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Particle balance calculations are done for the seeding species N _2 , Ne, Ar, Kr as well as He with the aim of obtaining a realistic description of the divertor and core plasma impurity content. Experimental time traces of main plasma impurity densities are fitted by a single, time-independent parameter v $_{z,\textrm{in}}^\textrm{eff}$ . This parameter represents the product of the impurity inward pinch in the pedestal, used for the description of gross fueling here, and the enrichment factor between the sub-divertor gas reservoir and the upstream separatrix. v $_{z,\textrm{in}}^\textrm{eff}$ depends strongly on the first ionization energy as well as on the charge Z or mass of the impurity. The prevailing dependence of the enrichment values on the ionization energy suggests the importance of the relative impurity and deuterium ionization lengths in the divertor. Regression analysis of v $_{z,\textrm{in}}^\textrm{eff}$ yields an expression for the impurity concentrations in the core and the divertor of ASDEX Upgrade which allows the prediction of the corresponding impurity densities and their divertor enrichment within a factor of 2 with only engineering parameters as input. A simple wall model has been introduced to take into account wall storage and release of impurities, e.g. for conditions of pre-loaded walls due to seeding in previous discharges. Wall effects are observed for all species considered, but wall storage turns out to be more important for N and He compared to Ne, Ar, Kr. Similar enrichment values are obtained for ELMy H-modes and EDA/QCE no-ELM regimes. A factor of approximately 1.4 reduction in enrichment is observed for divertor conditions for pronounced detachment with Ar and N _2 . The obtained analytical model for the core and sub-divertor impurity densities is well suited for integration into a discharge flight simulator or a real time state observer.

Experiments, gyrokinetic simulations and transport predictions were performed to investigate if a negative triangularity (NT) L-mode option for the Divertor Tokamak Test (DTT) full-power scenario would perform similarly to the positive triangularity (PT) H-mode reference scenario, avoiding the harmful edge localized modes (ELMs). The simulations show that a beneficial effect of NT coming from the edge/scrape-off layer (SOL) region $\rho_{\mathrm{tor}} \gt 0.9$ is needed to allow the actual NT L-mode option to perform like the PT H-mode. Dedicated experiments at TCV and AUG, with DTT-like shapes, show an optimistic picture. In TCV, experiments indicate that even with the relatively small triangularity of the DTT NT scenario, a large beneficial effect of NT comes from the plasma edge and SOL, allowing NT L-modes to outperform PT L-modes with the same power input, reaching the same central pressures as PT H-modes with twice as much applied heating power. For AUG, NT plasmas go into H-mode more easily than for TCV, but always present much smaller pedestals compared with PT plasmas with the same input power, showing a much weaker or absent ELM activity. However, NT has a smaller beneficial effect for AUG than for TCV, with NT pulses outperforming PT pulses with the same input power only for an ECRH-only case with relatively low input power. For the considered AUG cases, PT pulses perform better than NT ones at higher ECRH power or with mixed NBI and ECRH power. Based on this analysis, the NT option is a viable alternative for the DTT full power scenario, providing high performance plasmas with reduced or absent ELMs.

In present-day fusion devices, central wave heating is crucial to avoid core tungsten (W) accumulation. In this work, we present an integrated modelling framework that reproduces the reduction of core W peaking in ASDEX Upgrade experiments when multiple transport channels are self-consistently evolved, emphasizing the effects of wave heating on turbulent and neoclassical W transport. Predictions for the ITER 15 MA baseline are then provided. We show that the core of a reactor is in a different transport regime for W as compared to present-day tokamaks. The challenges introduced by W arise from global radiation losses that can hinder operation in H-mode, instead of local central accumulation.

Future fusion reactors require a safe, steady-state divertor operation. With deep divertor detachment, which is typically induced by impurity seeding, the radiation concentrates in a small region at the X-point or on closed flux surfaces above the X-point. This so-called X-point radiator (XPR) moves further inside the confined region with increasing seeding and the location can be actively controlled.At AUG, the parameter space for operation with an XPR was significantly extended, using active feedback on the XPR location. The XPR is observed in nearly the whole operational space of AUG in the high-densities or high collisionality regime. ELM suppression is consistently observed in all cases where the XPR was moved to a significant height above the X-point.Direct measurements of density and temperature from the region around the XPR using the new divertor Thomson scattering system at AUG indicate that the temperature at the location of the XPR remains high (>30eV) and only the region towards the X-point cools down further. In this cold XPR core, the temperature reduces to about 1eV.An XPR is also observed in TCV by the injection of nitrogen as extrinsic impurity. This highlights that the wall material (W for AUG, C for TCV) or machine size does not play a significant role for the existence of the regime. However, the scenario appears to be less stable in TCV. First experiments show the necessity of an active control for the XPR: Depending on the wall conditions and the nitrogen wall storage, the required nitrogen seeding level to achieve an XPR changes.Both, the low temperatures measured radially outside of the radiation zone at AUG, and the lower stability of the XPR regime at TCV with the presence of carbon are consistent with the predictions of a one-dimensional model of the XPR. However, the model would predict the development of the cold XPR core, and significant radiation at the X-point might already exist before reaching this cold temperature solution.

