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Funding Information: 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. The support from the Spanish Ministry of Science (Grant No. FPU15/06074) is gratefully acknowledged. The mega and ascot simulations reported herein were performed on the MARCONI cluster under the MEGAFILD project. | openaire: EC/H2020/633053/EU//EUROfusion The suppression and excitation of Alfvén eigenmodes have been experimentally obtained, for the first time, by means of externally applied 3D perturbative fields with different spatial spectra in a tokamak plasma. The applied perturbation causes an internal fast-ion redistribution that modifies the phase-space gradients responsible for driving the modes, determining, ultimately their existence. Hybrid kinetic-magnetohydrodynamic simulations reveal an edge resonant transport layer activated by the 3D perturbative field as the responsible mechanism for the fast-ion redistribution. The results presented here may help to control fast-ion driven Alfvénic instabilities in future burning plasmas with a significant fusion born alpha particle population.

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.

A baffled snowflake minus low-field side (SF-LFS) is geometrically-optimised in tokamak à configuration variable, increasing divertor neutral pressure, to evaluate the roles of divertor closure (comparing with an unbaffled SF-LFS) and magnetic geometry (comparing with a baffled single null (SN)) in power exhaust and core-divertor compatibility. Ohmically-heated L-mode discharges in deuterium, with a line-averaged core density of approximately 4.7 × 10 19 m−3, are seeded with nitrogen to approach detached conditions. Baffles in the SF-LFS configuration are found to reduce the peak outer target heat flux by up to 23 % , without significantly affecting the location of the inter-null radiation region or the core-divertor compatibility. When compared to the baffled SN, the baffled SF-LFS exhibits a reduction in the outer target heat flux by up to 66 % and the ability to balance the strike-point distribution of heat flux. These benefits are less significant with N2 seeding, with similar peak target quantities (such as heat flux, electron temperature and ion flux) and divertor radiated power. Despite a radiating region located farther from the confined plasma for the SF-LFS than the baffled SN, no change in core confinement is observed. Core effective charge even indicates an increase in core impurity penetration for the SF-LFS. These experiments constitute a good reference for detailed model validations and extrapolations, exploring important physics such as core impurity shielding and the dependence of divertor cross-field transport on magnetic geometry.

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 tokamak à configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019–20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T e/T i ∼ 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with ‘small’ (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019–20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

Strongly radiating and detached high-power discharges in present-day full-metal tokamaks have a characteristic radiation pattern involving condensation of radiation near the X-point, with significant radiative losses above the X-point. In contrast, Demonstration Fusion Power Plant (DEMO) divertor exhaust scoping studies using reduced physics models, including a fluid description for the neutrals, place the strongest radiation fronts in the divertor legs, near the separatrix. The present contribution studies sensitivity of the radiation pattern corresponding to maximal divertor impurity radiation to those physics models that are typically neglected in the simulations due to their computational expense: cross-field drifts, complex impurity models and kinetic neutrals. Model benchmarking is carried out in comparison to L-mode discharges, which are shown to feature both divertor and X-point radiation. The simulated plasma conditions with maximal divertor radiation have in–out asymmetries in the divertor legs and at the divertor entrance, and the asymmetries and the radiation patterns are observed to be sensitive to both cross-field drift effects and the neutral model. DEMO simulations, carried out using SOLPS-ITER, show an impact of cross-field drifts on the divertor asymmetries, but the impact is not large enough to move the radiation front from the divertor legs to regions above the X-point.

Experiments in ASDEX Upgrade (AUG) and JET with the ITER-like wall (JET-ILW) are performed to separate the pedestal and core contributions to confinement in H-modes with different main ion masses. A strong isotope mass dependence in the pedestal is found which is enhanced at high gas puffing. This is because the ELM type changes when going from D to H for matched engineering parameters, which is likely due to differences in the inter ELM transport with isotope mass. The pedestal can be matched in H and D plasmas by varying only the triangularity and keeping the engineering parameters relevant for core transport the same. With matched pedestals Astra/TGLF (Sat1geo) core transport simulations predict the experimental profiles equally well for H and D. These core transport simulations show a negligible mass dependence and no gyro-Bohm scaling is observed. However, to match the experimental observations at medium β it is required to take the fast-ion dilution and rotation into account. This is not enough for high β plasmas where for the first time a profile match between H and D plasmas was achieved experimentally. Under these conditions quasilinear modelling with TGLF over predicts the transport in the core of H and D plasmas alike.

