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Core particle fueling, an essential task in the European demonstration fusion power plant EU-DEMO, relies on adequate pellet injection. However, pellets are fragile objects, and their delivery efficiency can hardly be assumed to be unity. Exploring kinetic control of the EU-DEMO1 scenario indicates that such missed-out pellets do cause a considerable problem for keeping a burning plasma. Missed-out pellets can cause a severe drop of plasma density that in turn results in a potential drastic loss of burn power. Efforts are under way at the ASDEX Upgrade (AUG) tokamak aiming to provide real-time monitoring of pellet arrival and announcement of missed-out cases to the control systems. To further optimize the controllers, system identification experiments have been performed to identify the dynamic response of the system to the actuators.

The recently developed integrated model based on engineering parameters (IMEP) (Luda et al 2020 Nucl. Fusion 61 126048; Luda et al 2021 Nucl. Fusion 60 036023), so far validated on ASDEX Upgrade, has been tested on a database of 3 Alcator C-Mod and 55 JET-ILW ELMy (type I) H-mode stationary phases. The empirical pedestal transport model included in IMEP, consisting now of imposing a fixed value of R < ∇ T e > / T e , t o p = − 82.5 , allows an accurate prediction of the pedestal top temperature (when the pedestal top density is fixed to the experimental measurements) across these three machines with different sizes, when the pedestal is peeling–ballooning (PB) limited. Cases far from the ideal PB boundary, corresponding to high edge Spitzer resistivity, are instead strongly overpredicted by IMEP. A comparison between the predictions of Europed and IMEP for a subset of JET-ILW cases shows that IMEP can more accurately reproduce the experimental pedestal width. This allows IMEP to better capture profile effects on the pedestal stability, and therefore to correctly describe the negative effect of fueling on the pedestal pressure for PB limited cases. A strong correlation between the separatrix density and the fueling rate has been identified for a subset of JET-ILW cases, when taking into account different divertor configurations. Overall, these promising results encourage further developments of integrated models to obtain reliable predictions of pedestal and global confinement using only engineering parameters for present and future machines.

In this work, a very fast integrated transport model involving every region that interacts directly with the plasma of a tokamak, has been developed. The confined region is modeled in 1.5D, while the scrape-off layer has a 0D structure. For the core region, a physics-based analytical regression based on a set of simulations with the transport model TGLF [Staebler 2005 Phys. Plasmas 12 102508] has been produced. For the H-mode regime, an average edge-localized-modes model is applied in the pedestal region. In the scrape-off layer a two-point model for electron temperature (exhaust) and a particle balance for the species density at the separatrix have been implemented. All the models have first been validated individually in a standalone setting. Finally, six fully integrated simulations of an L-mode discharge, and five H-mode discharges, have been performed in the Fenix flight simulator [Janky et al 2019 Fusion Eng. Des. 146 1926, Fable et al 2022 Plasma Phys. Control. Fusion 64 044002], including transients, matching the experimental trajectories of an ASDEX upgrade discharge during flat-top and ramp-down. A broader validation including more discharges and the ramp-up phase is planned for the near future.

EU-DEMO will be the next step in Europe after ITER on the path toward a fusion power plant. The matter injection systems have to provide the requested material in order to establish, maintain, and terminate the burning plasma. Their main function is to fuel the plasma, but other tasks are addressed as well like delivering matter for generating sufficient core radiation and divertor buffering. In the preconceptual design phase performed from 2014 to 2020, the matter injection systems, in particular pellet injection and gas injection, have been assessed. This work describes the main findings and state of the art of the matter injection systems at the transition from the preconceptual design phase to the conceptual design phase.

Fusion systems codes (SCs) are deployed to produce the baseline of the European fusion power reactor (DEMO) within its conceptual design. A DEMO baseline is mostly defined by a radial/vertical reactor sketch and major reactor parameters, such as fusion and net electric power, magnetic fields, and plasma burn time. A baseline shall also meet a set of prescribed reactor requirements, constraints, and architectural features. According to the conceptual design workflow implemented within the EU-DEMO programme, the output from the SC is transferred to the detailed physics and engineering design codes. Presently-available fusion SCs rely on rather basic physics and engineering models (mostly at zero or one-dimensional level). The design codes, instead, are very detailed but run on much longer computing times. To fill the gap between systems and design codes, the multi-fidelity systems/design tool modular integrated reactor analysis (MIRA)—has been recently developed. MIRA incorporates the physics and the engineering insights of the utmost domains of tokamak reactors and relies on a higher spatial resolution, spanning from 1D up to 3D modelling frames. The MIRA approach has been applied to the DEMO 2017 baseline, generated by the EU reference SC PROCESS and used as input to MIRA. In the paper, the architectural and mathematical insights of the MIRA package are described, along with an EU-DEMO 2017 baseline analysis.