For gas puff imaging (GPI) studies of turbulent processes in the plasma edge, it is often assumed that the emitted light is mainly proportional to the local electron density. This, however, only holds for low electron densities in the far scrape-off layer. As GPI typically only detects one emission line, data interpretation further inside the plasma (i.e. around the separatrix) is complicated by two effects. First, there is the strongly non-linear connection between plasma temperature and density and the measured line emission. This can lead to incorrect data interpretation such as an artificial doubling of the frequency of periodic plasma perturbations. Second, a shadowing effect in the neutral distribution can hamper the localization of turbulent structures and modes. For this study thermal helium beam data of an ASDEX Upgrade EDA H-mode scenario are analyzed. By comparing these measurements to a synthetic diagnostic, we demonstrate the presence of diagnostic effects and illustrate their origins. It is shown that when line intensity ratios are used for the analysis, instead of emission from individual lines, the underlying plasma dynamics can be well recovered.

This letter reports on the efforts carried out at the ASDEX Upgrade tokamak to integrate I-mode plasmas with pellet fueling and to increase the I-mode Greenwald fraction $f_{\mathrm{GW}}$ , two important requirements for any DEMO operational scenario. For the first time, stationary I-mode plasmas have been achieved with pellet fueling and the core Greenwald fraction has been increased up to 0.8. Larger $f_{\mathrm{GW}}$ were not achieved due to technical constraints rather than to a physics-based limit. Pellet-fueled I-mode plasmas exhibit enhanced core and edge density, while core and edge temperature are reduced. Nonetheless, edge normalized gradients remain I-mode-like, namely shallow for the density and steep for the temperature. The I-mode energy confinement time is found to obey two distinct density dependencies: for $f_{\mathrm{GW}}\lt0.4$ it rises with increasing $f_{\mathrm{GW}}$ , while for $f_{\mathrm{GW}} \gt 0.4$ the energy confinement time plateaus with increasing $f_{\mathrm{GW}}$ , or even decreases with $f_{\mathrm{GW}}$ for the pellet-fueled plasmas. Similarities with Ohmic and L-mode energy confinement time dependency on density are discussed.

An efficient removal of He ash by active pumping in future fusion devices is necessary to avoid fuel dilution and not degrade the core confinement properties. Therefore, a deep understanding of the underlying physics mechanisms is mandatory. Helium exhaust has been experimentally investigated at the ASDEX Upgrade tokamak. This is an ideal test environment, thanks to the ITER-like divertor geometry, an extensive diagnostics coverage and the presence of plasma-facing components made of tungsten. The exhaust efficiency, characterized by the He compression in the divertor, was found to improve with increasing divertor neutral pressure but to degrade with detachment. A multi-reservoir particle balance model was developed to interpret the observed exhaust dynamics, accounting for plasma transport and wall retention. The limited performance of the pumping system and the efficient helium retention capability of the tungsten wall were identified to have the strongest impact on the exhaust dynamics.

Advanced momentum transport analysis is used to study matched hydrogen (H) and deuterium (D) plasmas in the core of ASDEX Upgrade. The aim is to validate gyrokinetic theory and assess a possible isotope dependence. The methodology extracts momentum diffusion, convection, and intrinsic torque as a function of time from experiments employing neutral beam injection (NBI) modulation. H and D plasma scenarios with comparable ion heat fluxes, NBI torque, electron densities, and several dimensionless parameters were designed to highlight any mass dependency. Linear gyrokinetic simulations predict that, for similar background gradients, the Prandtl and pinch numbers should be similar for H and D. This was confirmed by the experimental momentum transport analyses. The assessed intrinsic torques were found to be similar between H and D, co-current directed and located near the outermost region of the plasma core. The strength of the intrinsic torque is correlated with the amplitude of the plasma pressure gradient in the pedestal. Finally, a robust error analysis demonstrates the uniqueness of the parameters obtained together with their uncertainties. Neglecting the intrinsic torque, or its time dependence, systematically distorts the assessed momentum diffusion and convection. This is the first method to separate all three transport mechanisms from experimental data by retaining their time dependencies, that is found to match, quantitatively, the gyrokinetic predictions for Prandtl and pinch numbers, within experimental uncertainties.

