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

The 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.

Identification of the mechanisms for the H-mode density limit in machines with fully metallic walls, and their scaling to future devices is essential to find for these machines the optimal operational boundaries with the highest attainable density and confinement. Systematic investigations of H-mode density limit plasmas in experiments with deuterium external gas fuelling have been performed on machines with fully metallic walls, JET and AUG and results have been compared with one another. Basically, the operation phases are identical for both tokamaks: the stable H-mode phase, degrading H-mode phase, breakdown of the H-mode with energy confinement deterioration usually accompanied by a dithering cycling phase, followed by the l -mode phase. The observed H-mode density limit on both machines is found close to the Greenwald limit (n/n GW = 0.8-1.1 in the observed magnetic configurations). The similar behavior of the radiation on both tokamaks demonstrates that the density limit (DL) is neither related to additional energy losses from the confined region by radiation, nor to an inward collapse of the hot discharge core induced by overcooling of the plasma periphery by radiation. It was observed on both machines that detachment, as well as the X-point MARFE itself, does not trigger a transition in the confinement regime and thus does not present a limit on the plasma density. It is the plasma confinement, most likely determined by edge parameters, which is ultimately responsible for the transition from H-to L-mode. The measured Greenwald fractions are found to be consistent with the predictions from different theoretical models [16,30] based on MHD instability theory in the near-SOL. For complete list of authors see http://dx.doi.org/10.1016/j.nme.2017.01.005

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

Carbon (CD4) and nitrogen (N2) have been seeded in ASDEX Upgrade (AUG) with a tungsten wall and have both led to a 20–30% confinement improvement. The reference plasma is a standard target plasma with I p /B T = 1 MA/2.5 T, total input power P tot ~ 12 MW and normalized pressure of β N ~ 1.8. Carbon and nitrogen are almost perfectly exchangeable for the core, pedestal and divertor plasma in this experiment where impurity concentrations of C and N of 2% are achieved and Z eff only mildly increases from ~1.3 to ~1.7. As the radiation potentials of C and N are similar and peak well below 100 eV, both impurities act as divertor radiators and radiate well outside the pedestal region. The outer divertor is purposely kept in an attached state when C and N are seeded to avoid confinement degradation by detachment. As reported in earlier publications for nitrogen, carbon is also seen to reduce the high field side high density (the so-called HFSHD) in the scrape off layer above the inner divertor strike point by about 50%. This is accompanied by a confinement improvement for both low (δ ~ 0.25) and high (δ ~ 0.4) triangularity configurations for both seeding gases, due to an increase of pedestal temperature and stiff core temperature profiles. The electron density profiles show no apparent change due to the seeding. As an orthogonal effect, increasing the triangularity leads to an additionally increased pedestal density, independent of the impurity seeding. This experiment further closes the gap in understanding the confinement differences observed in carbon and metal wall devices; the absence of carbon can be substituted by nitrogen which leads to a similar confinement benefit. So far, no definite physics explanation for the confinement enhancement has been obtained, but the experimental observations in this paper provide input for further model development.

Recent experiments in ASDEX Upgrade aimed at improving the physics base for ITER and DEMO to prepare operation and aid the design. In order to increase its exhaust capabilities and operational flexibility, a new bulk W divertor as well as an adjustable cryopump had been installed prior to the 2014 campaign. In experiments with high-field-side pellet injection, central electron densities twice as high as the Greenwald density limit could be achieved without strongly increasing the pedestal density and deleterious effect on confinement. Due to its large installed heating power, a large normalized heat flux $P_{\mathrm{ sep}}/R = 10$ MWm $^{-1}$ has been reached, representing two-thirds of the ITER value, under partially detached conditions with a peak target heat flux well below 10 MWm $^{-2}$ . The divertor load could be further reduced by increasing the core radiation, still keeping the confinement in the range of $H_{98y2}\approx 1$ . Suppression of edge-localized modes (ELMs) at low collisionality has been observed in a narrow spectral window in contrast to earlier results at high densities. The ITER $Q = 10$ baseline scenario has been investigated, matching as close as possible the triangularity, the plasma beta, $q_{95}$ , and the distance to the L-H threshold. It turned out that the ELM frequency is low and consequently the energy ejected by a single ELM is very high and ELM mitigation appears to be difficult. As a possible alternative, a scenario has been developed achieving a similar performance at a lower plasma current (and consequently higher $q_{95})$ . Experiments using electron cyclotron current drive (ECCD) with feedback-controlled deposition have allowed successfully testing several control strategies for ITER, including automated control of (3, 2) and (2, 1) neoclassical tearing modes during a single discharge. Concerning advanced scenarios, experiments with central ctr-ECCD have been performed in order to modify the q-profile. A strong reversal of the q-profile could be stationarily achieved and an internal transport barrier could be triggered. In disruption mitigation studies with massive gas injection (MGI), a runaway electron beam could be provoked and mitigated by a second MGI. Ongoing enhancements aim at strengthening the power supplies in order to allow full use of the installed heating power, the exchange of two ion cyclotron resonance heating (ICRH) antennas to reduce the W influx during ICRH, and the upgrading of the electron cyclotron resonance heating (ECRH) system to 7–8 MW for 10 s.