The prediction of plasma rotation is of high interest for fusion research due to the effects of the rotation upon magnetohydrodynamic (MHD) instabilities, impurities, and turbulent transport in general. In this work, an analysis method was studied and validated to reliably extract momentum transport coefficients from neutral beam injection (NBI) modulation experiments. To this end, a set of discharges was created with similar background profiles for the ion and electron temperatures, the heat fluxes, the electron density, and the plasma rotation that, therefore, should exhibit similar momentum transport coefficients. In these discharges, a range of temporal perturbations were imposed by modulating and varying the power deposition of the NBI, electron-cyclotron-resonance heating (ECRH), and ion-cyclotron-resonance heating (ICRH). The transport model including diffusion, convection, and residual stress was implemented within the ASTRA code. The Prandtl number P r = χ φ / χ i was assessed via the GKW code. A convective Coriolis pinch was fitted and the intrinsic torque from the residual stress was estimated. The obtained transport coefficients agree within error bars for sufficiently small imposed temperature perturbations, as would be expected, from the similar background profiles. This successful validation of the methodology opens the door to study the parametric dependence of the diffusive and convective momentum transport of the main ions of the plasma as well as the turbulent intrinsic torque in a future work.

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.

The successful validation of theory-based models of transport, magnetohydrodynamic stability, heating and current drive, with tokamak measurements over the last 20 years, has laid the foundation for a new era where these models can be routinely used in a ‘predict first’ approach to design and predict the outcomes of experiments on tokamaks today. The capability to predict the plasma confinement and core profiles with a quantified uncertainty, based on a multi-machine, international, database of experience, will provide confidence that a proposed discharge will remain within the operational limits of the tokamak. Developing this predictive capability for the first generation of burning plasma devices, beginning with ITER, and progressing to tokamak demonstration reactors, is a critical mission of fusion energy research. Major advances have been made implementing this predict first methodology on today’s tokamaks. An overview of several of these recent advances will be presented, providing the integrated modeling foundations of the experimental successes. The first steps to include boundary plasmas, and tokamak control systems, have been made. A commitment to predicting experiments as part of the planning process is needed in order to collect predictive accuracy data and evolve the models and software into a robust whole discharge pulse design simulator.