A new approach to infer the momentum transport in tokamak core plasmas via perturbation experiments is presented. For the first time, the analysis self-consistently includes all momentum transport components and their time dependencies, which are essential to separate the momentum fluxes and closely match the experiment. The quantitative agreement between the experimentally inferred transport coefficients and the gyrokinetic predictions provides an unprecedented validation. This work shows that the new methodology and gyrokinetic predictions can now be utilized on the route to physics-based prediction of momentum transport in future reactor plasmas.

We study localized edge turbulence in the ASDEX Upgrade tokamak that appears if resonant magnetic perturbations (RMP) are applied to suppress edge localized modes (ELMs) in the high confinement mode. The concomitant density fluctuations are detected by microwave reflectometry at the outboard midplane. Two modes can be distinguished, (a) a broadband fluctuation below a threshold of the RMP field amplitude, and (b) a narrow-band quasi-coherent mode (QCM) above the threshold. The broadband fluctuation is toroidally spread out but disappears at the toroidal position of maximum E × B shear in the gradient region. Temporal and spatial correlation along field lines of the midplane density fluctuation and the divertor particle flux suggests that this mode is producing significant particle transport across the gradient region and into the divertor, hence contributing to the plasma density reduction that is often observed when applying RMP fields (the so-called ‘pump-out’ effect). The QCM is also toroidally localized, its radial extent grows with increasing RMP field amplitude, and leads to further increased divertor particle flux compared to the broadband mode. Our observations suggest that both modes not only play an important role in keeping the plasma density stationary in the absence of ELMs but also to reduce the plasma pressure such that the plasma edge becomes stable against ELMs.

In ASDEX Upgrade with full-tungsten wall, boronization is an important tool to control tungsten sources and allow reliable operation at low collisionality. The duration of the beneficial effects is limited by the erosion of the boron layer, in particular on the tungsten ion-cyclotron antenna limiters. We report results from experiments exploring injection boron-rich powders in tokamak plasmas as a way to replenish boron coating and extend the lifetime of boronization effects. Pure boron and boron nitride powders were introduced gravitationally in plasma discharges at rates up to 60 mg/s for durations up to 3 s. Boron injection with cumulative amounts >100 mg appeared to improve wall conditions similarly to boronization, with indications of reduced influx of oxygen and tungsten from the antenna limiter. Moreover, cumulative injection of 340 mg of B appeared sufficient to enable successful execution of low collisionality scenarios critically relying on freshly boronized walls.

An overview of results of negative triangularity (NT) studies from the ASDEX Upgrade tokamak (AUG) is given. Moderate values of the triangularity in the range of δ _average ≈ −0.2 have been obtained. Results from both unfavorable and favorable magnetic configuration in terms of high-confinement mode (H-mode) access are presented. In unfavorable configuration, the ion grad B drift points away from the active x-point, which usually results in a factor of 2 higher power thresholds to enter H-mode. In unfavorable configuration, low-confinement mode (L-mode) operation is confirmed at high auxiliary heating input powers $ > 10$ MW at low collisionality. Under these conditions, plasmas with positive triangularity (PT) are usually in H-mode. Highest plasma energies have been obtained using a combination of neutral beam injection, electron- and ion cyclotron resonance heating, where values of the plasma beta of β _N ≈ 1.7 have been reached in L-mode. While discharges with mixed auxiliary heating power input show moderate confinement on L-mode levels, plasmas heated with electron cyclotron resonance heating show H-mode like confinement. Operation in favorable configuration, however, leads to H-mode with type-I edge localized modes (ELMs). The power required for the L–H transition is comparable to the typical L–H power threshold in PT cases ( $< 2.6$ MW), and good confinement of typical PT H-modes is obtained. The ELM frequency is demonstrated to reach 700 Hz. These results show a clear effect of the NT configuration on the ELM behavior in H-modes, and that on AUG, unfavorable configuration is an essential ingredient if H-mode operation is to be avoided in NT plasmas. So far, in all AUG NT plasmas, power degradation is observed. Initial gyrokinetic simulations of the edge plasma in unfavorable configuration reveal the ion-temperature gradient mode as the fastest growing mode with a subdominant trapped-electron-mode.

