[en] The prediction of Q=10 for ITER is based on standard ELMy H-mode performance. In recent years robust operating scenarios have been developed which extrapolate for ITER to significantly higher Q-values at full current or longer inductive pulse lengths at reduced plasma current with a non-inductive current drive fraction above 50%, called hybrid operation. In particular the stationary improved H-Mode, realized at ASDEX Upgrade in 1998, combines improved core confinement and stability with an H-mode edge. Further developments of this regime by AUG and DIII-D, its demonstration at JET and the similarity to the high-beta pol plasmas at JT-60U have proven that it is a strong candidate for ITER hybrid operation. This scenario is stationary for several current re-distribution times and is characterized by a central q above 1, low central magnetic shear, fish bones or low (m,n) neoclassical tearing modes (NTM's) replacing sawteeth, and a confinement factor with respect to H-mode of up to 1.4. (3,2) NTM's remain small, enabling routine operation up to .N 3 at ITER relevant collisionalities for a broad range of q95 =3.2 4.5. Operation at low q maximises performance, while higher q95 maximise bootstrap current fractions and pulse length. Fusion performances in terms of HP-98.N / q95 20.4 have been stationary maintained over a broad density range compared with 0.2 assumed for the standard ITER scenario. Even at densities close to the Greenwald density needed for optimum exhaust . values up to .N3.5 and a fusion performance of 0.35 were achieved. Temperature profiles are still ITG/TEM turbulence dominated while density profiles are peaked even at high densities enhancing confinement. Extrapolated to ITER one still expects from quasi-linear models peaked density profiles (R/Ln 3). A further contribution to global confinement can arise from an increase of edge pedestal pressure combined with stiff temperature profiles. A main focus is to obtain improved H-modes over the widest range of non-dimensional parameters extending to ITER values. In AUG at low collisionality, .N close to 3 and HP-98>1 were achieved for .* ranging from 7.5 to 11-10-3 at fixed q954. At the lowest .* a heating power of 20 MW was needed. This result indicates a more favourable scaling of the maximum .N to ITER compared to standard H-modes. No insurmountable problem arose from impurity accumulation despite the peaked density profiles. Using tailored heat deposition with central wave heating and broad NI deposition a compromise in density peaking was found allowing enhanced confinement and keeping the tungsten concentration below 10-5 even with tungsten coated first wall and divertor structures. Experiments with dominant ICRH heating demonstrated operation slightly below .N3 with reactor relevant heating schemes without particle input. Finally, benign type II ELMs have been combined with improved H-modes at high densities and close to double null configurations. We present the existing database and the status of theoretical understanding of the underlying transport and the self-regulating nature of the current density profile, proceeding evidently through (possibly differing) benign instabilities. (Author)

[en] The toroidal rotation of H-mode plasmas in ASDEX Upgrade is studied in the outermost 5 cm of the confined plasma. The projection of the rotation velocity along the line of sight (approximately toroidal) is measured using charge exchange recombination spectroscopy, with a radial resolution of up to 3 mm and a temporal resolution of 1.9 ms. At about 1 cm inside the separatrix the rotation exhibits a local minimum. From there, the rotation in codirection increases towards the plasma center and towards the separatrix. The latter increase is the focus of this work. It is situated in the region of the edge transport barrier and amounts to 10-20 km/s. It is observed for D⁺, He²⁺, B⁵⁺, and C⁶⁺. The described rotation feature at the edge is not visible during an ELM crash and is probably connected to the occurrence of steep gradients in this plasma region

[en] During the combined plasma heating with neutral beam injection (NBI) and waves in the ion cyclotron (IC) range of frequencies, the NBI fast ions are preferentially accelerated by IC waves close to the IC harmonics, as a consequence of finite Larmor radius (FLR) effects. Since the NBI fast ions are expected to have a strong influence on the wave absorption and propagation, we have implemented a NBI source in the quasilinear Fokker-Planck SSFPQL code, interfaced with the toroidal full-wave TORIC solver. In this implementation the NBI ionization sources are obtained from the output of a Monte Carlo code, such as FAFNER. The numerical scheme adopted in the TORIC-SSFPQL package allows to describe very anisotropic sources, such as NBI, and to iterate the solution of Maxwell's equation taking into account selfconsistently the fast ion tails. As a first application, we present modeling of an ASDEX-Upgrade discharge with combined NBI and ICRF heating.