We present recent highlights from the most recent operation phases of Wendelstein 7-X, the most advanced stellarator in the world. Stable detachment with good particle exhaust, low impurity content, and energy confinement times exceeding 100 ms, have been maintained for tens of seconds. Pellet fueling allows for plasma phases with reduced ion-temperature-gradient turbulence, and during such phases, the overall confinement is so good (energy confinement times often exceeding 200 ms) that the attained density and temperature profiles would not have been possible in less optimized devices, since they would have had neoclassical transport losses exceeding the heating applied in W7-X. This provides proof that the reduction of neoclassical transport through magnetic field optimization is successful. W7-X plasmas generally show good impurity screening and high plasma purity, but there is evidence of longer impurity confinement times during turbulence-suppressed phases. 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. Peer Reviewed Article signat per 497 autors/es: Thomas Sunn Pedersen1,2,∗ , I. Abramovic3, P. Agostinetti4, M. Agredano Torres1, S. Äkäslompolo1, J. Alcuson Belloso1, P. Aleynikov1, K. Aleynikova1, M. Alhashimi1, A. Ali1, N. Allen5, A. Alonso6, G. Anda7, T. Andreeva1, C. Angioni8, A. Arkhipov8, A. Arnold1, W. Asad8, E. Ascasibar6, M.-H. Aumeunier9, K. Avramidis10, E. Aymerich11, S.-G. Baek3, J. Bähner1, A. Baillod12, M. Balden1, M. Balden8, J. Baldzuhn1, S. Ballinger3, M. Banduch1, S. Bannmann1, A. Banon Navarro8, A. Bañon Navarro ´ 1, T. Barbui13, C. Beidler1, C. Belafdil9, A. Bencze7, A. Benndorf1, M. Beurskens1, C. Biedermann1, O. Biletskyi14, B. Blackwell15, M. Blatzheim1, T. Bluhm1, D. Böckenhoff1, G. Bongiovi16, M. Borchardt1, D. Borodin17, J. Boscary8, H. Bosch1,18, T. Bosmann19, B. Böswirth8, L. Böttger1, A. Bottino8, S. Bozhenkov1, R. Brakel1, C. Brandt1, T. Bräuer1, H. Braune1, S. Brezinsek17, K. Brunner1, S. Buller1, R. Burhenn1, R. Bussiahn1, B. Buttenschön1, A. Buzás7, V. Bykov1, I. Calvo6, K. Camacho Mata1, I. Caminal20, B. Cannas11, A. Cappa6, A. Carls1, F. Carovani1, M. Carr21, D. Carralero6, B. Carvalho22, J. Casas20, D. Castano-Bardawil17, F. Castejon6, N. Chaudhary1, I. Chelis23, A. Chomiczewska24, J.W. Coenen13,17, M. Cole1, F. Cordella25, Y. Corre9, K. Crombe26, G. Cseh7, B. Csillag7, H. Damm1, C. Day10, M. de Baar27, E. De la Cal6, S. Degenkolbe1, A. Demby13, S. Denk3, C. Dhard1, A. Di Siena8,28, A. Dinklage12, T. Dittmar17, M. Dreval14, M. Drevlak1, P. Drewelow1, P. Drews17, D. Dunai7, E. Edlund3, F. Effenberg29, G. Ehrke1, M. Endler1, D.A. Ennis5, F.J. Escoto6, T. Estrada6, E. Fable8, N. Fahrenkamp1, A. Fanni11, J. Faustin1, J. Fellinger1, Y. Feng1, W. Figacz4, E. Flom13, O. Ford1, T. Fornal24, H. Frerichs13, S. Freundt1, G. Fuchert1, M. Fukuyama30, F. Füllenbach1, G. Gantenbein10, Y. Gao1, K. Garcia13, J.M. García Regaña6, I. García-Cortés6, J. Gaspar31, D.A. Gates29, J. Geiger1, B. Geiger13, L. Giudicotti32, A. González6, A. Goriaev26,33, D. Gradic1, M. Grahl1, J.P. Graves12, J. Green13, E. Grelier9, H. Greuner8, S. Groß1, H. Grote1, M. Groth34, M. Gruca24, O. Grulke1,35, M. Grün1, J. Guerrero Arnaiz1, S. Günter8, V. Haak1, M. Haas1, P. Hacker1, A. Hakola36, A. Hallenbert1, K. Hammond29, X. Han17,37, S.K. Hansen3, J.H. Harris38, H. Hartfuß1, D. Hartmann1, D. Hathiramani1, R. Hatzky8, J. Hawke39, S. Hegedus7, B. Hein8, B. Heinemann8, P. Helander12, S. Henneberg1, U. Hergenhahn8,40, C. Hidalgo6, F. Hindenlang8, M. Hirsch1, U. Höfel1, K.P. Hollfeld17, A. Holtz1, D. Hopf8, D. Höschen17, M. Houry9, J. Howard19, X. Huang41, M. Hubeny17, S. Hudson29, K. Ida9, Y. Igitkhanov10, V. Igochine8, S. Illy10, C. Ionita-Schrittwieser42, M. Isobe39, M. Jabłczynska ´ 24, S. Jablonski24, B. Jagielski1, M. Jakubowski1, A. Jansen van Vuuren1, J. Jelonnek10, F. Jenko8, F. Jenko8, T. Jensen35, H. Jenzsch1, P. Junghanns8, J. Kaczmarczyk24, J. Kallmeyer1, U. Kamionka1, M. Kandler8, S. Kasilov43, Y. Kazakov26, D. Kennedy1, A. Kharwandikar1, M. Khokhlov1, C. Kiefer8, C. Killer1, A. Kirschner17, R. Kleiber1, T. Klinger12, S. Klose1, J. Knauer1, A. Knieps17, F. Köchl44, G. Kocsis7, Ya.I. Kolesnichenko45, A. Könies1, R. König1, J. Kontula34, P. Kornejew1, J. Koschinsky, M.M. Kozulia14, A. Krämer-Flecken17, R. Krampitz1, M. Krause1, N. Krawczyk24, T. Kremeyer1, L. Krier10, D.M. Kriete5, M. Krychowiak1, I. Ksiazek46, M. Kubkowska24, M. Kuczynski1, G. Kühner1, A. Kumar15, T. Kurki-Suonio34, S. Kwak1, M. Landreman47, P.T. Lang8, A. Langenberg1, H.P. Laqua12, H. Laqua1, R. Laube1, S. Lazerson1, M. Lewerentz1, C. Li17, Y. Liang17, Ch. Linsmeier17, J. Lion1, A. Litnovsky17,48, S. Liu37, J. Lobsien1, J. Loizu12, J. Lore38, A. Lorenz1, U. Losada6, F. Louche26, R. Lunsford29, V. Lutsenko45, M. Machielsen12, F. Mackel8, J. Maisano-Brown3, O. Maj8, D. Makowski49, G. Manduchi50, E. Maragkoudakis6, O. Marchuk17, S. Marsen1, E. Martines4, J. Martinez-Fernandez6, M. Marushchenko1, S. Masuzaki41, D. Maurer5, M. Mayer8, K.J. McCarthy6, O. Mccormack4, P. McNeely1, H. Meister8, B. Mendelevitch8, S. Mendes1, A. Merlo1, A. Messian26, A. Mielczarek49, O. Mishchenko1, B. Missal1, R. Mitteau9, V.E. Moiseenko14, A. Mollen1, V. Moncada9, T. Mönnich1, T. Morisaki41, D. Moseev1, G. Motojima41, S. Mulas6, M. Mulsow1, M. Nagel1, D. Naujoks1, V. Naulin35, T. Neelis19, H. Neilson29, R. Neu8, O. Neubauer17, U. Neuner1, D. Nicolai17, S.K. Nielsen35, H. Niemann1, T. Nishiza1, T. Nishizawa1, T. Nishizawa8, C. Nührenberg1, R. Ochoukov8, J. Oelmann17, G. Offermanns17 K. Ogawa41, S. Okamura41, J. Ölmanns17, J. Ongena26, J. Oosterbeek1, M. Otte1, N. Pablant29, N. Panadero Alvarez6, N. Panadero Alvarez6, A. Pandey1, E. Pasch1, R. Pavlichenko14, A. Pavone1, E. Pawelec46, G. Pechstein1, G. Pelka24, V. Perseo1, B. Peterson41, D. Pilopp1, S. Pingel1, F. Pisano11, B. Plöckl8, G. Plunk1, P. Pölöskei1, B. Pompe2, A. Popov51, M. Porkolab3, J. Proll19, M.J. Pueschel19,27, M.