Previous studies (e.g. (Kappatou 2019 Nucl. Fusion59 056014)) have shown that discrepancies exist between experimental and modelled light impurity peaking in specific regimes. In particular, in NBI heated plasmas, the strong hollowness of boron profiles are not captured by the modelling. In this context, a dedicated ASDEX Upgrade discharge, including boron and helium density measurements, is studied where such discrepancies appear. Emphasis is given to the impact of fast ions on the turbulent and neoclassical impurity transport computed with the codes GKW and NEO, respectively. It is shown that fast ions increase the turbulent outward impurity convection via an increase of both thermo- and roto-diffusion. This increase is correlated with the stabilisation of the ion temperature gradient driven mode (ITG). Additionally, the ratio of the turbulent impurity diffusion over the ion heat diffusion is reduced with fast ions, which in turn, increases the neoclassical contributions to the impurity peaking. Finally, neoclassical convection is also enhanced through changes in the ion density gradient due to the modification of the quasi-neutrality condition in the presence of peaked fast ion profiles. The combination of these three mechanisms induced by the fast ion population has a significant impact on the development of hollow boron profiles, up to experimental levels. [ABSTRACT FROM AUTHOR]

The injection of boron (B) and boron nitride (BN) powders into ASDEX Upgrade H-mode discharges have demonstrated effective control of tungsten influx in low density/collisionality operational regimes, similar to conventional boronization methods. Sub-mm powder particles are gravitationally accelerated into the upper edge of a lower single null H-mode plasma with a boundary shape roughly conforming to the shape of the poloidal midplane limiters. Visible spectroscopy measurements at one of the outer limiter showed increases in both B and N signal levels, as well as elevated B levels in the divertor, and an increase in total radiated power by greater than a factor of two during BN injection. Globally the BN injection improved energy confinement by 10%–20%, associated with improvements in pedestal performance similar to gaseous N injection. Following conditioning discharges with B powder injection, three low gas-fueling discharges with magnetic perturbations for ELM suppression were successfully conducted. These first results suggest that the application of B containing powders can be used to both improve plasma performance in real-time, and to improve overall wall conditions for subsequent discharges. [ABSTRACT FROM AUTHOR]

A study of the properties of the turbulence-driven ion and electron heat fluxes, is presented. Dedicated H-mode experiments taking advantage of the on-axis and off-axis possibilities of both neutral beam injection and electron cyclotron resonance heating available on the ASDEX Upgrade tokamak were carried out. The experimental results are interpreted by comparisons with gyrokinetic calculations. Ion heat transport is, as expected, driven by the ion temperature gradient (ITG) instability with the features predicted by theory: increase of the driven heat flux above a threshold in normalised gradient. In addition the main effects known to impact on the stability of the turbulence, temperature ratio and fast ions population, are exhibited by the experimental results and agree with the gyrokinetic calculations. It is known that the ITG also contributes to the electron heat flux, but that an electron instability can be required in addition when the ITG contribution does not drive the whole imposed electron heating. This situation, investigated by adding electron cyclotron heating, indicates that in the experiments presented here the electron temperature gradient instability develops. [ABSTRACT FROM AUTHOR]

The design and optimization of a future tokamak fusion reactor, from the point of view of the plasma performance and output fusion power, requires the understanding of the underlying physics and the validation on present experiments of the theory-based tools used for the predictions. Present experimental efforts are devoted to approach reactor-relevant parameters (collisionality, , normalized heat fluxes and sources, temperatures ratio) as much as possible. In this work a series of discharges performed on ASDEX Upgrade are presented, where plasma density and auxiliary power levels are scanned to cover a parameter space that moves towards reactor relevant parameters. On this dedicated discharge dataset, a first-principle-based model is then applied for validation. The degree of agreement between code results and experimental measurements is shown and discussed. In cases where discrepancy is found, possible causes are identified. It is shown that the employed modeling tool can predict the overall trend, consistent with more fundamental theoretical considerations, although quantitative extrapolation still has to be performed with care. Key results of this work show that the density peaking is mainly sustained by turbulence, with a minor role for the fueling source, and how this experimental demonstration is important to predict future reactor density profiles. Moreover, the role of electromagnetic effects is also pointed out. [ABSTRACT FROM AUTHOR]

The presence of helium is fundamentally connected to the performance of a fusion reactor, as fusion-produced helium is expected to heat the plasma bulk, while He ‘ash’ accumulation dilutes the fusion fuel. An understanding of helium transport via experimentally validated theoretical models of the low-Z impurity turbulent transport is indispensable to predict the helium density profile in future fusion devices. At ASDEX Upgrade, detailed, multi-species investigations of low-Z impurity transport have been undertaken in dedicated experiments, resulting in an extensive database of helium and boron density profiles over a wide range of parameters relevant for turbulent transport (normalised gradients of the electron density, the ion temperature, and the toroidal rotation profiles, the collisionality and the electron to ion temperature ratio). Helium is not found to accumulate in the parameter space investigated, as the shape of the helium density profile follows largely that of the electron density. Helium is observed to be as peaked as the electron density at high electron cyclotron resonance heating fraction, and less peaked than the electron density at high neutral beam heating fraction. The boron density profile is found to be consistently less peaked than the electron density profile. Detailed comparisons of the experimental density gradients of both impurities with quasilinear gyrokinetic simulations have shown that a qualitative agreement between experiment and theory cannot always be obtained, with strong discrepancies observed in some cases. [ABSTRACT FROM AUTHOR]