The development of the baseline H-mode scenario foreseen for ITER on the ASDEX Upgrade tokamak, i.e. discharges at q 95 = 3, relatively low β N ~ 1.8, high normalized density n/n GW ~ 0.85 and high triangularity δ = 0.4, focused on the integration of elements foreseen for ITER and available on ASDEX Upgrade, such as ELM mitigation techniques and impurity seeding in combination with a metallic wall. Values for density and energy confinement simultaneously came close to the requirements of the ITER baseline scenario as long as β N stayed above 2. At lower heating power and thus lower β N normalized energy confinement H 98y2 ~ 0.85 is obtained. It has been found that stationary discharges are not easily achieved under these conditions due to the low natural ELM frequency occurring at the low q 95/high δ operational point. Up until now the ELM parameters were uncontrollable with the tools developed in other scenarios. Therefore studies on an alternative operational point at higher β N and q 95 have been conducted. In order to prepare for the ITER first non-activation operational phase, Helium operation has been investigated as well.

Recent experiments in ASDEX Upgrade aimed at improving the physics base for ITER and DEMO to aid the design and prepare operation. In order to increase its exhaust capabilities and operational flexibility a new bulk W divertor as well as an adjustable cryo-pump had been installed prior to the 2014 campaign. In experiments with high field side pellet injection central electron densities twice as high as the Greenwald density limit could be achieved without strongly increasing the pedestal density and deleterious effect on confinement. Due to its large installed heating power a large normalized heat flux Psep/R=10 MWm−1 has been reached, representing 2/3 of the ITER value, under partially detached conditions with a peak target heat flux well below 10 MWm−2. The divertor load could be further reduced by increasing the core radiation still keeping the confinement in the range of H98y≈1. ELM suppression at low collisionality has been observed in a narrow spectral window in contrast to earlier results at high densities. The ITER Q=10 baseline scenario has been investigated, matching as close as possible the triangularity, the plasma beta, q95 and the distance the L-H threshold. It turned out that the ELMs frequency is low and consequently the energy ejected by a single ELM is very high and ELM mitigation appears to be difficult. As a possible alternative an improved H-Mode scenario has been developed achieving a similar performance at lower plasma current (and consequently higher q95). Experiments using ECCD with feedback controlled deposition have allowed successfully testing several control strategies for ITER, including automated control of (3,2) and (2,1) NTMs during a single discharge. Concerning advanced scenarios, experiments with central ctr-ECCD have been performed in order to modify the q-profile. A strong reversal of the q-profile could be stationary achieved and an internal transport barrier could be triggered. In disruption mitigation studies with massive gas injection (MGI) a runaway electron beam could be provoked and mitigated by a second MGI. Ongoing enhancements aim at the strengthening of the power supplies in order to allow full use of the installed heating power, the exchange of two ICRH antennas to reduce the W influx during ICRH and the upgrading of the ECRH system to 7–8 MW for 10s.

Significant progress has been made on ASDEX Upgrade during the last two years in the basic understanding of transport, in the extension of the improved H-mode in parameter space and towards an integrated operating scenario and in the development of control methods for major performance limiting instabilities. The important features were the understanding of particle transport and the control of impurity accumulation based on it, the satisfactory operation with predominantly tungsten-clad walls, the improved H-mode operation over density ranges and for temperature ratios covering (non-simultaneously) the ITER requirements on ν*, n/nGW and Te/Ti, the ELM frequency control by pellet injection and the optimization of NTM suppression by DC-ECCD through variation of the launching angle. From these experiments an integrated scenario has emerged which extrapolates to a 50% improvement in n T τ or a 30% reduction of the required current when compared with the ITER base-line assumptions, with moderately peaked electron and controllable high-Z density profiles.

The scaling of the edge transport barrier (ETB) that sustains H-mode plasmas is crucial for the performance of next step tokamaks. At ASDEX Upgrade, the suite of edge diagnostics has been significantly improved to address this issue. High spatial resolution profiles of most of the key edge plasma parameters necessary to determine the magneto hydrodynamic (MHD) stability are now available. New high temporal resolution measurements give clear indications of the nonlinear evolution of the ELM crash. The correlation lengths of edge turbulence have been shown to be correlated with the edge radial electric field shear using a new correlation Doppler reflectometer system. The measured pressure gradient in the ETB is found to be consistent with ideal MHD stability limits, both for Type I and II ELMs. In addition, the edge electron temperature and density gradient lengths are found to be strongly correlated, leaving only the ETB width as a free parameter. In ASDEX Upgrade, the ETB width does not vary significantly over the entire H-mode edge database. Modelling of the transport of comparison discharges in hydrogen and deuterium shows that the expected mass effect on neutral penetration is largely compensated by more efficient heating of deuterium neutrals but requires a transport barrier in both the energy and particle channels in order to reproduce the measured edge temperature and density profiles.

Extensive studies of the H-mode density limit (DL) in JET gas-puffed discharges have been performed in the past four years targeting at an improved database for extrapolation to ITER. This paper reviews the arguments for the particular DL definition (pedestal density at the H-L boundary), the logic underlying the choice of parameters under focus (toroidal field, major radius, triangularity, safety factor) as well as some improvement in the interpretation of typical JET density ramp-up signatures that led to a critical review of the existing data. An empirical scaling is derived and compared with existing empirical and model based scalings. ASDEX Upgrade data are included in this analysis to provide information on the size dependence. The main results are: earlier findings on the B-t, R and q(95)-dependences are confirmed. The triangularity dependence, if any, is weak. The SOL-based BLS (Borrass, Lingertat, Schneider) scaling and the empirical scaling are virtually indistinguishable. The Greenwald scaling provides values in the right absolute range, but the overall fit is comparatively poor. The proposed scaling predicts ITER critical densities considerably below the reference value. Fuelling methods other than gas-puffing are outside the scope of this paper, but pellet fuelling, envisaged for ITER as an option to alleviate the situation, is discussed.