-E. Puiatti52, A. Puig Sitjes1, F. Purps1, K. Rahbarnia1, M. Rasinski ´ 17, J. Rasmussen35, A. Reiman29, F. Reimold1, M. Reisner8, D. Reiter17, M. Richou9, R. Riedl8, J. Riemann1, K. Riße1, G. Roberg-Clark1, V. Rohde8, J. Romazanov17, D. Rondeshagen1, P. Rong1, L. Rudischhauser1, T. Rummel1, K. Rummel1, A. Runov1, N. Rust1, L. Ryc24, P. Salembier20, M. Salewski35, E. Sanchez6, S. Satake41, G. Satheeswaran17, J. Schacht1, E. Scharff1, F. Schauer8, J. Schilling1, G. Schlisio1, K. Schmid8, J. Schmitt5, O. Schmitz13, W. Schneider1, M. Schneider1, P. Schneider8, R. Schrittwieser42, T. Schröder1, M. Schröder1, R. Schroeder1, B. Schweer26, D. Schwörer1, E. Scott1, E. Scott8, B. Shanahan1, G. Sias11, P. Sichta29, M. Singer1, P. Sinha29, S. Sipliä34, C. Slaby1, M. Sleczka53, H. Smith1, J. Smoniewski54, E. Sonnendrücker8, M. Spolaore4, A. Spring1, R. Stadler8, D. Stanczak24, T. Stange1, I. Stepanov26, L. Stephey13, J. Stober8, U. Stroth8,55, E. Strumberger8, C. Suzuki41, Y. Suzuki41, J. Svensson1, T. Szabolics7, T. Szepesi7, M. Szücs7, F.L. Tabares6, N. Tamura41, A. Tancetti35, C. Tantos10, J. Terry3, H. Thienpondt6, H. Thomsen1, M. Thumm10, J.M. Travere9, P. Traverso5, J. Tretter8, E. Trier8, H. Trimino Mora1, T. Tsujimura41, Y. Turkin1, A. Tykhyi45, B. Unterberg17, P. van Eeten1, B.Ph. van Milligen6, M. van Schoor26, L. Vano1, S. Varoutis10, M. Vecsei7, L. Vela56, J.L. Velasco6, M. Vervier17, N. Vianello50, H. Viebke1, R. Vilbrandt1, G. Vogel8, N. Vogt1, C. Volkhausen1, A. von Stechow1, F. Wagner1, E. Wang17, H. Wang57, F. Warmer1, T. Wauters26, L. Wegener1, T. Wegner1, G. Weir1, U. Wenzel1, A. White3, F. Wilde1, F. Wilms1, T. Windisch1, M. Winkler1, A. Winter1, V. Winters1, R. Wolf118, A.M. Wright29, G.A. Wurden39, P. Xanthopoulos1, S. Xu17, H. Yamada58, H. Yamaguchi41, M. Yokoyama41, M. Yoshinuma41, Q. Yu8, M. Zamanov14, M. Zanini1, M. Zarnstorff29, D. Zhang1, S. Zhou17, J. Zhu1, C. Zhu29, M. Zilker8, A. Zocco1, H. Zohm8, S. Zoletnik7 and L. Zsuga7 // 1 Max Planck Institute for Plasma Physics, Garching and Greifswald, Germany: 2 University of Greifswald, Greifswald, Germany; 3 Massachusetts Institute of Technology, 77 Massachusetts Ave, Cambridge, MA 02139, United States of America; 4 Consorzio RFX, Corso Stati Uniti, 4-35127 Padova, Italy; 5 Auburn University, Auburn, AL 36849, United States of America; 6 CIEMAT, Avenida Complutense, 40, 28040 Madrid, Spain; 7 Center for Energy Research, Konkoly-Thegeut 29-33, 1121 Budapest, Hungary; 8 Max-Planck-Institute for Plasma Physics, Boltzmannstraße 2, 85748 Garching bei München, Germany; 9 CEA Cadarache, 13115 Saint-Paul-lez-Durance, France; 10 Karlsruhe Institute of Technology, Kaiserstr. 12, 76131 Karlsruhe, Germany; 11 University of Cagliari, Via Universita, 40, 09124 Cagliari, Italy; 12 École Polytechnique Fédérale de Lausanne, Swiss Plasma Center, CH-1015 Lausanne, Switzerland; 13 University of Wisconsin–Madison, Engineering Drive, Madison, WI 53706, United States of America; 14 Institute of Plasma Physics, National Science Center ‘Kharkiv Institute of Physics and Technology’, Kharkiv, Ukraine; 15 The Australian National University, Acton ACT 2601, Canberra, Australia; 16 Department of Engineering, University of Palermo, Viale delle Scienze, Edificio 6, Palermo, 90128, Italy; 17 Forschungszentrum Jülich GmbH, Institut für Energie-und Klimaforschung—Plasmaphysik, 52425 Jülich, Germany; 18 Technical University of Berlin, Strasse des 17. Juni 135, 10623 Berlin, Germany; 19 Eindhoven University of Technology, 5600 MB Eindhoven, Netherlands; 20 Universitat Politècnica de Catalunya. BarcelonaTech, C. Jordi Girona, 31, 08034 Barcelona, Spain; 21 Culham Center for Fusion Energy, Abingdon OX14 3EB, United Kingdom; 22 Instituto de Plasmas e Fusao Nuclear, Av. Rovisco Pais, 1049-001 Lisboa, Portugal; 23 Department of Physics, National and Kapodistrian University of Athens, 15784 Athens, Greece; 24 Institute of Plasma Physics and Laser Microfusion, 23 Hery Str., 01-497 Warsaw, Poland; 25 ENEA—Centro Ricerche Frascati, Via Enrico Fermi, 45, 00044 Frascati RM, Italy; 26 Laboratory for Plasma Physics, LPP-ERM/KMS, TEC Partner, B-1000 Brussels, Belgium; 27 Dutch Institute for Fundamental Energy Research, PO Box 6336, 5600 HH Eindhoven, Netherlands; 28 University of Texas, Austin, TX, United States of America; 29 Princeton Plasma Physics Laboratory, Princeton, NJ 08543, United States of America; 30 Kyushu University, 744 Motooka Nishi-ku, Fukuoka 819-0395, Japan; 31 Aix-Marseille University, Jardin du Pharo, 58 Boulevard Charles Livon, 13007, Marseille, France; 32 Department of Physics and Astronomy, Padova University, Via Marzolo 8, 35131 Padova, Italy; 33 Department of Applied Physics, Ghent University, Sint-Pietersnieuwstraat 41 B4, 9000 Ghent, Belgium; 34 Aalto University, 02150 Espoo, Finland; 35 Department of Physics, Technical University of Denmark, Anker Engelunds Vej, 2800 Kgs Lyngby, Denmark; 36 VTT Technical Research Center of Finland Ltd., PO Box 1000, FI-02044 VTT, Finland; 37 Institute of Plasma Physics, Chinese Academy of Sciences, 230031 Hefei, Anhui, China; 38 Oak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN 37830, United States of America; 39 Los Alamos National Laboratory, NM 87545, United States of America; 40 Fritz-Haber-Institut der Max-Planck-Gesellschaft, 14195 Berlin, Germany; 41 National Institute for Fusion Science, National Institutes of Natural Sciences, 322-6 Oroshi-cho, Toki, Gifu Prefecture 509-5292, Japan; 42 Institute for Ion Physics and Applied Physics, University of Innsbruck, Innsbruck, Austria; 43 Graz University of Technology, Rechbauerstraße 12, 8010 GRAZ, Austria; 44 Austrian Academy of Science, Doktor-Ignaz-Seipel-Platz 2, 1010 Wien, Austria; 45 Institute for Nuclear Research, prospekt Nauky 47, Kyiv 03028, Ukraine; 46 University of Opole, plac Kopernika 11a, 45-001 Opole, Poland; 47 University of Maryland, Paint Branch Drive, College Park, MA 20742, United States of America; 48 National Research Nuclear University MEPhI, 115409 Moscow, Russian Federation; 49 Department of Microelectronics and Computer Science, Lodz University of Technology, Wolczanska 221/223, 90-924 Lodz, Poland; 50 Consiglio Nazionale delle Ricerche, Piazzale Aldo Moro, 7, 00185 Roma, Italy; 51 Ioffe Physical-Technical Institute of the Russian Academy of Sciences, 26 Politekhnicheskaya, St Petersburg 194021, Russian Federation; 52 Istituto di Fisica del Plasma Piero Caldirola, Via Roberto Cozzi, 53, 20125 Milano, Italy; 53 University of Szczecin, 70-453, aleja Papieza Jana Pawła II 22A, Szczecin, Poland; 54 Lawrence University, 711 E Boldt Way, Appleton, WI 54911, United States of America; 55 Physik-Department E28, Technische Universität München, 85747 Garching, Germany; 56 Universidad Carlos III de Madrid, Av. de la Universidad, 30 Madrid, Spain; 57 Yale University, New Haven, CT 06520, United States of America; 58 University of Tokyo, 5-1-5 Kashiwanoha, Kashiwa, Chhiab 277-0882, Japan Objectius de Desenvolupament Sostenible::7 - Energia Assequible i No Contaminant Objectius de Desenvolupament Sostenible::7 - Energia Assequible i No Contaminant::7.a - Per a 2030, augmentar la cooperació internacional per tal de facilitar l’accés a la investigació i a les tecnolo­gies energètiques no contaminants, incloses les fonts d’energia renovables, l’eficiència energètica i les tecnologies de combustibles fòssils avançades i menys contaminants, i promoure la inversió en infraestructures energètiques i tecnologies d’energia no contaminant