Experiments were performed that aimed to control the plasma isotopic mixture by delivering supplemental components into the plasma via pellet injection. Pellets, mm-sized solid bodies of frozen fuel, are either formed from a mixture of different hydrogen isotopes or from deuterium (D) with added elements. Mixed HD pellets were produced reliably and reproducibly providing the required 1:1 ratio of the protium (H) and D, mimicking the situation in a future fusion reactor with D and tritium (T). By applying pellet fuelling, control of the plasma isotope composition was established while gaining access to the reactor-relevant high-density regime. However, the approach was mainly to demonstrate the controlling capability. Yet, experiments provided information on the impact of the isotopic composition on the plasma performance. There is a strong indication that adding H to a D reference plasma causes a reduction of both the particle and energy confinement. The deleterious impact of H on the particle confinement is, apparently, even stronger than on the energy confinement. Furthermore, small amounts of H can cause a loss of energy confinement that is stronger than anticipated by the H98(y,2) scaling, which is usually employed in reactor design studies. In the course of this experiment, pellets were demonstrated to be an effective and powerful tool for detailed investigations of the isotope effect. A first test demonstrated that producing and launching pellets that contain a mixture of different elements is a feasible technique. The latter was achieved by doping a D host pellet despite technical reasons limiting the permissible amount of processed N. [ABSTRACT FROM AUTHOR]

Access conditions for full suppression of edge localised modes (ELMs) by magnetic perturbations (MP) in low density high confinement mode (H-mode) plasmas are studied in the ASDEX Upgrade tokamak. The main empirical requirements for full ELM suppression in our experiments are: 1. The poloidal spectrum of the MP must be aligned for best plasma response from weakly stable kink-modes, which amplify the perturbation, 2. The plasma edge density must be below a critical value, m−3. The edge collisionality is in the range (ions) and (electrons). However, our data does not show that the edge collisionality is the critical parameter that governs access to ELM suppression. 3. The pedestal pressure must be kept sufficiently low to avoid destabilisation of small ELMs. This requirement implies a systematic reduction of pedestal pressure of typically 30% compared to unmitigated ELMy H-mode in otherwise similar plasmas. 4. The edge safety factor q95 lies within a certain window. Within the range probed so far, , one such window, has been identified. Within the range of plasma rotation encountered so far, no apparent threshold of plasma rotation for ELM suppression is found. This includes cases with large cross field electron flow in the entire pedestal region. [ABSTRACT FROM AUTHOR]

We measure the deuterium density, the parallel drift velocity, and parallel and perpendicular temperatures (, ) in non-Maxwellian plasmas at ASDEX Upgrade. This is done by taking moments of the ion velocity distribution function measured by tomographic inversion of five simultaneously acquired spectra of -light. Alternatively, we fit the spectra using a bi-Maxwellian distribution function. The measured kinetic temperatures ( keV, keV) reveal the anisotropy of the plasma and are substantially higher than the measured boron temperature (7 keV). The Maxwellian deuterium temperature computed with TRANSP (6 keV) is not uniquely measurable due to the fast ions. Nevertheless, simulated kinetic temperatures accounting for fast ions based on TRANSP ( keV, keV) are in excellent agreement with the measurements. Similarly, the Maxwellian deuterium drift velocity computed with TRANSP (300 km s−1) is not uniquely measurable, but the simulated kinetic drift velocity accounting for fast ions agrees with the measurements (400 km s−1) and is substantially larger than the measured boron drift velocity (270 km s−1). We further find that ion cyclotron resonance heating elevates and each by 2 keV without evidence for preferential heating in the spectra. Lastly, we derive an expression for the 1D projection of an arbitrarily drifting bi-Maxwellian onto a diagnostic line-of-sight. [ABSTRACT FROM AUTHOR]

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

A newly installed core charge exchange recombination spectroscopy (CXRS) diagnostic at ASDEX Upgrade (AUG) enables the evaluation of the core poloidal rotation () through the inboard-outboard asymmetry of the toroidal rotation with an accuracy of 0.5 to 1 km . Using this technique, the total plasma flow has been measured in Ohmic L-mode plasmas across the transition from the linear to saturated ohmic confinement (LOC–SOC) regimes. The core poloidal rotation of the plasma around mid-radius is found to be always in the ion diamagnetic direction, in disagreement with neoclassical (NC) predictions. The edge rotation is found to be electron-directed and consistent with NC codes. This measurement provides as well the missing ingredient to evaluate the core velocity () from data only, which can then be compared to measurements of the perpendicular velocity of the turbulent fluctuations () to gain information on the turbulent phase velocity (). The non neoclassical from CXRS leads to good agreement between and indicating that is small and at similar values as found with gyrokinetic simulations. Moreover, the data shows a shift of in the ion-diamagnetic direction at the edge after the transition from LOC to SOC consistent with a change in the dominant turbulence regime. The upgrade of the core CXRS system provides as well a deeper insight into the intrinsic rotation. This paper shows that the reversal of the core toroidal rotation occurs clearly after the LOC–SOC transition and concomitant with the peaking of the electron density. [ABSTRACT FROM AUTHOR]