In some conditions, I-mode plasmas can feature pedestal relaxation events (PREs) that transiently enhance the energy reaching the divertor target plates. To shed light on their appearance, characteristics and energy reaching the divertor targets, a comparative study between two tokamaks — Alcator C-Mod and ASDEX Upgrade — is carried out. It is found that PREs appear only in a subset of I-mode discharges, mainly when the plasma is close to the H-mode transition. Also, a growing oscillating precursor before the PRE onset is observed in the region close to the separatrix in both devices, and a discussion on a possible triggering mechanism is outlined. The PRE relative energy loss from the confined region is found to increase with decreasing pedestal top collisionality ν ped * . Similarly, also the relative electron temperature drop at the pedestal top, which is related to the conductive energy loss, rises with decreasing ν ped * . Based on these relations, the PRE relative energy loss in future devices such as DEMO and ARC is estimated. Finally, the divertor peak energy fluence due to the PRE is measured on each device. Those values are then compared to the model introduced in Eich et al (2017 Nucl. Mater. Energy 12 84–90) for type-I edge localized modes. The model is shown to provide an upper boundary for PRE energy fluence data, while a lower boundary is found by dividing the model by three. These two boundaries are used to make projections of the PRE divertor energy fluence to DEMO and ARC.

The impact of externally applied magnetic perturbations (MPs) on fast-ion losses has been investigated by means of the light ion beam probe (LIBP) technique in the ASDEX Upgrade tokamak. The LIBP technique allows to experimentally infer the fast-ion orbit displacement induced by MPs via first-orbit losses using scintillator based fast-ion loss detector (FILD) measurements. The fast-ion orbit displacement against different applied MP spectra has been studied. These shots were conducted in ELM mitigated H-mode plasmas. A rigid rotation of the MP coils was applied with a frequency of 1 Hz, with an n = 2 configuration and changing the differential phase between the upper and lower set of coils (ΔΦul) on a shot-to-shot basis. Beam sources Q7 (tangential) and Q8 (radial) were used to probe different fast-ion orbits with FILD1. The measured fast-ion orbit displacement ranges from 3 to 20 mm approximately, and no qualitative difference is observed between ions from beam sources Q7 and Q8. The minimum is found for a ΔΦul ∼ 50°, which is shifted with respect to the minimum of the plasma boundary displacement, found at ΔΦul ∼ 0°. A first attempt to validate the orbit following code ASCOT—including the plasma response calculated with the MARS-F code—against these experimental measurements is performed. While the dependence of the first-orbit fast-ion displacement with ΔΦul does not match the experimental measurements, these simulations do capture other features such as the order of magnitude of the orbit displacement and the importance of the toroidal spectrum of the applied perturbation.

A model for simulating charge exchange (CX) of fast ions with background atoms in magnetically confined fusion plasmas has been implemented in the ASCOT orbit-following code. The model was verified by comparing simulated reaction mean free paths to analytical values across a range of fusion-relevant parameters. ASCOT was used to simulate beam ions slowing down in the presence of CX reactions in a MAST-U target scenario. ASCOT predicts the CX-induced loss of beam power to be 22 % , which agrees to within 15 % with the TRANSP prediction. Due to CX, plasma heating and current drive by beam ions are strongly reduced towards the edge. However, an overall lower but noticeable increase of up to 20 % in current drive is predicted closer to the core. The simulated deposition of fast CX atoms on the wall is concentrated around the outer midplane, with estimated peak power loads of 70–80 kW m−2 on the central poloidal field coils (P5) and the vacuum vessel wall between them. This analysis demonstrates that ASCOT can be used to simulate fast ions in fusion plasmas where CX reactions play a significant role, e.g. in spherical tokamaks and stellarators.

The tokamak a configuration variable (TCV) continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to address critical issues in preparation for ITER and a fusion power plant. For the 2019-20 campaign its configurational flexibility has been enhanced with the installation of removable divertor gas baffles, its diagnostic capabilities with an extensive set of upgrades and its heating systems with new dual frequency gyrotrons. The gas baffles reduce coupling between the divertor and the main chamber and allow for detailed investigations on the role of fuelling in general and, together with upgraded boundary diagnostics, test divertor and edge models in particular. The increased heating capabilities broaden the operational regime to include T (e)/T (i) similar to 1 and have stimulated refocussing studies from L-mode to H-mode across a range of research topics. ITER baseline parameters were reached in type-I ELMy H-modes and alternative regimes with `small' (or no) ELMs explored. Most prominently, negative triangularity was investigated in detail and confirmed as an attractive scenario with H-mode level core confinement but an L-mode edge. Emphasis was also placed on control, where an increased number of observers, actuators and control solutions became available and are now integrated into a generic control framework as will be needed in future devices. The quantity and quality of results of the 2019-20 TCV campaign are a testament to its successful integration within the European research effort alongside a vibrant domestic programme and international collaborations.

