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

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

The reproducibility of diamagnetic flux measurements in vacuum field shots for the internal and external diamagnetic flux measurements in the ASDEX Upgrade tokamak is investigated. A comparison of diamagnetic flux measurements and predictions from equilibrium reconstruction, including pressure constraints from thermal and fast-ion contributions, is reported. The ideal magnetohydrodynamic model involved in interpreting diamagnetic flux measurements in a tokamak with anisotropic pressure is summarised. The plasma energy calculated from equilibrium reconstruction and inferred from diamagnetic flux measurements is compared with the respective values calculated by modelling with TRANSP and IDE/RABBIT. It is found that at low densities, in discharges with parallel neutral beam sources, the plasma energy inferred from the diamagnetic flux measurements is smaller than the plasma energy calculated from the equilibrium reconstruction. In addition, details of the modelling are discussed to stress the interdependence of the various modelling inputs and the interpretation of the results. A comparison of diamagnetic flux measurements with calculations from equilibrium reconstruction by IDE and fast-ion profiles from RABBIT and TRANSP simulations can identify anisotropic plasma discharges and provide a consistency check of the input data for modelling.

Internal kink (IK) instability is investigated for European demonstration fusion reactor (EU DEMO) plasmas in both negative triangularity (NT) and positive triangularity (PT) configurations. For NT plasmas, the IK becomes more unstable as an ideal conformal wall moves away from the plasma boundary, with the mode growth rate saturating at the wall radial location of about b / a = 1.5 , where a is the plasma minor radius and b the wall radial location. The plasma resistivity destabilizes the IK mode. The effect of sub-sonic toroidal plasma flow is sufficiently weak and can thus be ignored for these EU DEMO equilibria. These results are consistent with those for PT plasmas, albeit with larger mode growth rate in the NT configuration. Both perturbative and self-consistent magneto-hydrodynamic (MHD)-kinetic hybrid calculations predict (partial) stabilization of the IK modes in both NT and PT configurations, with inclusion of various kinetic contributions. Precessional drift motion of trapped fusion-born alphas in EU DEMO produces weak stabilization to the IK mode. Stronger stabilization occurs with the toroidal precession of trapped thermal particles (ions and electrons) and the bounce-transit motion of thermal ions. The stabilization is similar between the NT and PT configurations, due to the similarity of the mode eigenfunction (occupying a nearly circular region in the plasma core) despite the sign difference in the triangularity. The non-perturbative MHD-kinetic hybrid model predicts much less stabilization of the mode than the perturbative model, primarily due to the self-consistent determination of the mode eigenvalue in the former. Generally, no significant difference in the IK mode stability is found between the NT and PT plasmas in EU DEMO.

Integrated edge-localized-mode (ELM)-free or small ELM scenarios for the demonstration fusion power plant (DEMO) are investigated in ASDEX Upgrade using argon seeding for radiative power removal mainly in the pedestal region. An important aspect is the modification of the electron pressure in the pedestal by the additional radiative power losses. Full ELM suppression could be achieved in a no-ELM H-mode scenario featuring an edge electromagnetic quasicoherent mode up to a heating power of 5 MW, where argon radiation allowed the extension of the heating power operational space. At higher powers up to 12 MW (reaching the beta limit), ELMs of reduced size prevail and detachment is obtained by the argon seeding. Control of the position of a radiating zone localized inside the X-point was found to be favorable compared to the control of the separatrix power for low P s e p / P h e a t . Integration of pedestal cooling with Ar and ELM suppression by resonant magnetic perturbations (RMP) allowed an increase of the core radiation and a partial recovery of normalized confinement to H98 = 1. This favorable behavior is finally limited by the loss of the RMP density pump-out effect, followed by an ELM-free H-mode phase and re-occurrence of ELMs. For an active tailoring of the pedestal pressure profile, precise knowledge of the radiation profile is required. Modeling of the argon radiation profile with the STRAHL code showed good agreement with bolometry only if charge exchange of neutral deuterium with argon ions was taken into account, highlighting the importance of the neutral density in the pedestal region. In view of best integration of a no-ELM scenario with divertor detachment and high heating power, the divertor compression and enrichment of argon and neon were compared using dynamic gas puff experiments. Argon shows more than a factor of 3 higher divertor enrichment compared to neon, but the absolute values decrease with higher neutral deuterium pressure.

