the ITER base-line scenario, respond very well to pellet injection, showing efficient fuelling with no measurable impact on the edge density profile.

The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q95 = 5.5, ) at low density. Higher installed electron cyclotron resonance heating power 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG. Stable high-density H-modes with MW m−1 with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently . This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated. Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of and Er allow for inter ELM transport analysis confirming that Er is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of 'natural' no ELM regimes have been extended. Stable I-modes up to have been characterised using -feedback. Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle—measured for the first time—or the cross-phase angle of and fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvénic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 and 2019–2020 under grant agreement No. 633053. The views and opinions expressed herein do not necessarily reflect those of the European Commission. "Article signat per més de 100 autors/es: H. Meyer, for the AUG Team: D. Aguiam, C. Angioni, C.G. Albert, N. Arden, R. Arredondo Parra, O. Asunta, M. de Baar, M. Balden, V. Bandaru, K. Behler, A. Bergmann, J. Bernardo, M. Bernert, A. Biancalani, R. Bilato, G. Birkenmeier, T.C. Blanken, V. Bobkov, A. Bock, T. Bolzonella, A. Bortolon, B. Böswirth, C. Bottereau, A. Bottino, H. van den Brand, S. Brezinsek, D. Brida, F. Brochard, C. Bruhn, J. Buchanan, A. Buhler, A. Burckhart, Y. Camenen, D. Carlton, M. Carr, D. Carralero, C. Castaldo, M. Cavedon, C. Cazzaniga, S. Ceccuzzi, C. Challis, A. Chankin, S. Chapman, C. Cianfarani, F. Clairet, S. Coda, R. Coelho, J.W. Coenen, L. Colas, G.D. Conway, S. Costea, D.P. Coster, T.B. Cote, A. Creely, G. Croci, G. Cseh, A. Czarnecka, I. Cziegler, O. D'Arcangelo, P. David, C. Day, R. Delogu, P. de Marné, S.S. Denk, P. Denner, M. Dibon, A. Di Siena, D. Douai, A. Drenik, R. Drube, M. Dunne, B.P. Duval, R. Dux, T. Eich, S. Elgeti, K. Engelhardt, B. Erdös, I. Erofeev, B. Esposito, E. Fable, M. Faitsch, U. Fantz, H. Faugel, I. Faust, F. Felici, J. Ferreira, S. Fietz, A. Figuereido, R. Fischer, O. Ford, L. Frassinetti, S. Freethy, M. Fröschle, G. Fuchert, J.C. Fuchs, H. Fünfgelder, K. Galazka, J. Galdon-Quiroga, A. Gallo, Y. Gao, S. Garavaglia, A. Garcia-Carrasco, M. Garcia-Muñoz, B. Geiger, L. Giannone, L. Gil, E. Giovannozzi, C. Gleason-González, S. Glöggler, M. Gobbin, T. Görler, I. Gomez Ortiz, J. Gonzalez Martin, T. Goodman, G. Gorini, D. Gradic, A. Gräter, G. Granucci, H. Greuner, M. Griener, M. Groth, A. Gude, S. Günter, L. Guimarais, G. Haas, A.H. Hakola, C. Ham, T. Happel, N. den Harder, G.F. Harrer, J. Harrison, V. Hauer, T. Hayward-Schneider, C.C. Hegna, B. Heinemann, S. Heinzel, T. Hellsten, S. Henderson, P. Hennequin, A. Herrmann, M.F. Heyn, E. Heyn, F. Hitzler, J. Hobirk, K. Höfler, M. Hölzl, T. Höschen, J.H. Holm, C. Hopf, W.A. Hornsby, L. Horvath, A. Houben, A. Huber, V. Igochine, T. Ilkei, I. Ivanova-Stanik, W. Jacob, A.S. Jacobsen, F. Janky, A. Jansen van Vuuren, A. Jardin, F. Jaulmes, F. Jenko, T. Jensen, E. Joffrin, C.