[en] This paper describes and further develops the understanding of toroidal momentum transport. Nonlinear gyro-kinetic simulations of the ion temperature gradient mode with adiabatic electrons show a strong coupling between the momentum and ion heat transport, with the ratio of the transport coefficients close to 1. Linear theory using a global description predicts an off-diagonal contribution to the momentum flux even in the absence of a radial electric field. The influence of the toroidal velocity gradient on the ion temperature profile is found to be small in the H-mode. These predictions are in qualitative agreement with experimental observations

[en] Due to wall lifetime requirements and the problem of tritium co-deposition in hydrocarbon layers, a future burning plasma will most probably have a full high-Z wall. The prime candidate material is tungsten, which exhibits good thermo-mechanical properties and has a high energy threshold for physical sputtering. To investigate the reactor-compatibility of this wall material, ASDEX Upgrade is being converted into a full-tungsten coated tokamak in a step-by-step approach, with presently almost 70 % of W wall coverage. The effect of the reduction of primary carbon coverage on the plasma is so far moderate. Under tokamak conditions, carbon behaves like a recycling impurity, due to the deposition and re-erosion of soft hydrocarbon layers on the tungsten surface. During high density H-mode operation, the central tungsten concentrations remain typically low, i.e. well under 10-5. The situation is more critical in the improved H-mode or hybrid scenario. Here, the combination of hot edge conditions and peaked central density profiles result in high central tungsten concentrations of up to 10-4, which would be critical in a reactor. However, core electron density peaking is reduced by use of central ICR or ECR heating and thus in turn suppresses central tungsten accumulation. For extrapolation to reactor conditions, we need to separate the effects of the tungsten wall source, the penetration over the edge transport barrier (ETB) and the core transport with its strong neoclassical contribution. These issues are addressed by inspecting the tungsten behaviour in various discharge scenarios and parameters in ASDEX Upgrade. These include radiative cooling by medium-Z seed impurities and ELM frequency control by pellet injection to simulate a reactor plasma with small edge and divertor impurity radiation levels and a separatrix power flux close to the H-L threshold. Fast ions produced by NBI and ICR heating at the low field side appear to be an important tungsten erosion mechanism, as observed by spectroscopic influx measurements of WI. The penetration of tungsten over the ETB is highly affected by ELMs, showing a clear reduction of the core tungsten content with rising ELM frequency. This effect is explained by the fact that the ELMs expel a major fraction of the tungsten ions which accumulate inside the ETB during the inter-ELM phase. Core neoclassical ion transport acts as a multiplier for the pedestal density, which depends on the temperature and density gradients in the core plasma. The ASDEX Upgrade results will be complemented by results obtained with molybdenum in Alcator C-Mod and with a tungsten limiter in TEXTOR. (Author)

[en] EDGE2D-EIRENE (the ‘code’) simulations show that radial electric field, E _r, in the near scrape-off layer (SOL) of tokamaks can have large variations leading to a strong local E × B shear greatly exceeding that in the core region. This was pointed out in simulations of JET plasmas with varying divertor geometry, where the magnetic configuration with larger predicted near SOL E _r was found to have lower H-mode power threshold, suggesting that turbulence suppression in the SOL by local E × B shear can be a player in the L–H transition physics (Delabie et al 2015 42nd EPS Conf. on Plasma Physics (Lisbon, Portugal, 22–26 June 2015) paper O3.113 (http://ocs.ciemat.es/EPS2015PAP/pdf/O3.113.pdf), Chankin et al 2017 Nucl. Mater. Energy 12 273). Further code modeling of JET plasmas by changing hydrogen isotopes (H–D–T) showed that the magnitude of the near SOL E _r is lower in H cases in which the H-mode threshold power is higher (Chankin et al 2017 Plasma Phys. Control. Fusion 59 045012). From the experiment it is also known that hydrogen plasmas have poorer particle and energy confinement than deuterium plasmas, consistent with the code simulation results showing larger particle diffusion coefficients at the plasma edge, including SOL, in hydrogen plasmas (Maggi et al 2018 Plasma Phys. Control. Fusion 60 014045). All these experimental observations and code results support the hypothesis that the near SOL E × B shear can have an impact on the plasma confinement. The present work analyzes neutral ionization patterns of JET plasmas with different hydrogen isotopes in L-mode cases with fixed input power and gas puffing rate, and its impact on target electron temperature, T _e, and SOL E _r. The possibility of a self-feeding mechanism for the increase in the SOL E _r via the interplay between poloidal E × B drift and target T _e is discussed. It is also shown that reducing anomalous turbulent transport coefficients, particle diffusion and electron and ion heat conductivities, leads to higher peak target T _e and larger E _r, suggesting the possibility of a positive feedback loop, under an implicitly made assumption that the E × B shear in the SOL is capable of suppressing turbulence. (paper)