The sawtooth activity is investigated for an EU DEMO reference plasma, including kinetic effects from both thermal particles and fusion-born alphas. Kinetic effects are studied in conjunction with modification of the magnetic shear near the q= 1 surface (q is the safety factor) due to local current drive. Kinetic stabilization on the internal kink mode is found by both the non-perturbative MHD-kinetic hybrid code MARS-K (Liu et al 2008 Phys. Plasmas 15 112503) and the perturbative semi-analytic Porcelli model. The latter predicts full stabilization of the mode, when the local magnetic shear s 1 at the q= 1 surface is less than 1 (e.g. s 1 ∼ 0.6 as is the case for the target EU DEMO plasma without local current drive). By increasing the local magnetic shear with local current drive up to s 1 ∼ 2.8, one of the Porcelli sawtooth crash criteria can be satisfied, by accessing the ion-kinetic regime. Direct MARS-K non-perturbative eigenvalue computations, on the other hand, predict less kinetic stabilization of the internal kink and more robust triggering of the sawtooth crash in EU DEMO. The general trend of the predicted results remains similar though, between the MARS-K model and the Porcelli model.

The paper describes under which plasma and machine conditions runaway electrons (REs) are generated during ASDEX Upgrade plasma disruptions. The REs are created by argon injection to investigate methods of current and energy dissipation. The RE beams are stable and last up to a few hundred milliseconds. The experimental findings are described and made available for the validation of theoretical models. In the following, a simple 0D fluid model is used to simulate the observed RE current magnitude and time behavior. In spite of its simplicity, the model is consistent with the measured RE current. The injection of heavy gases into the RE beam is then discussed in some detail: The injected gas penetrates into the low density background plasma and through the beam without ionizing significantly. Therefore its effect on the REs can be tested directly and it is found to be consistent with the known Coulomb collisional theory.

In the present work, the tolerable impurity level and composition for a reactor plasma using several sets of model assumptions are evaluated. Special care was taken to evaluate a comprehensive and consistent set of atomic data for 35 different elements, such that the impurity level for various elements may be studied as a function of their nuclear charge. The data set may not only be useful for the presented work or for system codes which design fusion reactors, but also for interpretation of bolometric measurements. Additionally, the predictions of the spectral distribution of the radiated power is of high quality such that soft x-ray broadband measurements may be interpreted. In the present work the data is used for predicting the radiated power in a reactor plasma, using a 0D, several variants of a 0.5D model and a realistic 1D ASTRA modelling of a DEMO plasma, i.e. the EU DEMO1 2015 design. The maximal or appropriate impurity content of a reactor plasma for all models can be determined, such that the predictions from a simplistic 0D model can be compared to less simplistic models and a proper reactor simulation. These comparisons suggest that with the simplistic models the impurity content may be estimated within a factor of about 1.5, independent of the realization of the reactor plasma. At the same time this study underlines the sensitivity of the reactor performance on the impurity mixture and especially of the He content of the plasma. Additionally, an extended 0.5D model is presented which is able to predict variations of the fusion yield Q and the He concentration, when both is known for a reference scenario. These predictions prove to be of high accuracy when compared to the 1D ASTRA modelling and thus, allow the net impact of an increased dilution and a simlutaneous temperature rise at constant plasma pressure to be evaluated. Furthermore, the parameter space is scanned with more than 105 model reactor plasmas demonstrating that the use of a low-Z impurity diminishes the possibility of an economical feasible reactor plasma. The main results of the parameter scan are made available via scaling formulae.

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, $\chi_{\rm i}$ , 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 $\chi_{\rm i}$ 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.

Electron internal transport barriers (eITBs) are obtained in TCV with different heating and current drive schemes. They are sustained in steady-state conditions for several energy confinement and current redistribution times. In these scenarios, the density profile displays a different behaviour with respect to normal L-mode plasmas, with or without auxiliary heating. In fully non-inductive discharges developing an eITB, the density profile shape is strongly correlated with the electron temperature profile, with the normalized density gradient equal to 0.45 times the normalized temperature gradient, revealing the existence of a significant inward pinch of a thermodiffusive type. The coupling of the two profiles is observed from the 'foot' of the barrier inwards. The effect of small inductive current perturbations on fully non-inductive sustained eITBs shows that ∇n/n is only indirectly coupled to the current profile, through its effect on local confinement, contrary to the standard L-mode.