The edge ion heat transport is analyzed in ASDEX Upgrade (AUG) by combining a comprehensive set of pedestal measurements with both interpretive and predictive modelling. The experimentally determined ion heat diffusivities, , are compared with neoclassical theory and the impact of edge localized modes (ELMs) on the edge ion heat transport level is studied in detail. Pedestal matching experiments in deuterium and hydrogen plasmas show that the inter-ELM pedestal remains close to the neoclassical value. The additional power needed in hydrogen to get similar pedestal temperatures as in deuterium plasmas mostly affects the electron heat channel, i.e. the electron heat diffusivity increases while the ion heat diffusivity stays at the same level within the uncertainties. Sub-ms measurements of the edge ion temperature allows us to extend the analysis to the entire ELM cycle. During the ELM crash, the ion heat transport is increased by an order of magnitude. The perturbed heat flux increases first at the separatrix, i.e. first the separatrix ion temperature increases, leading to a flatter ion temperature gradient, followed by a decrease of the whole pedestal profile. The ion heat transport returns to its pre-ELM neoclassical level 3–4 ms after the ELM crash. [ABSTRACT FROM AUTHOR]

Modeling of the linear-to-saturated ohmic confinement transitions is performed on the ohmic discharges database collected at ASDEX Upgrade. Ion and electron temperatures and densities are predicted using the TGLF turbulence transport model. Self-consistent classical energy equipartition is retained, as well as the impact of low-Z impurities, the core concentration of which is largest at the lowest plasma densities and decreases with increasing electron density. The simulation results are then compared to the experimentally estimated confinement time behavior, and locally to the energy diffusivities behavior across the density ramps. Both density and plasma current are scanned. The key result, which somewhat confirms previous studies, is that the ion energy transport channel is dominated by the ITG instability, which becomes stronger at higher densities due to the reduction in impurity content (as such, reduction in dilution), as well as a moderate increase in ion temperature normalized gradient. A direct impact of the transition between electron modes (TEM) and ITG on the LOC–SOC is not observed. However, this is necessary to explain the behavior of the density profile. The linear phase of the confinement is simply attributed to the decrease of electron heat transport as the electron and ion temperatures become more coupled. [ABSTRACT FROM AUTHOR]

In ASDEX Upgrade (AUG), the normalised gyroradius was varied via a hydrogen isotope scan while keeping other dimensionless parameters constant. This was done in L-mode, to minimise the impact of pedestal stability on confinement. Power balance and perturbative transport analyses reveal that the electron heat transport is unaffected by the differences in isotope mass. Nonlinear simulations with the Gene code suggest that these L-mode discharges are ion temperature gradient (ITG) dominated. The different gyroradii due to the isotope mass do not necessarily result in a change of the predicted heat fluxes. This result is used in simulations with the Astra transport code to match the experimental profiles. In these simulations the experimental profiles and confinement times are reproduced with the same transport coefficients for hydrogen and deuterium plasmas. The mass only enters in the energy exchange term between electrons and ions. These numerical observations are supported by additional experiments which show a lower ion energy confinement compared to that of the electrons. Additionally, hydrogen and deuterium plasmas have a similar confinement when the energy exchange time between electrons and ions is matched. This strongly suggests that the observed isotope dependence in L-mode is not dominated by a gyroradius effect, but a consequence of the mass dependence in the collisional energy exchange between electrons and ions. [ABSTRACT FROM AUTHOR]

Tearing mode formation after internal crash events like sawteeth or fishbones is one of the most important MHD processes that results in a big island structure and associated confinement degradation. The process implies magnetic reconnection at the rational surface, which has been investigated in great detail in the ASDEX Upgrade tokamak. Using direct local measurements, it is found that the crash leads to the formation of an ideal kink mode with large saturated amplitude at the resonant surface immediately after the sawtooth crash. This kink mode transforms into a tearing mode on a much longer timescale than the crash itself. The ideal kink mode, formed at the resonant surface after the crash, provides the driving force for the magnetic reconnection. The conversion of the ideal kink mode into a tearing mode after the internal crash is similar for various values of plasma rotation and normalized pressure. [ABSTRACT FROM AUTHOR]