Charge exchange recombination spectroscopy, utilizing a fast lithium beam as the recombination source, has been successfully applied to measuring edge ion temperature profiles of fully stripped carbon and helium ions in Ohmic, L-mode and H-mode plasmas with a radial resolution of about 6 mm. The temperatures of the carbon and helium ions agree within their respective experimental error bars in discharges where both could be measured. Depending on the plasma scenario, impurity content, beam penetration, etc, either helium or carbon can be the better choice for accurate edge measurements. The impact of edge localized modes (ELMs) on this diagnostic method has been noted. Temperature profiles in between ELMs can be determined as long as the ELM frequency is less than half the frame rate of the detector, i.e. less than about 60 Hz. In a first application, accurate measurements in L-mode plasmas with dominant electron heating and low density show much higher ion temperatures than electron temperatures across the separatrix.

Detachment of high power discharges is obtained in ASDEX Upgrade by simultaneous feedback control of core radiation and divertor radiation or thermoelectric currents by the injection of radiating impurities. So far 2/3 of the ITER normalized heat flux Psep/R = 15 MW m−1 has been obtained in ASDEX Upgrade under partially detached conditions with a peak target heat flux well below 10 MW m−2. When the detachment is further pronounced towards lower peak heat flux at the target, substantial changes in edge localized mode (ELM) behaviour, density and radiation distribution occur. The time-averaged peak heat flux at both divertor targets can be reduced below 2 MW m−2, which offers an attractive DEMO divertor scenario with potential for simpler and cheaper technical solutions. Generally, pronounced detachment leads to a pedestal and core density rise by about 20–40%, moderate (

The high confinement mode (H-mode) is the operational scenario foreseen for ITER, DEMO and future fusion power plants. At high densities, which are favorable in order to maximize the fusion power, a back transition from the H-mode to the low confinement mode (L-mode) is observed. In present tokamaks, this H-mode density limit (HDL) occurs at densities on the order of, but below, the Greenwald density.In gas ramp discharges at the fully tungsten covered ASDEX Upgrade tokamak (AUG), four distinct operational phases are identified in the approach towards the HDL. These phases are a stable H-mode, a degrading H-mode, the breakdown of the H-mode and an L-mode. They are reproducible, quasi-stable plasma regimes and provide a framework in which the HDL can be further analyzed. During the evolution, energy losses are increased and a fueling limit is encountered. The latter is correlated to a plateau of electron density in the scrape-off layer (SOL). The well-known extension of the good confinement at high density with high triangularity is reflected in this scheme by extending the first phase to higher densities.In this work, two mechanisms are proposed, which can explain the experimental observations. The fueling limit is most likely correlated to an outward shift of the ionization profile. The additional energy loss channel is presumably linked to a regime of increased radial filament transport in the SOL. The SOL and divertor plasmas play a key role for both mechanisms, in line with the previous hypothesis that the HDL is edge-determined.The four phases are also observed in carbon covered AUG, although the HDL density exhibits a different dependency on the heating power and plasma current. This can be attributed to a changed energy loss channel in the presented scheme.

Modelling calculations for a fast He beam on its way through a tokamak plasma have been performed at TU Wien. They are based on the collisional-radiative ADAS model for He beams which delivers radially resolved population density and line intensity profiles for any initial population of a diagnostic He beam and any plasma structure. The effect of the metastable triplet population in the incident beam on the line emission profiles is shown. A sensitivity study of the optical lines with respect to electron temperature and density has been performed in the plasma edge and core regions. To validate these model calculations, experiments using fast He beams and beam emission spectroscopy have been performed at ASDEX Upgrade (IPP Garching) and JET by injecting He gas into the ion sources of the H- or D-heating beam systems.

Systematic investigations of gas fuelled, high density H mode discharges with the new closed Divertor II of ASDEX Upgrade are reported. The focus is on operational limits (H-L transition, density limit) over a wide range of externally controllable plasma parameters. It is found that the H mode threshold power dramatically exceeds the generally accepted prediction Pthreshheat ∝ eBt when high densities are approached. Highest densities (density limit) are achieved close to the H-L mode back-transition and are generally associated with complete divertor detachment. The scaling of the density limit is assessed and discussed with a view to detachment based models and the Greenwald limit.