The rapid formation of an electron internal transport barrier (eITB) is observed during a slow evolution (similar or equal to 200 ms) from a centrally peaked to a hollow current density profile, while all external actuators remain constant. The time constant for the barrier formation appears to be shorter than the electron energy confinement time. The improved confinement associated with the barrier formation occurs first in a localized region off-axis. Then the effects propagate to inner and outer flux surfaces on a confinement time scale. The temporal and spatial localizations of the barrier formation suggest a threshold in the magnetic shear profile that triggers the onset of the eITB. The threshold-like behaviour can be attributed to the sudden appearance of a minimum q (or zero-shear) flux surface in the safety factor profile. Based on modelling of the current profile evolution using the ASTRA transport code, the appearance of the zero-shear flux surface is correlated with the temporal and spatial formation of the barrier.

On the TokamakConfiguration Variable (TCV), electron internal transport barriers (eITBs) can be formed during a gradual evolution from a centrally peaked to a hollow current profile while all external actuators are held constant. The formation occurs rapidly (

A novel method for inverting time-resolved line-integrated interferometric plasma density measurements is described. The method uses singular value decomposition of local density profiles from Thomson scattering measurements obtained at low sampling rates in the same or equivalent plasmas to determine a set of orthogonal spatial basis functions which is well adapted to the physical processes under investigation. The sought-for density profile is expanded into a limited series of these functions and a solution is calculated by using a simple least-squares fit method. The new method overcomes the difficulties encountered with other methods, such as regularization methods, which smoothen gradients and depend on the availability of accurate measurements in the plasma edge region. The small number of computations required provides for a fast algorithm. This method, which combines the high bandwidth of interferometer systems with the spatial accuracy of Thomson scattering, is applied to invert interferometer measurements in a wide variety of operational regimes in the Tokamak a Configuration Variable and Joint European Torus (JET) tokamaks. In particular, the collisionality dependence of density peaking observed in ASDEX Upgrade is confirmed in JET H-modes using this method.

An updated and improved version of the 0D divertor and scrape-off layer (SOL) model published in Siccinio et al (2016 Plasma Phys. Control. Fusion 58 125011) was coupled with the 1.5D transport code ASTRA (Pereverzev 1991 IPP Report 5/42, Pereverzev and Yushmanov 2002 IPP Report 5/98 and Fable et al 2013 Plasma Phys. Control. Fusion 55 124028). The resulting numerical tool was employed for various scans in the major radius R and in the toroidal magnetic field B T—for different safety factors q, allowable loop voltages V loop and H factors—in order to identify the most convenient choices for an electricity producing tokamak. Such a scenario analysis was carried out evaluating self-consistently, and simultaneously, the core profile and transport effects, which significantly impact on the fusion power outcome, and the divertor heat loads, which represent one of the most critical issues in view of the realization of fusion power plants (Zohm et al 2013 Nucl. Fusion 53 073019 and Wenninger et al 2017 Nucl. Fusion 57 046002). The main result is that, when divertor limits are enforced, the curves at constant electrical power output are closed on themselves in the plane, and a maximum achievable power exists—i.e. no benefits would be obtained from a further increase in R and B T once the optimum is reached. This result appears as an intrinsic physical limit for all those devices where a radiative SOL is needed to deal with the power exhaust, and where a lower limit on the power crossing the separatrix (e.g. because of the L–H transition) is present.

Recent improvements to the heating and diagnostic systems on the ASDEX Upgrade tokamak allow renewed investigations into non-inductive operation scenarios with improved confinement in a full-metal device. Motivated by this, a scenario with , and a high non-inductive current fraction has been developed. The scenario offers good confinement with and normalised ion temperature gradients . Moreover, it is robust against resistive magnetohydrodynamic (MHD) instabilities, but does suffer from ideal MHD instability when . To verify the understanding of the plasma transport processes, the heat transport was modelled using TGLF. This revealed that electromagnetic effects at high β and/or from fast ions appear to be missing from TGLF's physics model. As accurate reconstruction of the plasma equilibrium is crucial for studies of advanced scenarios, this publication also documents the presence of polarised background light that can contaminate motional stark effect measurements and thus interfere with equilibrium reconstruction.