Gas puff modulation experiments are performed at ASDEX Upgrade in L-mode plasmas. We model the discharge with the ASTRA transport code in order to determine transport coefficients outside of a normalized radius of ρ pol = 0.95. The experimental data is consistent with a range of particle diffusivities and pinch velocities of the order of D = (0.20 ± 0.13) m2 s−1 and v = (−1 ± 2) m s−1, respectively. The electron temperature response caused by the gas modulation permits to estimate also that heat diffusivity χ e increases almost linearly when collisionality rises due to fuelling. The fuelling particle flux is amplified by recycling, overcompensating losses.

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 this work we present the extensive validation of a refined version of the integrated model based on engineering parameters (IMEP) introduced in reference (Luda et al 2020 Nucl. Fusion 60 036023). The modeling workflow is now fully automated, computationally faster thanks to the reduced radial resolution of the TGLF calculation, and it includes the modeling of the toroidal rotation, which was still taken from experimental measurements in our previous work. The updated model maintains the same accuracy as its previous version when tested on the cases presented in the initial publication. The confined plasma, from the magnetic axis to the separatrix, is simulated without using any experimental information from profiles measurements, and the inputs of IMEP are the same engineering parameters used when programming a plasma discharge. The model validation database consists of 50 ASDEX Upgrade (AUG) stationary (over a few energy confinement time) H-mode phases, which largely cover the entire AUG operational domain. The prediction of IMEP is compared with experimental measurements and with scaling laws, such as the IPB98(y,2), the ITPA20-IL, and AUG specific regressions. This modeling framework has proven to be very accurate over the entire set of 50 cases, with a significantly lower mean relative error with respect to each of the scaling laws considered, accurately reproducing the change in pedestal and core confinement caused by a change in plasma current, heating power, fueling rate, triangularity, magnetic field, NBI voltage (i.e. the effect of a change in the core particle source), and heating mix (e.g. correctly predicting the effect on confinement caused by a change in T e/T i). Plasma confinement is correctly described by IMEP also for two particular operating regimes, such as the ITER baseline scenario, and the QCE regime (quasi continuous exhaust, also referred as type-II and small ELMs). This work clearly demonstrates the power of this approach in pulling out physics mechanisms to interpret subtle interdependencies and that a 1D integrated model can reproduce experimental results over very large parameter variations with a higher accuracy than any statistical regression. This approach has therefore the potential to improve the prediction of the fusion performance in future tokamak reactors.