Experiments on electron heat transport were performed in the tokamak ASDEX Upgrade, mainly in ohmically heated plasmas, applying either edge cooling with laser blow-off impurity injection or edge heat pulses with ECRH. Repetitive pulses within one plasma discharge were made allowing Fourier transformation of the temperature perturbation, yielding a good signal to noise ratio up to high harmonics and a detailed investigation of the pulse propagation. For densities lower than 1.8 x 10(19) m(-3), an increase of central electron temperature was found as the response to edge cooling via impurity injection similar to observations made in the TEXT, TFTR, RTP and Tore Supra tokamaks. The inversion does not appear instantaneously, but with a time delay roughly compatible with diffusion. Modelling of the propagation of the cold pulses in the framework of the IFS-PPPL model yields qualitative agreement. However, the predicted increase of ion temperature is not observed experimentally on a fast timescale. The response to ECRH heat pulses is not perfectly symmetrical to cold pulse experiments, but the similarities suggest a common underlying physical mechanism. No inversion of the heat pulse is found, instead the initial pulse from the edge is associated with a second, much slower, heat pulse in the centre. This central reaction is in fact quite similar (and not symmetrical) to that of the cold pulses. By varying the position of the ECRH deposition it could be shown that the central increase is related to the arrival of the pulse close to the inversion radius and not to the initial pulse. Modelling was performed to investigate the possible physics mechanisms and for comparison with the cold pulses.

The effect of plasma shape variation (in particular, variation of the upper and lower triangularity ?) on edge localized modes (ELMs) and H-mode pedestal properties in ASDEX Upgrade is reported here. Strongly shaped plasmas (high ?) show an increased edge pressure gradient and, without external gas puff, type-I ELMs generally have larger losses and lower frequency than in plasmas with low ?. With external gas puff, ELM losses at high ? are reduced and in the same range as found for low ?. The average ELM power loss is a constant fraction of the total loss power, independent of triangularity. The width of the steep electron temperature and pressure gradient zone remains essentially constant at low ? while at high ? it shows a variation inconsistent with a poloidal gyroradius scaling. The edge pressure gradient and the pedestal pressure scales strongly with plasma current and triangularity. In type-I ELM H-modes, the pedestal pressure is directly related to the global stored plasma energy, independent of plasma shape.

Internal transport barriers with the central electron temperature as large as the central ion temperature both in excess of 10 keV have been achieved in the Axi-symmetric divertor experiment (ASDEX Upgrade) [H. Vernickel et al., J. Nucl. Mater. 128, 71 (1984)]. By applying central electron cyclotron heating and current drive to negative central shear discharges, established by neutral beam heating in the current ramp, the core electron temperatures could be raised by more than a factor of 2. Despite the fivefold increase of the central electron heat flux, the ion and electron energy and also angular momentum transport did not deteriorate. For neutral beam injection alone and also with additional central electron cyclotron counter-current drive, a double tearing mode and the associated detrimental effect on the plasma confinement is destabilized only transiently, when the minimum of the safety factor (qmin) passes through 2. For co-current drive, however, the confinement does not recover after qmin has dro...

Operation in ASDEX Upgrade and JET with different divertor geometries is compared with emphasis upon energy and particle exhaust and the impact of the divertor on core plasma performance. Closing the divertor increases the ratio of the neutral flux density in the divertor to that in the main chamber and facilitates improved pumping. The influence of closure on H-mode performance turns out to be weak, the main chamber physics dominates. The electron density/neutral hydrogen flux in the scrape-off layer is found to be a key parameter for particle and energy confinement as well as for the radiative losses. The density in the outer scrape-off layer is related to main chamber recycling, while the separatrix density is primarily affected by divertor geometry. Reduced target plate power load due to an increased divertor radiation level is obtained in ASDEX Upgrade Div II. The radiated power fraction in JET is systematically lower, which is partly explained by the lower edge densities. Core plasma Zeff and carbon concentrations are similar in both machines and do not depend on divertor geometry.

Steady-state discharges with improved core confinement and H-mode edge with edge localized modes (ELMs) are investigated. In plasmas with an upper triangularity top close to zero an H-factor of HITER89-P = 2.7 and N = 2.2 could be maintained for 1 s and HITER89-P = 2.4 and N = 2.0 for 6 s, the latter corresponding to 40 confinement times or 2 1/2 resistive time scales for current redistribution, only limited by the duration of the possible discharge length. At a line averaged density of 4 × 1019 m-3 the central temperatures reach values of Ti = 10 keV and Te = 6.5 keV. The stationarity of the current profile is explained by magnetic reconnection driven by strong (m = 1, n = 1) fishbones, which, in the absence of sawteeth, also expel energy and impurities. Further increasing the pressure, is limited by neoclassical tearing modes. Raising the density by edge gas fuelling and the simultaneous increase of the neutral beam power, HITER89-P remained unchanged up to ne = 5.5 × 1019 m-3, accompanied by a substantial reduction of Zeff. Increasing top to 0.2, both confinement and -limit improved reaching values of HITER89-P = 3.0 and N = 2.4 at densities above ne = 5 × 1019 m-3. This resulted in the highest fusion product of nD,0Ti,0E = 0.9 × 1020 keV s m-3 so far observed in ASDEX Upgrade.

Stationary discharges with $H$-mode edge and internal transport barrier for energy, momentum, and particle transport have been obtained on ASDEX Upgrade. At a line averaged density of $4\ifmmode\times\else\texttimes\fi{}{10}^{19}{\mathrm{m}}^{\ensuremath{-}3}$ an $H$ factor of ${H}_{\mathrm{ITER}89\ensuremath{-}P}\phantom{\rule{0ex}{0ex}}=\phantom{\rule{0ex}{0ex}}2.4$ and ${\ensuremath{\beta}}_{N}\phantom{\rule{0ex}{0ex}}=\phantom{\rule{0ex}{0ex}}2$ could be maintained for 6 s corresponding to 40 confinement times, limited only by the possible discharge duration. The $q$ profile is flat in the center and close to 1. Current diffusion calculations suggest that reconnection is required to clamp the safety factor at 1, which in the absence of sawteeth is explained by the strong fishbone activity.