-P. Käsemann, A. Kallenbach, S. Kálvin, M. Kantor, A. Kappatou, O. Kardaun, J. Karhunen, S. Kasilov,, Y. Kazakov, W. Kernbichler, A. Kirk, S. Kjer Hansen, V. Klevarova, G. Kocsis, A. Köhn, M. Koubiti, K. Krieger, A. Krivska, A. Krämer-Flecken, O. Kudlacek, T. Kurki-Suonio, B. Kurzan, B. Labit, K. Lackner, F. Laggner, P.T. Lang, P. Lauber, A. Lebschy, N. Leuthold, M. Li, O. Linder, B. Lipschultz, F. Liu, Y. Liu, A. Lohs, Z. Lu, T. Luda di Cortemiglia, N.C. Luhmann, R. Lunsford, T. Lunt, A. Lyssoivan, T. Maceina, J. Madsen, R. Maggiora, H. Maier, O. Maj, J. Mailloux, R. Maingi, E. Maljaars, P. Manas, A. Mancini, A. Manhard, M.-E. Manso, P. Mantica, M. Mantsinen, P. Manz, M. Maraschek, C. Martens, P. Martin, L. Marrelli, A. Martitsch, M. Mayer, D. Mazon, P.J. McCarthy, R. McDermott, H. Meister, A. Medvedeva, R. Merkel, A. Merle, V. Mertens, D. Meshcheriakov, O. Meyer, J. Miettunen, D. Milanesio, F. Mink, A. Mlynek, F. Monaco, C. Moon, F. Nabais, A. Nemes-Czopf, G. Neu, R. Neu, A.H. Nielsen, S.K. Nielsen, V. Nikolaeva, M. Nocente, J.-M. Noterdaeme, I. Novikau, S. Nowak, M. Oberkofler, M. Oberparleiter, R. Ochoukov, T. Odstrcil, J. Olsen, F. Orain, F. Palermo, O. Pan, G. Papp, I. Paradela Perez, A. Pau, G. Pautasso, F. Penzel, P. Petersson, J. Pinzón Acosta, P. Piovesan, C. Piron, R. Pitts, U. Plank, B. Plaum, B. Ploeckl, V. Plyusnin, G. Pokol, E. Poli, L. Porte, S. Potzel, D. Prisiazhniuk, T. Pütterich, M. Ramisch, J. Rasmussen, G.A. Rattá, S. Ratynskaia, G. Raupp, G.L. Ravera, D. Réfy, M. Reich, F. Reimold, D. Reiser, T. Ribeiro, J. Riesch, R. Riedl, D. Rittich, J.F. Rivero-Rodriguez, G. Rocchi, M. Rodriguez-Ramos, V. Rohde, A. Ross1, M. Rott, M. Rubel, D. Ryan, F. Ryter, S. Saarelma, M. Salewski, A. Salmi, L. Sanchis-Sanchez, J. Santos, O. Sauter, A. Scarabosio, G. Schall, K. Schmid, O. Schmitz, P.A. Schneider, R. Schrittwieser, M. Schubert, T. Schwarz-Selinger, J. Schweinzer, B. Scott, T. Semer, E. Seliunin, M. Sertoli, A. Shabbir, A. Shalpegin, L. Shao, S. Sharapov, G. Sias, M. Siccinio, B. Sieglin, A. Sigalov, A. Silva, C. Silva, D. Silvagni, P. Simon, J. Simpson, E. Smigelskis, A. Snicker, C. Sommariva, C. Sozzi, M. Spolaore, A. Stegmeir, M. Stejner, J. Stober, U. Stroth, E. Strumberger, G. Suarez, H.-J. Sun, W. Suttrop, E. Sytova, T. Szepesi, B. Tál, T. Tala, G. Tardini, M. Tardocchi, M. Teschke, D. Terranova, W. Tierens, E. Thorén, D. Told, P. Tolias, O. Tudisco, W. Treutterer, E. Trier, M. Tripský, M. Valisa, M. Valovic, B. Vanovac, D. van Vugt, S. Varoutis, G. Verdoolaege, N. Vianello, J. Vicente, T. Vierle, E. Viezzer, U. von Toussaint, D. Wagner, N. Wang, X. Wang, M. Weiland, A.E. White, S. Wiesen, M. Willensdorfer, B. Wiringer, M. Wischmeier, R. Wolf, E. Wolfrum, L. Xiang, Q. Yang, Z. Yang, Q. Yu, R. Zagórski, I. Zammuto, W. Zhang, M. van Zeeland, T. Zehetbauer, M. Zilker, S. Zoletnik, H. Zohm and the EUROfusion MST1 Team55