MAST-U has recently started operating with a Super-X divertor, designed to increase total flux expansion and neutral trapping, both predicted through simple analytic models and SOLPS calculations to reduce the plasma and impurity density detachment thresholds. In this study, utilising the SOLPS-ITER code, we are quantifying the possible gain allowed by the MAST-U Super-X and neutral baffling geometry, in terms of access to detachment. We show that a significant reduction of the upstream density detachment threshold (up to a factor 1.6) could be achieved in MAST-U, for the Super-X, as opposed to conventional divertor geometry, mainly through an increased total flux expansion, neutral trapping being found very similar between the different configurations. We also show that variations of the strike-point angle are complex to interpret in such a tightly baffled geometry, and that a case in which the target normal points more towards the separatrix does not necessarily imply a lower detachment threshold. As in previous calculations for TCV, we quantify the role of neutral effects through developing and applying a quantitative definition of neutral trapping.

This work presents the use of a collisional-radiative model to infer plasma parameters from 2D emissivities of several deuterium Balmer lines. The emissivities were obtained by MANTIS, an absolutely calibrated, ten-camera imaging polychromator with < 5 m m spatial resolution, up to 800 Hz frame rate, viewing the lower divertor tangentially. Our analysis of those image frames generates 2D maps of plasma parameters such as electron density, temperature, neutral atomic density and the reaction rates for ionisation, recombination and charge exchange as a function of time. The analysis is compared and validated against a SOLPS-ITER simulation accounting for drifts. The results are compared against the simulation in 2D, radial and poloidal profiles to probe the numerous effects of particle transport in the scrape-off layer, in particular in the approach to detached divertor leg conditions. The inferred inner divertor leg radial profiles of the electron density and temperature were consistent with the SOLPS-ITER predictions. A significant transport of particles to the private flux region is found experimentally, that is not captured in the simulation. The simulation diverges from the experiment at the outer divertor target, where the plasma emission appears to be consistent with the emission driven by plasma–molecule interactions. Our analysis also shows prospects for aiding the power exhaust control efforts by potentially providing an optical tracking of the particle balance in the divertor.

Experiments on nitrogen assisted divertor detachment in the improved energy confinement mode (I-mode) are reported from the ASDEX Upgrade tokamak. When nitrogen is introduced into the divertor and radiation losses cool the divertor plasma down, a loss of core confinement is observed, concomitant with an increase in low frequency edge fluctuation levels. The loss in confinement can be compensated and the I-mode can be maintained by additional heating power input. Detachment of the inner divertor leg has been observed for the first time in an I-mode discharge. The outer divertor leg remains attached in these experiments. Good energy confinement properties (H 98(y, 2) = 0.9) during the detachment of the inner divertor leg are reported.

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.

The data collected during ASDEX Upgrade experiments in which external 3D fields have been deployed in the attempt of mitigating runaway electrons (RE) are interpreted by a numerical test particle approach. To this end the Hamiltonian guiding center code ORBIT has been used, with the implementation of the magnetic perturbation spectrum modeled by the code MARS-F, which also takes into account the plasma response to the applied 3D fields. In agreement with the observed phenomenology, ORBIT simulations show that the configuration of the currents in the top/bottom arrays of error field coils, which maximizes the plasma response to the external perturbations, is the one that most affects the high energy test electron trajectories in the edge region, thus leading to an enhancement of the energetic electron losses. This occurs in particular during the disruption, i.e. taking into account the increased toroidal electric field associated with the fast plasma cooling. Used in a predictive way, the numerical results suggest which coil configuration could further improve the RE mitigation.

The multi-machine International Tokamak Physics Activity (ITPA) Global H-mode Confinement Database has been upgraded with new data from JET with the ITER-like wall and ASDEX Upgrade with the full tungsten wall. This paper describes the new database and presents results of regression analysis to estimate the global energy confinement scaling in H-mode plasmas using a standard power law. Various subsets of the database are considered, focusing on type of wall and divertor materials, confinement regime (all H-modes, ELMy H or ELM-free) and ITER-like constraints. Apart from ordinary least squares (OLS), two other, robust regression techniques are applied, which take into account uncertainty on all variables. Regression on data from individual devices shows that, generally, the confinement dependence on density and the power degradation are weakest in the fully metallic devices. Using the multi-machine scalings, predictions are made of the confinement time in a standard ELMy H-mode scenario in ITER. The uncertainty on the scaling parameters is discussed with a view to practically useful error bars on the parameters and predictions. One of the derived scalings for ELMy H-modes on an ITER-like subset is studied in particular and compared to the IPB98(y, 2) confinement scaling in engineering and dimensionless form. Transformation of this new scaling from engineering variables to dimensionless quantities is shown to result in large error bars on the dimensionless scaling. Regression analysis in the space of dimensionless variables is therefore proposed as an alternative, yielding acceptable estimates for the dimensionless scaling. The new scaling, which is dimensionally correct within the uncertainties, suggests that some dependencies of confinement in the multi-machine database can be reconciled with parameter scans in individual devices. This includes vanishingly small dependence of confinement on line-averaged density and normalized plasma pressure (β), as well as a noticeable, positive dependence on effective atomic mass and plasma triangularity. Extrapolation of this scaling to ITER yields a somewhat lower confinement time compared to the IPB98(y, 2) prediction, possibly related to the considerably weaker dependence on major radius in the new scaling (slightly above linear). Further studies are needed to compare more flexible regression models with the power law used here. In addition, data from more devices concerning possible ‘hidden variables’ could help to determine their influence on confinement, while adding data in sparsely populated areas of the parameter space may contribute to further disentangling some of the global confinement dependencies in tokamak plasmas.