Recently, there has been increased interest in the ratio nS/nC, where nS and nC are the separatrix and pedestal densities, respectively. A semi-empirical model for this ratio is proposed which puts particular emphasis on the role of edge physics. The empirical elements of the model and the detailed comparison with data rely on the extensive edge density database established at ASDEX Upgrade. The derived scaling can be cast into a form where nS/nC is essentially determined by the proximity to the Greenwald limit.

In 1997 the new `LYRA' divertor went into operation at ASDEX Upgrade and, in parallel, the neutral beam heating power was increased to 20 MW by installation of a second injector leading to a P/R value of 12 MW/m. Experiments have shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. There is an overall reduction of the maximum heat flux in the LYRA divertor by about a factor of 2 compared with the previous open divertor Div I. This reduction is mainly due to increased radiative losses inside the divertor region, which are caused by an effective reflection of hydrogen neutrals into the hot separatrix region. The main channel of radiative loss is carbon radiation, which cools the divertor plasma down to a few electronvolts, where hydrogen radiation losses become significant. The radiative losses preferentially reduce the power flux at the separatrix, leading to early detachment around the strike point position. With increasing density, the detached region extends upwards on the vertical target. The power fraction radiated in the LYRA divertor is around 45% and nearly independent of the heating power. This value is a factor of 2 higher than the typical radiation fraction in Div I. B2-EIRENE modelling of the performed experiments supports the experimental finding and refines the understanding of loss processes in the divertor region.

The divertor geometry on ASDEX-Upgrade has changed from an open to a more closed configuration. The effects of this change on midplane scrape-off layer profiles for temperature and density are discussed. Investigations concerning the accuracy of separatrix positions are described. The relation between neutral gas flux density in the divertor and midplane density is analysed and scalings are presented. No decrease in the maximum achievable density in the H-mode with the new closed divertor in comparison to the former open one has been found. For comparable discharges the scrape-off layer density for the closed divertor has been found to be higher or at least as high as for the open divertor.

The change of the divertor plasma behaviour from Div-I to Div-II for ASDEX Upgrade as measured by a set of reconstructed or newly designed divertor diagnostics is presented. The Div-II configuration is characterised – due to the highly inclined target plates – by reflection of neutrals towards the separatrix. Therefore, detachment in Div-II starts rather early localised close to the separatrix. In contrast, the complete global divertor detachment is practically unchanged, because the outer parts of the SOL stay attached even for high densities. The power distribution on the divertor in Div-II is much broader than in Div-I, resulting in a large reduction of peak power loads both in L and H-mode (factor of 2–4) due to larger divertor radiation losses. The larger losses are caused by larger hydrogen losses, enhancement of carbon radiation due to radial transport and convective energy transport into the radiation zone and larger radial energy transport in the divertor. The new Div-II geometry shows larger neutral gas densities in the divertor for the same line averaged density and a much faster helium exhaust rate ( τ ∗ He /τ E ≈4 ) in H-mode. The neon compression is worse compared with helium. No strong effect on impurity compression and overall divertor performance was seen by puff and pump experiments, neither in Div-I nor in Div-II.

Operational regime boundaries for High confinement mode (H-mode) plasmas with high edge density are investigated on the ASDEX Upgrade tokamak. In addition to the MHD stability boundary imposed by Edge Localized Modes (ELMs) and the H-mode threshold, the onset of complete divertor detachment limits the accessible separatrix density. At the H-mode density limit, a transport enhancement is observed that reduces the temperature gradient in the edge of the main plasma and can lead to a disruption by loss of H-mode and and subsequent radiation and resistive MHD instability. This `transport limit' occurs parallel to full divertor detachment at high edge collisionalities. A possible origin is drift-Alfven-ballooning turbulence driven by the pressure gradient at high collisionality.

We present an experimental and theoretical study on the modelling of fast Li beams which are used for diagnostics of fusion edge plasmas. The atomic collision database utilized for modelling the Li beam attenuation has been revised and extended by detailed investigations of populations resulting from ion impact excitation. We obtain good agreement between measured and calculated Li(nl) populations for an Li beam passing the edge plasma region of the ASDEX Upgrade tokamak and WENDELSTEIN 7 AS stellarator, respectively, at IPP Garching.

A key element in future toroidal magnetic fusion machines like ITER is the design of a divertor, which allows for safe particle and power exhaust in parallel with high bulk plasma performance. Correspondingly, the definition and design of an optimized divertor is a major task within the ongoing international ITER research and development effort. In order to provide a profound physics basis for such a divertor optimization, different divertor geometries are being tested on major tokamaks. This paper describes the effects of these divertor modifications on plasma performance in ASDEX Upgrade and in JET. In conclusion, increasing closure improves divertor performance without limiting the core plasma performance in ELMy H-modes.

Particle and energy losses due to type-I edge-localized modes (ELMs) and the effect of H-mode pedestal parameters on global confinement are studied for H-Mode discharges in ASDEX Upgrade. ELM frequency as well as particle and heat losses during ELMs are strongly affected by gas puffing, impurity injection, variation of plasma current and heating power. However the average power loss due to ELMs P-ELM = f(ELM)Delta W shows much less variation, and cannot account for the variations of stored energy encountered, e.g. in a plasma current scan. A good correlation between pedestal pressure and stored energy is found for type-I ELMy discharges.