The ITER first wall is designed for start-up and ramp-down in limiter configuration. The wall panels are toroidally shaped in order to spread the incident parallel power flux q|| uniformly, assuming a single decay length λq whose value is not known from first principles. In order to study the scaling of q|| with plasma parameters, infra-red viewing of specially-designed limiters has been used on the COMPASS tokamak in ∼100 discharges with scans in Ip, ne and for all combinations of magnetic field and Ip directions. The IR measurement clearly shows that in addition to the main SOL heat flux profile with λq > 40 mm, a steep gradient (λqnear = 4 ± 2 mm) dominates q|| near separatrix. This appears independently of limiter shaping, insertion with respect to neighbors and incident field-line angles. Good agreement is found between the measured λqnear and the prediction of a heuristic drift-based model.

The development of the plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) is facing unprecedented challenges. The DEMO D&C system has to operate with very high reliability, since any loss of plasma control may result in machine damage. On the same time, high accuracy of the D&C system is needed in order to allow for plasma operation near operational limits, where the fusion power is maximized. The implementation and performance of diagnostic components is however limited by space restrictions (optimization of the tritium breeding rate; integrity of the first wall and divertor against loads), and by adverse effects acting on the front-end components (neutron and gamma radiation, heat loads, erosion and deposition). Finally, the capabilities of the available actuators (poloidal field coils, external heating and fueling) are limited as well. As part of the European DEMO conceptual design studies, the development of the D&C system has recently been launched [1]. A preliminary suite of candidate diagnostics for DEMO have been selected, aiming to cover all the main plasma control quantities with some redundancy, and choosing types and locations for diagnostic front-end components such that long-term durability with minimum need for maintenance can be expected under typical loads (e.g. neutron radiation, particle fluxes and fluencies). Specifically, only robust metallic or ceramics diagnostic components shall be used inside the vacuum vessel, while the more sensitive components shall be located at more remote (protected) positions. This initial plasma diagnostic suite comprises microwave diagnostics, magnetic coil based and Hall sensors, passive spectroscopy and radiation power measurements, divertor thermo-current measurement, infrared interferometry/polarimetry and neutron-gamma measurements. In the first R&D phase the possibilities and conditions for integration of diagnostic sightlines and front-end components into the machine have been investigated, and an understanding of the required number of channels and components has been obtained. The conditions for controllability of the DEMO plasma are being analyzed by numerical simulations. To this purpose, the transport modelling code ASTRA, coupled to a radiation module, has been connected to the Simulink simulation framework, and the performance properties of diagnostics and actuators are being added [2]. Similarly, the existing CREATE equilibrium code is being amended to include limitations of diagnostics and actuators, such that the controllability e.g. of fast VDEs can be simulated under DEMO relevant settings. Furthermore, predictive control oriented models such as RAPTOR are being further developed as an alternative approach to analyze the controllability of DEMO. The common goal is to arrive at numerical simulations which closely mimic the control of the DEMO plasma such that the controllability based on the available diagnostics and actuators can be demonstrated quantitatively. Based on these R&D results, an initial version of the DEMO diagnostic and control concept has been elaborated and will be presented in this paper.

The Thomson scattering diagnostics (TS) at the COMPASS tokamak (operated in the Institute of Plasma Physics ASCR, Prague, Czech Republic) is based on two high-power Nd:YAG lasers, which have a pulse energy of 1.5 J and repetition rate of 30 Hz each. A new timing and synchronization unit (TSU) based on a Field Programmable Gate Array (FPGA) was designed and constructed in order to assure reliable and synchronized control of flash lamps and Q-switches of both lasers. Correct delay between laser pulses is necessary to obtain the highest possible laser energy and to prevent the active medium, the Nd:YAG rod, from thermal damage. An FPGA development kit was chosen as core hardware for the TSU and dedicated input and output circuits were manufactured to adapt the kit interfaces for COMPASS TS needs. In addition, the FPGA firmware, the control software, and the graphical user interface were developed to control the TSU hardware. This paper describes the TS regimes that are addressed by the new TSU. Further, the TSU design, the operation flow, and the firmware and software development are presented. Finally, the results of commissioning are shown.