[en] Recent EDGE2D-EIRENE simulations of JET plasmas showed a significant difference between radial electric field ( E _r) profiles across the separatrix in two divertor configurations, with the outer strike point on the horizontal target (HT) and vertical target (VT) (Chankin et al 2016 Nucl. Mater. Energy , doi: 10.1016/j.nme.2016.10.004). Under conditions (input power, plasma density) where the HT plasma went into the H-mode, a large positive E _r spike in the near scrape-off layer (SOL) was seen in the code output, leading to a very large E  ×  B shear across the separatrix over a narrow region of a fraction of a cm width. No such E _r feature was obtained in the code solution for the VT configuration, where the H-mode power threshold was found to be twice as high as in the HT configuration. It was hypothesised that the large E  ×  B shear across the separatrix in the HT configuration could be responsible for the turbulence suppression leading to an earlier (at lower input power) L–H transition compared to the VT configuration. In the present work these ideas are extended to cover some other experimental observations on the H-mode power threshold variation with parameters which typically are not included in the multi-machine H-mode power threshold scalings, namely: ion mass dependence (isotope H–D–T exchange), dependence on the ion ∇ B drift direction, and dependence on the wall material composition (ITER-like wall versus carbon wall in JET). In all these cases EDGE2D-EIRENE modelling shows larger positive E _r spikes in the near SOL under conditions where the H-mode power threshold is lower, at least in the HT configuration. (paper)

[en] The electron heat transport in low density H-mode plasmas heated by neutral beam injection (NBI) is investigated in ASDEX Upgrade using electron cyclotron heating (ECH) combining both steady-state and transient response analysis by modulating the ECH power. Under these conditions, more than 60% of the NBI power (>3 MW) is delivered to the ions, while approximately 20% (∼1 MW) is delivered to the electrons. In the confinement region, the electron-to-ion temperature ratio, T_e/T_i, varies between 0.5 and 0.7 in the NBI-only phase and between 0.8 and 1.0 when the ECH is also applied. Due to the low collisional coupling, the power in the electron channel is locally more than doubled by applying up to the available 2 MW of ECH, while the power in the ion channel is locally increased by less than 30%. A dependence on the density of the reaction of the plasma parameters to the ECH is observed. For plasmas with average density nbar_e < 4.0 x 10¹⁹ m^-3 (defined as 'hot-ion' H-modes), when the ECH is applied, T_e increases, the central T_i drops and the density flattens. These effects disappear with increasing density and are not observed for nbar_e > 4.0 x 10¹⁹ m^-3 (defined as 'regular' H-modes). Power balance analysis of both the hot-ion and regular H-modes points to a strong resilient behaviour of the T_e profiles. In the hot-ion cases, the ECH heating induces a strong increase in transport in the ion channel. Power balance and transient response analysis of the regular H-modes are consistent with an inverse scale length transport model with a threshold in R/L_Te=R · vertical bar ∇T_e|/T_e, above which the electron heat transport is increased. Comparison with recent studies in pure EC heated L-modes points to a stronger resilience of T_e in the NBI heated H-modes

[en] In the ASDEX Upgrade tokamak, the power deposition structures on the divertor target plates during type-I edge localized modes (ELMs) have been investigated by infrared thermography. In addition to the axisymmetric strike line, several poloidally displaced stripes are resolved, identifying an ELM as a composite of several subevents. This pattern is interpreted as being a signature of the helical perturbations in the low field side edge during the non-linear ELM evolution. Based on this observation, the ELM related magnetic perturbation in the midplane can be derived from the target load pattern. In the start phase of an ELM collapse, average toroidal mode numbers around n ∼ 3-5 are found evolving to values of n ∼ 12-14 during the ELM power deposition maximum. Further information about the non-linear evolution of the ELM mode structure is obtained from statistical analyses of the spatial distribution, heat flux amplitudes and number of single stripes

[en] A pedestal prediction model Europed is built on the existing EPED1 model by coupling it with core transport simulation using a Bohm-gyroBohm transport model to self-consistently predict JET-ILW power scan for hybrid plasmas that display weaker power degradation than the IPB98(y, 2) scaling of the energy confinement time. The weak power degradation is reproduced in the coupled core-pedestal simulation. The coupled core-pedestal model is further tested for a 3.0 MA plasma with the highest stored energy achieved in JET-ILW so far, giving a prediction of the stored plasma energy within the error margins of the measured experimental value. A pedestal density prediction model based on the neutral penetration is tested on a JET-ILW database giving a prediction with an average error of 17% from the experimental data when a parameter taking into account the fuelling rate is added into the model. However the model fails to reproduce the power dependence of the pedestal density implying missing transport physics in the model. The future JET-ILW deuterium campaign with increased heating power is predicted to reach plasma energy of 11 MJ, which would correspond to 11–13 MW of fusion power in equivalent deuterium–tritium plasma but with isotope effects on pedestal stability and core transport ignored. (paper)