The ion heat transport in the pedestal of H-mode plasmas is investigated in various H-mode discharges with different pedestal ion collisionalities. Interpretive modelling suggests that in all analyzed discharges the ion heat diffusivity coefficient, , in the pedestal is close to the neoclassical prediction within the experimental uncertainties. The impact of changing the deposition location of the electron cyclotron resonance heating on the ion heat transport has been studied. The effect on the background profiles is small. The pre-ELM (edge localized modes) edge profiles as well as the behaviour of the electron temperature and density, ion temperature and impurity toroidal rotation during the ELM cycle are very similar in discharges with on- and off-axis ECRH heating. No significant deviation of from neoclassics is observed when changing the ECRH deposition location to the plasma edge. [ABSTRACT FROM AUTHOR]

The Type I ELM behavior in ASDEX Upgrade with full W plasma facing components is studied in terms of time scales and energy losses for a large set of shots characterized by similar operational parameters but different nitrogen seeding rate and input power. ELMs with no nitrogen can have two typical behaviors, that can be classified depending on their duration, the long and the short ELMs. The work shows that both short and long ELMs have a similar first phase, but the long ELMs are characterized by a second phase with further energy losses. The second phase disappears when nitrogen is seeded with a flux rate above 1022 (e s−1). The phenomenon is compatible with a threshold effect. The presence of the second phase is related to a high divertor/scrape-off layer (SOL) temperature and/or to a low pedestal temperature. The ELM energy losses of the two phases are regulated by different mechanisms. The energy losses of the first phase increase with nitrogen which, in turn, produce the increase of the pedestal temperature. So the energy losses of the first phase are regulated by the pedestal top parameters and the increase with nitrogen is due to the decreasing pedestal collisionality. The energy losses of the second phase are related to the divertor/SOL conditions. The long ELMs energy losses increase with increasing divertor temperature and with the number of the expelled filaments. In terms of the power lost by the plasma, the nitrogen seeding increases the power losses of the short ELMs. The long ELMs have a first phase with power losses comparable to the short ELMs losses. Assuming no major difference in the wetted area, these results suggest that (i) the nitrogen might increase the divertor heat fluxes during the short ELMs and that (ii) the long ELMs, despite the longer time scale, are not beneficial in terms of divertor heat loads. [ABSTRACT FROM AUTHOR]

The I-mode is a plasma regime obtained when the usual L-H power threshold is high, e.g. with unfavourable ion direction. It is characterised by the development of a temperature pedestal while the density remains roughly as in the L-mode. This leads to a confinement improvement above the L-mode level which can sometimes reach H-mode values. This regime, already obtained in the ASDEX Upgrade tokamak about two decades ago, has been studied again since 2009 taking advantage of the development of new diagnostics and heating possibilities. The I-mode in ASDEX Upgrade has been achieved with different heating methods such as NBI, ECRH and ICRF. The I-mode properties, power threshold, pedestal characteristics and confinement, are independent of the heating method. The power required at the L-I transition exhibits an offset linear density dependence but, in contrast to the L-H threshold, depends weakly on the magnetic field. The L-I transition seems to be mainly determined by the edge pressure gradient and the comparison between ECRH and NBI induced L-I transitions suggests that the ion channel plays a key role. The I-mode often evolves gradually over a few confinement times until the transition to H-mode which offers a very interesting situation to study the transport reduction and its link with the pedestal formation. Exploratory discharges in which n  =  2 magnetic perturbations have been applied indicate that these can lead to an increase of the I-mode power threshold by flattening the edge pressure at fixed heating input power: more heating power is necessary to restore the required edge pressure gradient. Finally, the confinement properties of the I-mode are discussed in detail. [ABSTRACT FROM AUTHOR]

The impact of lithium (Li) on plasma performance was investigated at the ASDEX Upgrade tokamak, which features a full tungsten wall. Li pellets containing 1.6  ×  1020 Li atoms were launched with a speed of 600 m s−1 to achieve deep penetration into the plasma and minimize the impact on the first wall. Homogeneous transient Li concentrations in the plasma of up to 15% were established. The Li sustainment time in the plasma decreased with an increasing heating power from 150 to 40 ms. Due to the pellet rate being restricted to 2 Hz, no Li pile-up could take place. No significant positive impact on plasma properties, as reported from other tokamak devices, could be found; the Li pellets rather caused a small reduction in plasma energy, mainly due to enhanced radiation. Due to pellet injection, a short-lived Li layer was formed on the plasma-facing components, which lasted a few discharges and led to moderately beneficial effects during plasma start-up. Most pellets were found to trigger type-I ELMs, either by their direct local perturbation or indirectly by the altered edge conditions; however, reliability was less than 100%. [ABSTRACT FROM AUTHOR]