This paper reports on probe measurements of the electron energy distribution function and plasma potential in the divertor region of the COMPASS tokamak during D-shaped plasmas. The probe data have been processed using the novel first-derivative technique. A comparison with the results obtained by processing the same data with the classical probe technique, which assumes Maxwellian electron energy distribution functions is presented and discussed. In the vicinity of the inner and outer strike points of the divertor the electron energy distribution function can be approximated by a bi-Maxwellian, with a dominating low-energy electron population (4-7 eV) and a minority of higher energy electrons (12-25 eV). In the private flux region between the two strike points the electron energy distribution function is found to be Maxwellian with temperatures in the range of 7-10 eV. The comparative analysis using both techniques has allowed a better insight into the underlying physical processes at the divertor region of the COMPASS tokamak. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

The COMPASS plasma control system is based on the MARTe real-time framework. Thanks to MARTe modularity and flexibility new algorithms have been developed for plasma diagnostic (plasma position calculation), control (shaping field control), and protection systems (central solenoid protection). Moreover, the MARTe framework itself was modified to broaden the communication capabilities via Aurora. This paper presents the recent upgrades and improvements made to the COMPASS real-time plasma control system, focusing on the issues related to precision of the real-time calculations, and discussing the improvements in terms of discharge parameters and stability. In particular, the new real-time system has given the possibility to analyze and to minimize the transport delays of each control loop.

This paper presents an overview of the Control, Data Acquisition, and Communication system (CODAC) at the COMPASS tokamak: the hardware set-up, software implementation, and communication tools are described. The diagnostics and the data acquisition are tailored for high spatial and temporal resolution required by the COMPASS physics programme, which aims namely at studies of the plasma edge, pedestal, and Scrape-off-Layer (SOL). Studies of instabilities and turbulence are also an integral part of the programme. Therefore, the data acquisition consists of more than 1000 channels, sampled at rates from 500 kS/s up to 2 GS/s. Presently, the feedback system controls the plasma position and shape, plasma current, and density and it includes 32 analogue input channels as well as 1 digital input/output channel and 8 analogue outputs. The feedback control runs within the Multi-threaded Application Real-Time executor (MARTe) framework with two threads, a 500 μs cycle to control slow systems and a 50 μs cycle to control the fast feedback power supplies for plasma position control. In this paper, special attention is paid to the links between the systems, to the hardware and software connections, and to the communication. The hardware part is described, the software framework is addressed, and the particular implementation – the dedicated software modules, communication protocols, and links to the database are described.

The main magnetic fields in COMPASS – i.e. The Toroidal, Magnetising, Equilibrium, and Shaping Fields – are created by a set of four corresponding thyristor power supplies controlled in a 0.5 ms loop. The plasma position has to be controlled both radially and vertically by two additional magnetic fields provided by two fast amplifiers (FAs) based on MOSFET technology, each supplying ±100 V and up to ±5 kA. Currently, an ongoing project aims at ELM triggering by fast changes of the vertical position of the plasma column, also referred to as vertical kicks. For this purpose, a new Vertical Kicks Power Supply (VKPS) capable of quick change of vertical plasma position is being constructed. This power supply should operate at up to 1.2 kV with switching frequency up to 5 kHz. It is designed as a H-bridge but based on IGBT transistors which can be operated at higher voltages than MOSFETs. In this paper, we focus on the FAs and VKPS engineering design and required output parameters. Both the power supplies are based on modern components with highest available ratings in their categories. Unique design of the power supplies takes advantage of the short duration of the COMPASS discharge by overloading the transistors above their maximal steady-state rating. The FA is regularly operating, so that in addition to describing its design, we also describe the achieved performance parameters. Finally, the common controller unit, communication, and error handling is described.