Global and local H-mode threshold analyses are presented and discussed. The density window was extended up to the density limit towards which the power threshold exhibits a dramatic increase. First results with the new divertor are compared to the previous threshold scaling. Local edge measurements at the L-H transition show that the temperature decreases slightly with density and increases with magnetic field, and that the ion collisionality at the plasma edge is always clearly above unity, between 5 and 15. Experiments at high power threshold exhibit improved L-mode confinement and suggest that two mechanisms are required for the L-H transition.

A new set of diagnostics has been implemented on ASDEX Upgrade to measure the input impedance of the ICRF antennas, in the form of a voltage and current probe pair installed on each feeding line of every antenna. Besides allowing the measurement of the reflection coefficient Gamma of each antenna port, the probes have two advantages: first, they are located close to the antenna ports (similar to 3 m) and thus the measurements are not affected by the uncertainties due to the transmission and matching network; second, they are independent of matching conditions. These diagnostics have been used to study the behavior of the ASDEX Upgrade antennas while changing the plasma shape (low to high triangularity) and applying magnetic perturbations (MPs) via saddle coils. Scans in the separatrix position R-sep were also performed. Upper triangularity delta(o) was increased from 0.1 to 0.3 (with the lower triangularity delta(o) kept roughly constant at 0.45) and significant decreases in vertical bar Gamma vertical bar (up to similar to 30%, markedly improving antenna coupling) and moderate changes in phase (up to similar to 5 degrees) off on each feeding line were observed approximately at delta(o) >= 0.29. During MPs (in similar to 0.5 s pulses with a coil current of 1 kA), a smaller response was observed: 6% - 7% in vertical bar Gamma vertical bar, with changes in phase of 5 apparently due to R p scans only. As 1 is usually in the range 0.8 - 0.9, this still leads to a significant increase in possible coupled power. Numerical simulations of the antenna behavior were carried out using the FELICE code; the simulation results are in qualitative agreement with experimental measurements. The results presented here complement the studies on the influence of gas injection and MPs on the ICRF antenna performance presented in [4].

Edge-localized mode (ELM) triggering and pacing in an all-metal wall environment shows significant differences to a first-wall configuration containing carbon. Here we report on experiments performed at ASDEX Upgrade revisiting the issue with all plasma-facing surfaces now fully replaced by tungsten. This investigation was motivated by experimental findings indicating that ELM triggering becomes more intricate when the carbon is replaced by a metal wall. ELM pacing could no longer be achieved by magnetic triggering in ASDEX Upgrade under conditions that previously showed a positive response. Also, recent investigations at JET indicate that a lag time occurs in pellet ELM triggering when operating with the new ITER-like wall. The ASDEX Upgrade centrifuge-based launching system was revitalized and upgraded for this study, now allowing detailed analysis of the ELM trigger response. The appearance of a lag time for pellet ELM triggering in an all-metal wall environment was confirmed. While different lag time durations were found for several type-I ELMy H-mode scenarios, the magnitude of the pellet perturbation was found to cause no difference. Reducing the auxiliary heating power for ELM triggering clearly makes the pellet tool less efficient for ELM control purposes; however, this affords a major benefit when applied for fuelling. Plasma operation with benign ELM behaviour at core densities far beyond the Greenwald limit was demonstrated, this being fully reversible and not affecting the energy confinement.

Local edge parameters on the ASDEX Upgrade tokamak are investigated at the L-mode to H-mode transition, during phases with various types of edge-localized modes (ELMs), and at the density limit. A scaling law for the boundary electron temperature, , is found which describes the H-mode threshold for deuterium-puffed discharges with favourable ion -drift direction. The region of stable operation is bounded by type I ELMs near the ideal ballooning limit and by a minimum temperature necessary to avoid thermal instability of the plasma edge. Stationary operation with type III ELMs imposes an upper limit on the edge temperature. Within the entire range of boundary densities investigated , both L-mode and H-mode are found to be accessible. During type I ELMy H-mode, a relation of global confinement with the edge pressure gradient is found which is connected with a loss of the favourable density dependence predicted by the ITER-92P and ITER-93H ELMy H-mode scalings. At high density, better confinement is achieved in H-modes with an edge pressure gradient below the ideal ballooning limit, e.g. during type III ELMy H-mode with impurity-seeded radiation.

The L mode and H mode density operational window in the vicinity of the density limit has been investigated with a combination of gas puff refuelling and improved fine tuning of neutral beam injection (NBI) heating power. In this way, a novel strategy is achieved by means of a parallel increase of density and heating power. As the density limit is approached, H modes degrade into L modes independently of heating power; this is in contrast to the generally accepted L to H mode threshold scaling PheatL-H varies as neB. Furthermore, contrary to the well known heating power independent Greenwald limit, the L mode density limit increases moderately with rising heating power, neDL varies as Pheat0.3+or-0.1, if a simple power law is assumed. The power dependence becomes more obvious when analysed in terms of edge densities and powers flowing across the separatrix into the scrape-off layer, nesep varies as Psep0.6+or-0.2. The corresponding H mode studies show that before an H mode quenches into an L mode the maximum achievable density (i.e. The H mode density limit) is practically independent of the heating power, as observed on many machines

The flow of heat in the scrape-off layer (SOL) region of ASDEX Upgrade is investigated and compared with simple modelling. Parallel heat transport is found to be consistent with electron heat conduction based on Spitzer - H?rm conductivity. Cross-field heat transport is characterized using radial e-folding distances for power, temperature and plasma pressure, which are all found to vary weakly over a wide range of discharge conditions. Type I ELMs, also characterized, introduce a discreteness to the power flow into the SOL and carry approximately half of the power exhaust from the discharge. The divertor plates are effectively screened from the ELM energy, even in low-radiation discharges, suggesting enhanced radiation rates during ELMs.