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.

The optimum conditions for access to and sustainment of H-mode plasmas and their expected plasma parameters in the pre-fusion power operation 1 (PFPO-1) phase of the ITER research plan, where the additional plasma heating will be provided by 20 MW of electron cyclotron heating, are assessed in order to identify key open R&D issues. The assessment is performed on the basis of empirical and physics-based scalings derived from present experiments and integrated modelling of these plasmas including a range of first-principle transport models for the core plasma. The predictions of the integrated modelling of ITER H-mode plasmas are compared with ITER-relevant experiments carried out at JET (low-collisionality high-current H modes) and ASDEX Upgrade (significant electron heating) for both global H-mode properties and scale lengths of density and temperature profiles finding reasonable agreement. Specific integration issues of the PFPO-1 H-mode plasma scenarios are discussed taking into account the impact of the specificities of the ITER tokamak design (level of ripple, etc).

Kinetic profile predictions of ITER PFPO-1 plasmas require high accuracy in the central electron temperatures to be applied to the calculation of third harmonic electron cyclotron absorption. Correctly predicting the transition from L-mode to H-mode further requires precise estimates of the ion heat flux in the periphery of the plasma. Recent versions of the quasi-linear transport models TGLF and QuaLiKiz were tested against an extensive set of experimental results from ASDEX Upgrade (AUG) and JET-ILW, where the focus was put on AUG plasmas heated by ECRH. Spectra obtained from TGLF are compared to a set of linear gyrokinetic simulations performed with GKW. Electron and ion temperature profiles obtained with TGLF-SAT1geo show good agreement with the experimental profiles, but there is a slight tendency to underpredict central T e and T i at high ratios T e/T i. QuaLiKiz yields reasonable results for T e and T i profiles in plasmas where the ion temperature gradient mode is dominant, but predicts a significantly too weak transport in the presence of dominant trapped electron modes in conditions of strong central electron heating.

Within the current EU pre-conceptual design activities, the reference demonstrative fusion power plant concept relies on ITER-like, pulsed H-mode plasmas (“DEMO1”). Simultaneously, alternative design options are being considered. An example is “Flexi-DEMO”, which is designed for steady-state plasma discharges, but can also be run pulsed at the same fusion power with the same auxiliary heating systems if the plasma confinement time turns out to be not sufficient for non-inductive operations. Preliminary investigations have started also for ELM-free regimes, such as I-mode or QH-mode, due to the issues encountered in ELMs active control at reactor scales. Neutral beam injection (NBI) is one of the candidates for a DEMO heating and current-drive system. Depending on the system design, NBI can be optimized for e.g. central heating (as for DEMO1), current-drive (as for Flexi-DEMO) or even to provide torque to the plasma (which may favour the access to ELM-free QH-mode regime). In the present contribution we describe the main NBI features in terms of design parameters (NBI energy, injection geometry) that would optimize these alternative DEMO plasma scenarios, in particular regarding Flexi-DEMO and QH-mode regime. Wide-range parametric scans through METIS numerical simulations of DEMO plasmas indicate the optimal design windows for NBI system in each plasma configuration. These results need however further validations against technical constraints, available technology limits and detailed plasma transport simulations.