The COMPASS tokamak is expected to produce several gigabytes of data per shot in near future. A new storage system is needed to accommodate and access all the data. It should be scalable, fault-tolerant, and parallel. It should not be based on proprietary solutions to maintain independence from hardware and software manufacturers and preferably it should be built on inexpensive commodity hardware. One of the promising distributed parallel fault-tolerant file systems, GlusterFS, was selected for testing. The aim of the work was to make initial tests of a particular small GlusterFS setup to confirm its aptitude for the COMPASS storage system. Aggregated reading throughput from multiple NFS clients was one of the most important figures that were benchmarked, it scaled well with the number of clients, starting just above 60 MiB/s 1 for 1 client and going slightly over 300 MiB/s for 6 clients.

The COMPASS tokamak, a divertor device with clear H-mode and ITER-relevant geometry, has been re-installed in IPP Prague after its transport from CCFE in UK. The first plasma was achieved in December 2008. Many new diagnostic tools with both high temporal and spatial resolutions have been developed to address the scientific programme of COMPASS focused on H-mode physics and pedestal investigations. In the paper, an overview of existing and in near future planned diagnostics (magnetic, spectroscopy, microwave, probe and beam/particle) on COMPASS is given including their basic technical specifications and achievements.

An efficient horizontal and vertical stabilization of the plasma column position are essential for a reliable tokamak operation. Plasma position is generally determined by plasma current, plasma pressure and external vertical and horizontal magnetic fields. Such fields are generated by poloidal field coils and proper algorithm for the current control have to by applied, namely, in case of fast feedback loops. This paper presents a real-time plasma position reconstruction algorithms developed for the COMPASS tokamak. Further, its implementation in the MARTe (Multithreaded Application Real-Time executor) is described and the first results from test of the algorithm for real-time control of horizontal plasma positions are presented.

Modern fusion experiments require the presence of several subsystems, responsible for the different parameters involved in the operation of the machine. With the migration from the pre-programmed to the real-time control paradigm, their integration in Control, Data Acquisition, and Communication (CODAC) systems became an important issue, as this implies not only the connection to a main central coordination system, but also communications with related diagnostics and actuators. A subsystem for the control and operation of the vacuum, gas injection and baking was developed and installed in the COMPASS tokamak. These tasks are performed by dsPIC microcontrollers that receive commands from a hub computer and send information regarding the status of the operation. Communications are done in the serial protocol RS-232 through fibre optics. Java software, with an intuitive graphical user interface, for controlling and monitoring of the subsystem was developed and installed in a hub computer. In order to allow operators to perform these tasks remotely besides locally, this was integrated in the FireSignal system. Taking advantage of FireSignal features, it was possible to provide the users with, not only the same functionalities of the local application but also a similar user interface. An independent FireSignal Java Node bridges the central server and the control application. This design makes possible to easily reuse the Node for other subsystems or integrate the vacuum slow control in the other CODAC systems. The complete system, with local and remote control, has been installed successfully on COMPASS and has been in operation since April this year.

The COMPASS tokamak has started its operation recently in Prague and to meet the necessary operation parameters its real-time system, for data processing and control, must be designed for both flexibility and performance, allowing the easy integration of code from several developers and to guarantee the desired time cycle. For this purpose an Advanced Telecommunications Computing Architecture based real-time system has been deployed with a solution built on a multi-core x86 processor. It makes use of two software components: the BaseLib2 and the MARTe (Multithreaded Application Real-Time executor) real-time frameworks. The BaseLib2 framework is a generic real-time library with optimized objects for the implementation of real-time algorithms. This allowed to build a library of modules that process the acquired data and execute control algorithms. MARTe executes these modules in kernel space Real-Time Application Interface allowing to attain the required cycle time and a jitter of less than 1.5 μs. MARTe configuration and data storage are accomplished through a Java hardware client that connects to the FireSignal control and data acquisition software. This article details the implementation of the real-time system for the COMPASS tokamak, in particular the organization of the control code, the design and implementation of the communications with the actuators and how MARTe integrates with the FireSignal software.