A simple analytic model of divertor radiation is used as a framework for identifying different routes to divertor detachment. Three means of achieving detachment are demonstrated experimentally including, increasing plasma density, purposely increasing the impurity level and increasing the scrape-off layer connection length. The resulting low electron temperatures in the divertor causes friction processes to dominate ionization in the recycling region, giving rise to plasma pressure loss along field lines. The pressure loss is in approximate agreement with simple analytic modelling.

Power exhaust is one of the critical questions to be solved for ITER, especially with the side condition of sufficient particle exhaust. Since the H-mode threshold might require a high power flux over the separatrix, only small power losses inside the bulk plasma might be permitted, and thus eventually a large fraction of the power must be removed in the scrape-off layer and divertor plasma. Neutral energy losses can only remove a small fraction of the power, and therefore most of the power flux has to be radiated, either by hydrogen isotopes, by the intrinsic impurities (mainly carbon) or by additionally seeded impurities. However, all these scenarios with reduced power flow to the divertor imply also a reduced particle flow into the divertor. This might be critical for particle exhaust, especially concerning the helium ash. Reduction of the power flow to the divertor by enhanced radiation losses has been investigated in the large divertox experiments like JET, DIII-D, ASDEX Upgrade, and JT-60U, and an experimental overview of the present status will be given in this paper, together with a discussion of impurity exhaust (mainly He and Ne) in these scenarios.

The influence of impurity radiation on the and transitions is investigated for highly radiative divertor discharges in the ASDEX Upgrade tokamak. The transitions between H- and L-mode depend on the net heat flow across the separatrix, , calculated from the heating power and bolometric radiation profiles. For typical radiative boundary conditions in ASDEX Upgrade, the radiation distribution is dominated by line emission in the vacuum-ultraviolet region and peaks near the separatrix. For the case of neon used as seed impurity, about 2/3 of the main chamber radiation is emitted inside, but close to the separatrix. Argon seed results in a higher fraction of core radiation, while the line emission is shifted further outside for nitrogen. The radiation-corrected threshold is not affected by gas puffing and is described by [MW, , T, amu]. The threshold power, which is typically lower by a factor of two without strong deuterium puffing, is increased by heavy gas puffing leading to . In the vicinity of the radiation-induced transition, a general alignment of H and L mode is observed with regard to global energy confinement time and edge density and temperature profiles. The transition itself exhibits a smooth evolution in time. Reduction of target plate power load down to about 10% of the total heating power is easily achieved by edge radiation in the CDH-mode for low conditions. However, this reduction is attributed mainly to radiation from inside the separatrix and is connected to relatively high values of the core . These results emphasize the importance of the development of more closed divertor concepts, allowing for higher divertor radiation levels in connection with lower core .

Characteristics of the various types of edge localized modes (ELMs) observed during the H-mode of the ASDEX Upgrade tokamak are presented. Magnetic high-mode-number ELM precursor activity is found for type I ELMs and type III ELMs, the former only during neutral beam injection in the counter direction. The effect of ELMs on electron temperature and density profiles is discussed.

ITER or any other future fusion device will operate in H-mode to take advantage of the improved confinement. Stationary operation in H-mode, however, requires the occurrence of ELMs, and the resulting power load on the divertor target plates is not acceptable from the engineering point of view. Beyond this, the present divertor concepts rely on a cold and dense (possibly detached) divertor plasma. Large type-I ELMs, however, carry too large a power and possibly burn through such cold divertor plasmas. The completely detached H-mode (CDH-mode), recently established in ASDEX Upgrade shows detachment at both divertor target plates and small ELMs (so-called type-III ELMs), that do not burn through the divertor plasma. The extrapolation of this scenario to ITER, however, is an important question.

The radiation and transport characteristics of ASDEX Upgrade discharges with a neon-driven radiative mantle are modelled using a one-dimensional radial impurity transport code that has been coupled to a simple divertor model describing particle recycling and pumping. The code is well suited to describing the measured impurity line radiation, total, soft x-ray and bremsstrahlung radiation in regions of the plasma that are not dominated by two-dimensional effects. The recycling and pumping behaviour of neon as well as the bulk transport of neon for radiative boundary scenarios are discussed.