This paper describes joint ITPA studies of the I-mode regime, which features an edge thermal barrier together with L-mode-like particle and impurity transport and no edge localized modes (ELMs). The regime has been demonstrated on the Alcator C-Mod, ASDEX Upgrade and DIII-D tokamaks, over a wide range of device parameters and pedestal conditions. Dimensionless parameters at the pedestal show overlap across devices and extend to low collisionality. When they are matched, pedestal temperature profiles are also similar. Pedestals are stable to peeling–ballooning modes, consistent with lack of ELMs. Access to I-mode is independent of heating method (neutral beam injection, ion cyclotron and/or electron cyclotron resonance heating). Normalized energy confinement H98,y2  ⩾  1 has been achieved for a range of 3  ⩽  q95  ⩽  4.9 and scales favourably with power. Changes in turbulence in the pedestal region accompany the transition from L-mode to I-mode. The L–I threshold increases with plasma density and current, and with device size, but has a weak dependence on toroidal magnetic field BT. The upper limit of power for I-modes, which is set by I–H transitions, increases with BT and the power range is largest on Alcator C-Mod at B  >  5 T. Issues for extrapolation to ITER and other future fusion devices are discussed. [ABSTRACT FROM AUTHOR]

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 Ip /BT  =  1 MA/2.5 T, total input power Ptot ~ 12 MW and normalized pressure of βN ~ 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 Zeff 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. [ABSTRACT FROM AUTHOR]

The poloidal and toroidal impurity flows in the edge transport barrier of H-mode plasmas have been studied over a wide range of pedestal top ion collisionalities. A comparison of the edge poloidal rotation measurements to neoclassical predictions shows good agreement in all cases. The measured edge impurity toroidal rotation is observed to change sign from co-current to counter-current with decreasing collisionality. The switch occurs at the same collisionality at which neoclassical theory predicts the main ion poloidal rotation to change from the electron to the ion diamagnetic direction. The behaviour of these two species, when used to calculate the main ion toroidal rotation via the radial force balance equation, leads to fairly constant co-current main ion toroidal rotation. Hence, at low collisionality, due to a reduced frictional coupling, the main ion-impurity differential rotation can be quite large. The behaviour of impurity ion flows on a flux surface has also been investigated in detail and it was found that the measurements are consistent with the continuity equation only if the poloidally asymmetric impurity density distribution is taken into account. The asymmetry is found to be the result of the interplay of all forces in the parallel momentum balance, with the friction force providing the dominant drive. Close to the separatrix the poloidal centrifugal force, which is usually neglected, also gives an additional contribution to the impurity density asymmetry. Within the experimental uncertainties the ion temperature and the electrostatic potential are simultaneous flux functions, despite the presence of a poloidally asymmetric impurity density profile. [ABSTRACT FROM AUTHOR]

The neutral beam current drive efficiency has been investigated in the ASDEX Upgrade tokamak by replacing on-axis neutral beams with tangential off-axis beams. A clear modification of the radial fast-ion profiles is observed with a fast-ion D-alpha diagnostic that measures centrally peaked profiles during on-axis injection and outwards shifted profiles during off-axis injection. Due to this change of the fast-ion population, a clear modification of the plasma current profile is predicted but not observed by a motional Stark effect diagnostic.The fast-ion transport caused by MHD activity has been studied in low collisionality discharges that exhibit strong modes. In particular due to sawtooth crashes, significant radial redistribution of co-rotating fast-ions is observed which can very well be described by the Kadomtsev model. In addition, first tomographic reconstructions of the central 2D fast-ion velocity space in the presence of sawtooth crashes allow the investigation of the pitch dependence of the mode-imposed redistribution: a stronger redistribution of mainly co-rotating fast ions is observed than of those with smaller pitch values. [ABSTRACT FROM AUTHOR]

The influence of electron heating compared to ion heating on plasma performance has been analysed in order to make valid projections towards future devices. The capabilities of the newly upgraded electron cyclotron resonance heating system at ASDEX Upgrade make this analysis feasible by replacing neutral beam injection. Dominantly electron heated plasmas are analysed and compared to dominantly ion heated plasmas. It is investigated if they behave systematically different or if the change of heated species is fully compensated by heat exchange from electrons to ions. Studies of plasmas at high collisionalities are presented in Sommer et al (2012 Nucl. Fusion52 114018). Here, these former investigations are extended towards lower collisionalities. The global plasma parameters show a slight reduction with increasing electron heating arising from a significant decrease of the ion temperature, whereas the electron temperature profile is unchanged. The density profile shows a strong peaking which remains unchanged when modifying the heating mix. The power balance analysis shows an important impact of the heat exchange between electrons and ions. The electron and ion temperatures and the plasma density are modelled with the transport model TGLF. The experimental observations are reproduced verifying the applied code. Linear gyrokinetic calculations with GS2 found the ion temperature gradient mode to be the dominant microinstability in all analysed cases. [ABSTRACT FROM AUTHOR]