The common way to predict energy confinement in future devices such as ITER is to use scaling laws, based on parameters regression of large cross-machine databases. However, this approach is limited: the regression variables are not purely engineering parameters, physics quantities such as the plasma density n e are also input; power regressions fail to capture important physics such as regime transitions; profile effects such as T i/T e or reverse magnetic shear are not retained. As a consequence, the scatter is large, but even some dependences are known to be of limited validity, such as the n e or P dependences of the IPB98(y, 2) scaling. Dimensionless physics, if validated, provides a reliable basis for predicting confinement in a future device. However, there are no experimental data to use as boundary condition, nor empirical knowledge based on that device. Also, the use of full gyro-kinetic codes with all relevant ingredients is computationally unfeasible for a full-radius modelling. Integrated modelling with the recently developed IMEP workflow allows to simulate from the separatrix to the plasma centre, without making use of any experimental input. Although physics-based models are used to predict core transport, core transport is shown to be predicted with less accuracy than the pedestal energy. Therefore, we validate the most established quasi-linear models TGLF and QuaLiKiZ in different regimes and experimental conditions, assessing their strengths and limits. The quasi-linear models are shown to predict transport accurately in a variety of experimental conditions, including dominant ion and electron heating, respectively. The well-known shortcoming near the pedestal top is solved, while a new collision operator in the QuaLiKiZ model improves the electron temperature simulation significantly. Also the Neural Network version of QuaLiKiZ, with very fast computation time, appears to be well applicable in ion heated H-modes.

A new method was developed to model the neutral population produced by the gas puff based charge exchange recombination spectroscopy systems at ASDEX Upgrade (AUG). With this method, the edge impurity density on the high field side (HFS) and low field side (LFS) can be obtained without the need to apply a neutral beam injection system. The neutral penetration needed for the calculation of the impurity density is obtained with a new gas puff module implemented in the FIDASIM code. The LFS impurity density profile evaluated with the new gas puff module matches the impurity density calculated with standard beam-based charge exchange diagnostics. Impurity temperature, rotation and density profiles at the HFS and LFS of an AUG H-mode discharge are presented. Edge impurity toroidal and poloidal flows show asymmetric structures. The impurity density asymmetries obtained with the new gas puff module are consistent with the observed flow structure.

The key result of the present work is the theoretical prediction and observation of the formation of a new type of transport barrier in fusion plasmas, called F-ATB (fast ion-induced anomalous transport barrier). As demonstrated through state-of-the-art global electrostatic and electromagnetic simulations, the F-ATB is characterized by a full suppression of the turbulent transport - caused by strongly sheared, axisymmetric $E \times B$ flows - and an increase of the neoclassical counterpart, albeit keeping the overall fluxes at significantly reduced levels. The trigger mechanism is shown to be a mainly electrostatic resonant interaction between supra-thermal particles, generated via ion-cyclotron-resonance heating, and plasma micro-turbulence. These findings are obtained by realistic simulations of the ASDEX Upgrade discharge $\#36637$ - properly designed to maximized the beneficial role of the wave-particle resonance interaction - which exhibits the expected properties of improved confinement produced by energetic particles., Version 2

One of the three goals of the ITER project is to demonstrate fusion gain Q ≥ 5 in steady-state operation (SSO). A reassessment of the necessary conditions for Q ≥ 5 SSO in ITER has been carried out for steady-state scenarios assuming no internal transport barrier in the core plasma and with the sole use of neutral beam injection (NBI) and electron cyclotron heating and current drive (EC H&CD) as heating and current drive sources in these scenarios. The parametric operational space for SSO in ITER has been reassessed utilizing an inverse evaluation approach that takes into account the baseline design of the NBI and EC H&CD systems with PNBI = 33 MW, PEC = 20 MW and their possible upgrade capabilities included in their design, PNBI ≤ 49.5 MW, and PEC ≤ 30 MW. The optimal operational points from this evaluation approach have been chosen for detailed 1.5-D transport modelling by ASTRA and followed-up by magnetohydrodynamic (MHD) stability analysis to demonstrate the conditions in which the Q = 5 SS goal can be achieved in ITER where the plasma current is fully accounted for by the driven current by NBI and EC H&CD and the self-driven bootstrap current. The ideal MHD stability of plasma configurations with self-consistently simulated plasma profiles and equilibrium for the chosen OPs has been analysed by the KINX code. Using this analysis, the possibility to control the MHD stability of these steady-state plasmas by tailoring the current profile with the flexibility provided by design of the systems for electron cyclotron current drive and neutral beam current drive in ITER has been demonstrated. Issues related to the realization of such scenarios from the point of view of plasma physics, experimental demonstration and design limits of the ITER systems and components is discussed.