Divertor plasmas with strong external gas puffing in ASDEX Upgrade have shown very efficient impurity retention, increasing with the divertor neutral gas density. The experiments presented here use feedback-controlled gas puffs in discharges with different pumping speed to keep the divertor neutral gas flux density the same. This allows for the first time a decoupling of the divertor neutral gas flux density and the external gas flow. The resulting plasmas are almost identical and show identical impurity retention, clearly demonstrating the importance of the divertor neutral gas density over the externally induced flow. {copyright} {ital 1996 The American Physical Society.}

ASDEX Upgrade (AUG) has been converted to all W plasma facing components (PFCs) in 2007 and JET has implemented the ITER like wall (ILW) project (2011) using the same PFC configuration as ITER during its active phase, namely Be in the main chamber and tungsten in the divertor. As a result of the all metal PFCs both devices much less surface conditioning is needed to arrive at reproducible wall conditions. Specifically the Be PFCs of JET led to a very small low-Z content (reduction of C and O by at least a factor of 10), reducing the edge radiation in steady state operation as well as during disruptions. Both devices successfully employ massive gas injection to mitigate disruption forces and power loads to PFCs by radiating up to 100% of the available energy. Hydrogen retention is strongly reduced (AUG: factor 5, JET: factor 10) and the remaining retention is still dominated by co-deposition with residual C in AUG and intrinsic Be in JET. The very low edge and divertor radiation could be compensated by impurity seeding either by a single gas species (N2) (AUG and JET) or by combining N2 and Ar (AUG) injection for divertor and main chamber radiation, respectively. The W sputtering in the divertor increases when seeding small amounts of N2, but decreases for higher fluxes due to the plasma cooling provided by the nitrogen radiation. The tungsten content is controlled by the source as well as by its edge and central transport. It could be kept sufficiently small by using a minimum gas fuelling to reduce the W erosion and to diminish the W penetration. The control of the central W transport by central (wave) heating had been well established in AUG, however in both devices the W content is increased during ICRH operation most probably due to increased W sputtering caused by rectified sheaths. The H-Mode threshold is reduced by 20-30% in AUG and JET, but on average the confinement is lower in JET-ILW than with C PFCs. To date it is not yet clear, whether the reduced H-Mode confinement has to be attributed to the use of W PFCs, since such a clear trend as in JET was not found in AUG. The increase of confinement with N2 seeding observed in both devices hints to the fact, that low-Z impurities like carbon or nitrogen play a beneficial role for the pedestal confinement.

The completely detached high confinement (CDH) regime established recently in high power ASDEX-Upgrade single null divertor discharges is described. The standard CDH mode scenario with feedback controlled external deuterium and neon puffing is analysed in detail with respect to its divertor detachment and confinement properties. Impurity transport and pumping effects and the density operation windows of the CDH mode are outlined. First results with other impurity radiators (N,Ar) are given. Finally the CDH mode operation space is discussed in terms of main chamber and scrape-off layer plus divertor radiation and possible consequences for future fusion experiments are outlined.

ASDEX Upgrade (AUG) is an ITER-shaped divertor tokamak with versatile heating, fueling, exhaust, and control systems. All plasma-facing components (PFCs) are coated with tungsten layers. Plasma scenarios have been adopted that avoid central tungsten accumulation, which can lead to an H-L transition due to excessive central radiative losses. Compared to a carbon-PFC tokamak, the AUG operational space is slightly more weighted toward higher densities and collisionalities. Actual and future planned extensions aim toward reducing the core collisionality while maintaining good power and particle exhaust. These extensions include a solid tungsten outer divertor target, improved pumping, higher ECRH power, and modified ICRF antennas that reduce tungsten sources. The newest element for advanced plasma control is the first set of eight magnetic perturbation coils, which already achieved type-I edge localized mode mitigation in multiple plasma scenarios. Another eight coils have been installed in autumn 2011, allowing the production of mode spectra with n >; 2. In parallel to the improved actuator set, an increasing number of diagnostics are brought into real-time state, allowing versatile profile and stability control.

ASDEX Upgrade is an ITER shaped divertor tokamak with versatile heating, fueling, exhaust and control systems. All plasma facing components are coated with tungsten layers. Plasma scenarios have been adopted to avoid central tungsten accumulation, which can lead to an H-L transition due to excessive central radiative losses. Compared to a carbon-PFC tokamak, the AUG operation space is slightly more weighted towards higher densities and collisionalities. Actual and future planned extensions aim towards reducing the core collisionality while maintaining good power and particle exhaust. These extensions include a solid tungsten outer divertor target, improved pumping, a higher ECRH power and modified ICRF antennas for reduced tungsten sources. The newest element for advanced plasma control is the first set of 8 magnetic perturbation coils, which already achieved type-I ELM mitigation in various plasma scenarios. Another 8 coils will be installed in autumn 2011 allowing to produce mode spectra with n > 2. In parallel to the improved actuator set, an increasing number of diagnostics is brought into real-time state, allowing versatile profile and stability control.

ASDEX Upgrade is a poloidal divertor experiment very similar to ITER with respect to magnetic field properties and especially to the plasma boundary geometry. A large part of the programme on ASDEX Upgrade is therefore dedicated to investigating and optimizing reactor-relevant plasma boundary issues. This contribution is concerned with H-mode studies and edge physics in general. The L/H-power threshold, H-mode confinement, details of ELM dynamic and 'dithering' L/H-transitions will be discussed. The edge physics investigations are concerned with model calculations and validations, the characteristic 'approach' to the density limit (DL), target plate sputtering and the influence of hydrocarbons, asymmetries and drifts depending on the direction of the toroidal magnetic field, and finally first results of high-Z material experiments.

This addendum applies to papers authored by contributors from the Axially Symmetric Divertor Experiment (ASDEX) Upgrade Team published in Review of Scientific Instruments. This addendum provides the full list of ASDEX Upgrade Team contributors and their affiliations.