The Compass tokamak is currently being installed in Prague, Czech Republic. Its control and data acquisition system is being redesigned and built from scratch and it will be based on novel digital technologies. The proposed hardware for the controller is based on the PICMG 3.0 Advanced Telecommunications Computing Architecture ™ (ATCA) standard. The platform contains one ATCA controller with a Gigabit Ethernet interface, up to 12 ATCA Digitizer-Generator-Processor (DGP) cards and trigger and clock inputs, all on a 12U shelf. The multi-core × 86-based General Purpose Processor (GPP) controller will be connected to the DGP cards by Peripheral Component Interconnect Express ™ (PCIe) point-to-point links through the ATCA backplane. Multiple-input multiple-output (MIMO) signal processing will be shared by the DGP cards using the built-in Field Programmable Gate Arrays (FPGAs) and the controller’s × 86 general processor. Eleven Aurora ™ 2.5 Gbit/s links allow to further parallelize the execution of code among several FPGAs. In order to guarantee real-time execution of the control codes a framework based on Linux and the Real-Time Application Interface (RTAI) will be used. This will explore the features provided by the new multi-core technologies. Synchronization between the subsystems will be guaranteed by a real-time event network. The interface to the system will be provided by the FireSignal control and data acquisition system. This will allow the operators and diagnostic coordinators to configure the hardware, prepare the discharges, pre-program events of interest and follow results from the discharge. FireSignal will also orchestrate the data flow coming from the different diagnostics into the database and to registered data clients.

The electron density on COMPASS is measured using 2 mm microwave interferometer. Interferometer signal is used as an input for the feedback control loop, running under the MARTe real-time framework. Two different threads are used to calculate (fast 50 μs thread) and to control (slow 500 μs thread) the electron density. The interferometer measures a line averaged density along a measurement chord. This paper describes an approach to control the line-averaged electron density in a real-time loop, using a correction to the real plasma shape, the plasma position, and non-linear effects of the electron density measurement at high densities. Newly developed real-time electron density control give COMPASS the chance to control the electron density more accurately which is essential for parametric scans for diagnosticians, for physics experiments and also for achieving plasma scenarios with H-mode.

The core Thomson scattering diagnostic (TS) on the COMPASS tokamak was put in operation and reported earlier. Implementation of edge TS, with spatial resolution along the laser beam up to ∼1/100 of the tokamak minor radius, is presented now. The procedure for spatial calibration and alignment of both core and edge systems is described. Several further upgrades of the TS system, like a triggering unit and piezo motor driven vacuum window shutter, are introduced as well. The edge TS system, together with the core TS, is now in routine operation and provides electron temperature and density profiles.

We present a complex data handling system for the COMPASS tokamak, operated by IPP ASCR Prague, Czech Republic [1] . The system, called CDB (COMPASS DataBase), integrates different data sources as an assortment of data acquisition hardware and software from different vendors is used. Based on widely available open source technologies wherever possible, CDB is vendor and platform independent and it can be easily scaled and distributed. The data is directly stored and retrieved using a standard NAS (Network Attached Storage), hence independent of the particular technology; the description of the data (the metadata) is recorded in a relational database. Database structure is general and enables the inclusion of multi-dimensional data signals in multiple revisions (no data is overwritten). This design is inherently distributed as the work is off-loaded to the clients. Both NAS and database can be implemented and optimized for fast local access as well as secure remote access. CDB is implemented in Python language; bindings for Java, C/C++, IDL and Matlab are provided. Independent data acquisitions systems as well as nodes managed by FireSignal [2] are all integrated using CDB. An automated data post-processing server is a part of CDB. Based on dependency rules, the server executes, in parallel if possible, prescribed post-processing tasks.

Recent results of measured fast neutrons created in the D-D reaction on the COMPASS tokamak during ohmic discharges are presented in this paper. Two different type detectors were used during experiment. He-3 detectors and bubble detectors as a support. The measurements are an introduction for neutron diagnostic on tokamak COMPASS and monitoring neutrons during discharges with Neutral Beam Injection (NBI). The He-3 counters and bubble detectors were located in two positions near tokamak vacuum chamber at a distance less than 40 cm to the centre of plasma. The neutrons flux was observed in ohmic discharges. However, analysis of our results does not indicate any clear source of neutrons production during ohmic discharges.