The electron cyclotron resonance heating (ECRH) system of the ASDEX Upgrade tokomak has been upgraded over the last 15 years from a 2 MW, 2 s, 140 GHz system to an 8 MW, 10 s, dual frequency system (105/140 GHz). The power exceeds the L/H power threshold by at least a factor of two, even for high densities, and roughly equals the installed ion cyclotron range of frequencies power. The power of both wave heating systems together (>10 MW in the plasma) is about half of the available neutral beam injection (NBI) power, allowing significant variations of torque input, of the shape of the heating profile and of Qe/Qi, even at high heating power. For applications at a low magnetic field an X3-heating scheme is routinely in use. Such a scenario is now also forseen for ITER to study the first H-modes at one third of the full field. This versatile system allows one to address important issues fundamental to a fusion reactor: H-mode operation with dominant electron heating, accessing low collisionalities in full metal devices (also related to suppression of edge localized modes with resonant magnetic perturbations), influence of Te/Ti and rotational shear on transport, and dependence of impurity accumulation on heating profiles. Experiments on all these subjects have been carried out over the last few years and will be presented in this contribution. The adjustable localized current drive capability of ECRH allows dedicated variations of the shape of the q-profile and the study of their influence on non-inductive tokamak operation (so far at q 95 > 5.3). The ultimate goal of these experiments is to use the experimental findings to refine theoretical models such that they allow a reliable design of operational schemes for reactor size devices. In this respect, recent studies comparing a quasi-linear approach (TGLF) with fully non-linear modeling (GENE) of non-inductive high-beta plasmas will be reported.

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.

We present the first successful simulation of a induced disruption in ASDEX Upgrade from massive material injection (MMI) up to established runaway electron (RE) beam, thus covering pre-thermal quench, thermal quench and current quench (CQ) of the discharge. For future high-current fusion devices such as ITER, the successful suppression of REs through MMI is of critical importance to ensure the structural integrity of the vessel. To computationally study the interplay between MMI, background plasma response, and RE generation, a toolkit based on the 1.5D transport code coupling ASTRA-STRAHL is developed. Electron runaway is described by state-of-the-art reduced kinetic models in the presence of partially ionized impurities. Applied to argon MMI in ASDEX Upgrade discharge #33108, key plasma parameters measured experimentally, such as temporal evolution of the line averaged electron density, plasma current decay rate and post-CQ RE current, are well reproduced by the simulation presented. Impurity ions are transported into the central plasma by the combined effect of neoclassical processes and additional effects prescribed inside the $q = 2$ rational surface to explain experimental time scales. Thus, a thermal collapse is induced through strong impurity radiation, giving rise to a substantial RE population as observed experimentally.

Experiments have been performed in the ASDEX Upgrade tokamak where plasmas are kept just below the H-mode threshold using feedback on the radiated power via seeded impurity. The resulting high-power L-modes show high confinement properties and no ELM activity due to the reduced pedestal height. The physics of these high-confinement L-modes is explored by means of integrated modeling to clarify the origin of the improved confinement.

One of the main issues in view of the realization of a DEMOnstration fusion reactor is the availability of a sufficient external supply of tritium (T) to start operation. T is an unstable nuclide, which is almost absent in nature and is currently available as by-product in e.g. CANDU, whose operation in the next decades (both in terms of life extension of existing reactors and construction of new ones) is at the moment under debate. During DEMO operation, T will be generated on-site by breeding blanket, employing the neutrons originating from D-T reaction. However, it is considered that a certain initial amount of T is needed to start operation, the so-called start-up inventory. An alternative approach consists of obtaining the start-up inventory exploiting reactions occurring in a D -D plasma, which generate T both directly in the plasma and via breeding in the breeding blanket. In the present paper, the conditions under which D -D start-up becomes a favorable option for a power plant are discussed. The analysis mainly focuses on the EU-DEMO reactor concept, for which design data are sufficient for a fairly quantitative evaluation of the relevant parameters. It is found that the unavoidable presence of elements requiring saturation before they are able to release significant amounts of T clamps the T production rate to the same order of magnitude as D -D reaction rate. Thus, under very optimistic assumptions, several hundreds of full-current D -D discharges are necessary for T to be available for plasma fueling, but more realistic estimates let this number raise up to several thousands.

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.

The main elements defining the tungsten content of an ITER plasma, i.e. the gross erosion, the prompt redeposition and the transport in the H-mode edge transport barrier are reviewed and it is investigated, whether the low tungsten concentration limit in the core calls for additional control of the divertor parameters or the ELM energy beyond the requirements set by the power handling capability of the divertor. The outcome is rather positive, since no further control seems to be required mainly because of the favourable transport in the H-mode edge transport barrier which provokes hollow tungsten profiles in the edge transport barrier for the expected pedestal parameters. Keywords: First wall, Tungsten, Erosion, ELM, Cross field transport, ITER