openaire: EC/H2020/633053/EU//EUROfusion Gross and net erosion of tungsten (W) and other plasma-facing materials in the divertor region have been investigated in deuterium (D) and helium (He) plasmas during dedicated experiments in L- and H-mode on ASDEX Upgrade and after full-length experimental campaigns on the WEST tokamak. Net erosion was determined via post-exposure analyses of plasma-exposed samples and full-size wall components, and we conclude that the same approach is applicable to gross erosion if marker structures with sub-millimeter dimensions are used to eliminate the contribution of prompt re-deposition. In H-mode plasmas, gross erosion during ELMs may exceed the situation in inter-ELM conditions by 1-2 orders of magnitude while net erosion is typically higher by a factor of 2-3. The largest impact on net erosion is attributed to the electron temperature while the role of the impurity mixtures is weaker, even though both on ASDEX Upgrade and WEST significant amounts of impurities are present in the edge plasmas. Impurities, on the other hand, will lead to the formation of thick co-deposited layers. We have also noted that with increasing surface roughness, net erosion is strongly suppressed and the growth of co-deposited layers is enhanced. In He plasmas, gross erosion is increased compared to D due to the higher mass and charge states of the plasma particles, resulting from larger energies due to sheath acceleration, but strong impurity fluxes can result in apparent net deposition in the divertor. Our results from ASDEX Upgrade and WEST are comparable and indicate typical net-erosion rates of 0.1-0.4 nm s(-1), excluding the immediate vicinity of the strike-point regions.

The Tokamak à Configuration Variable (TCV) tokamak is in the midst of an upgrade to further its capability to investigate conventional and alternative divertor configurations. To that end, modular and removable gas baffles have been installed to decrease the coupling between the divertor and the plasma core. The baffles primarily seek to suppress the transit of recycling neutrals to closed flux surfaces. A first experimental campaign with the gas baffles has shown that the baffled divertor remains compatible with a wide range of configurations including snowflake and super-X divertors. Plasma density ramp experiments reveal an increase of the neutral pressure in the divertor by up to a factor ×5 compared to the unbaffled divertor and thereby qualitatively confirm simulations with the SOLPS-ITER code that were used to guide the baffle design. Together with a range of new and upgraded divertor diagnostics, the baffled TCV divertor is now used to validate divertor models for ITER and next step devices with particular emphasis on geometric variations.

The ITPA multi-machine confinement database has been recently extended, in particular to account for the effect of the metallic plasma facing components installed in the ASDEX Upgrade and JET tokamaks during the last decade. Confinement data from the ASDEX Upgrade tokamak have been contributed to the ITPA confinement database since 1994. The present version, which includes data from 1994 to 2017, is described and discussed in this article. The aim of this work is to document the database before it is made public and report on the main results on the confinement properties reflected in this database. Through the transition from carbon to metallic wall, the operation at low density originally accessible with the carbon wall, has been strongly limited by the occurrence of tungsten accumulation. However, at medium and high densities, the confinement properties with the carbon and metallic walls are rather similar. The database analysis yields in both cases a rather strong plasma current dependence and a clearly negative magnetic field dependence. The density dependence is negligible, while the well-known positive impact of triangularity appears very clearly. These dependences are partly at variance with the widely-used IPB98(y,2) confinement scaling, but in agreement with the analysis of the upgraded ITPA multi-machine and JET databases. The confinement degradation observed at high density with strong gas puffing can be clearly attributed to the density profile in the outer plasma region, as reported in previous studies.

Full conversion from ohmic to runaway electron (RE) driven current was observed in the tokamak à configuration variable (TCV) following massive injection of neon through a disruption mitigation valve into a low-density limited circular plasma. Following a partial disruption, a stable 200 kA RE beam is maintained for more than 1 s. Controlled ramp-down of the RE beam with adjustable decay rate was demonstrated. Control of the beam vertical position was achieved down to a RE current of 20 kA. RE beam formation is observed in elongated plasma configurations up to κ = 1.5. A reproducible scenario for RE beam generation without loss of circulating current is of particular interest for disruption modelling applications.