The contribution focuses on plasma tomography via the minimum Fisher regularisation (MFR) algorithm applied on data from the recently commissioned tomographic diagnostics on the COMPASS tokamak. The MFR expertise is based on previous applications at Joint European Torus (JET), as exemplified in a new case study of the plasma position analyses based on JET soft x-ray (SXR) tomographic reconstruction. Subsequent application of the MFR algorithm on COMPASS data from cameras with absolute extreme ultraviolet (AXUV) photodiodes disclosed a peaked radiating region near the limiter. Moreover, its time evolution indicates transient plasma edge cooling following a radial plasma shift. In the SXR data, MFR demonstrated that a high resolution plasma positioning independent of the magnetic diagnostics would be possible provided that a proper calibration of the cameras on an x-ray source is undertaken.

The scrape-off-layer (SOL) parameters in the COMPASS tokamak are studied by using a Langmuir probe mounted on a horizontal reciprocating manipulator. The radial profiles of the plasma potential, the electron energy distribution function and the electron densities are derived from the measured current-voltage probe characteristics by applying the firstderivative probe technique (FDPT). It is shown that close to the tokamak wall the electron energy distribution function is Maxwellian, while in the SOL, in the vicinity of the last closed flux surface and inside the confined plasma, the electron energy distribution function is bi-Maxwellian with a low-energy electron fraction dominating over a higher energy one. The radial profiles of the electron pressure and the parallel electron power flux density in COMPASS are also presented.

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.

The transition reported in L-mode plasmas leading to the formation of a density “shoulder” in the Scrape-off Layer (SOL) of tokamaks is investigated in the three ITER stepladder machines: COMPASS, ASDEX Upgrade (AUG) and JET. In AUG and JET, the broadening of the profiles can be observed at similar relative densities (ne/nGW ∼ 0.4–0.5) and coincides with a clear increase in the size of SOL filaments and the onset of the outer divertor detachment. In COMPASS, no transition is observed in the same density range: the divertor does not detach and no clear effect is observed neither in the profiles nor in the turbulence. This is explained as the result of an increase of intermittent transport triggered by the increased resistivity along filaments associated to divertor detachment. The transition conditions in AUG and JET are compared to foreseen ITER SOL parameters, and the implications are briefly discussed.

Real-time plasma electron density profile estimation and control are essential in the operation of future tokamaks. In particular, the robustness against diagnostics failure and disturbances is important for long pulse operation., A model-based approach to profile estimation is implemented on the control systems of ASDEX Upgrade and TCV, which is able to merge information from various diagnostics for both core and edge density, as well as systematically handling diagnostic failure. The model used for profile estimation is employed to tune a feedback controller before an experiment, thereby reducing the experimental time required for tuning., Subsequently, this observer and controller have been employed in scientific experiments on ASDEX Upgrade and TCV. On ASDEX Upgrade, the density profile estimator was used in high-density pellet-fuelled discharges, providing a more reliable real-time estimate of the core density for feedback control than previously achieved. On TCV, in experiments on integrated pressure and safety factor profile control, the density profile estimator and feedback controller provide a constant density despite disturbances from time-varying ECCD power. Additionally, the real-time density profiles provide an essential input for other real-time plasma state estimation codes including Electron Cyclotron ray tracing codes, contributing to a complete real-time estimation of the entire plasma state.

Fenix [1] , [2] is the ASDEX Upgrade (AUG) flight simulator based on the 1-D transport code ASTRA coupled with the 2-D equilibrium solver [ 3 ] SPIDER and Simulink (simulation platform integrated with MATLAB). Fenix is designed to simulate the entire discharge from the ramp up of the poloidal field (PF) coils, through the plasma phase, finishing with the ramp down of the PF coils. Currently, controllers for position, shape, fuelling and heating are implemented according to the AUG control system setup. Also, a model of the poloidal field coils and the models of their corresponding power supplies are provided, as well as basic models of heating actuators such as electron cyclotron resonant heating (ECRH), neutral beam injection (NBI) and fuelling using either pellets or gas. As a next step, the Fenix simulations have to be validated against an already existing database of the discharges. This article compares the behaviour of Fenix simulations against various AUG AUGAUGAUG discharge measurements. It shows that it succeeds in its goal as a comprehensive tokamak discharge simulator.