A gyro-kinetic analysis of intrinsic rotation is presented for the ASDEX Upgrade tokamak. The gyro-kinetic turbulence code, GKW and the neoclassical transport code, NEO are coupled so that the neoclassical equilibrium distribution function is included in the background distribution function in the gyro-kinetic turbulence simulation. This implementation is benchmarked against a similar implementation in the gyro-kinetic code, GS2 (Dorland et al 2000 Phys. Rev. Lett. 85 5579) and against analytical predictions. A quasi-linear and non-linear gyro-kinetic turbulence analysis is performed on Ohmic L-mode ASDEX Upgrade plasmas showing that the symmetry breaking effects due to neoclassical background flows can produce significant toroidal momentum transport. While its magnitude is of the order of other symmetry breaking mechanisms, such as the Coriolis pinch, up–down asymmetry in the magnetic flux surfaces and flow shear, the flow gradients it can sustain are appreciably smaller than the maximum gradients measured at the mid-radius of the ASDEX Upgrade tokamak core, which can be up to an order of magnitude larger. It is found that the gradient of the diamagnetic flow, and therefore the second derivatives of the density and temperature gradients are critical to the production of residual toroidal momentum flux. A quasi-linear estimate indicated that the second derivatives required to match the experimental flow gradient are up to an order of magnitude higher than the measured second derivatives. This analysis suggests that turbulent transport driven by neoclassical flows is not sufficient to explain the maximum flow gradients observed in ASDEX Upgrade.

The most recent experiments of disruption mitigation by massive gas injection in ASDEX Upgrade have concentrated on small-relatively to the past-quantities of noble gas injected, and on the search for the minimum amount of gas necessary for the mitigation of the thermal loads on the divertor and for a significant reduction of the vertical force during the current quench. A scenario for the generation of a long-lived runaway electron beam has been established; this allows the study of runaway current dissipation by moderate quantities of argon injected. This paper presents these recent results and discusses them in the more general context of physical models and extrapolation, and of the open questions, relevant for the realization of the ITER disruption mitigation system.

As a next-step in the tokamak-based fusion programme, the DEMO fusion reactor is foreseen to produce relevant output electricity, in the order of ∼500 MW delivered to the network. The scenarios that are being presently investigated consist of a pulsed device, called DEMO1, and a steady-state device, called DEMO2. In this work, which is focused on the pulsed device DEMO1, scenarios are studied from the point of view of core transport, to assess plasma performance and limitations due to core microinstabilities. The role of radiated power, aspect ratio, and height of temperature pedestal are assessed as they impact both core energy and particle transport. Open issues in this framework are also discussed.

In systems modelling for fusion power plants, it is essential to robustly predict the performance of a given machine design (including its respective operating scenario). One measure of machine performance is the energy confinement time that is typically predicted from experimentally derived confinement scaling laws (e.g. IPB98(y,2)). However, the conventionally used scaling laws have been derived for ITER which—unlike a fusion power plant—will not have significant radiation inside the separatrix. In the absence of a new high core radiation relevant confinement scaling, we propose an ad hoc correction to the loss power used in the ITER confinement scaling and the calculation of the stored energy by the radiation losses from the 'core' of the plasma . Using detailed ASTRA / TGLF simulations, we find that an appropriate definition of is given by 60% of all radiative losses inside a normalised minor radius . We consider this an improvement for current design predictions, but it is far from an ideal solution. We therefore encourage more detailed experimental and theoretical work on this issue.

International audience; in the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

New or upgraded diagnostics of the edge transport barrier allow investigations of the dominant transport mechanisms in the pedestal. The density build-up after the L–H transition can be explained with a mainly diffusive edge transport barrier. A small inward convection term improves the agreement between modelling and experiment, but its existence cannot be confirmed due to the uncertainty in the neutral sources. Measurements of the impurity ion flow asymmetry as well as the edge current density are in agreement with neoclassical modelling. The inter-ELM pedestal recovery was traced with ideal peeling–ballooning modelling, which shows that the stability boundary moves closer to the operational point as the pedestal becomes wider. Gyrokinetic modelling of the different phases reveal that density gradient driven trapped electron modes are dominant during the early recovery, while electron temperature gradient modes or kinetic ballooning modes determine the temperature gradient in the final phase. Microtearing modes are modelled and also experimentally determined at the top of the pedestal. Non-linear coupling between modes could explain the failure of ideal linear MHD modelling.

A toroidal momentum transport equation is presented, which is suited for implementation in 1D transport codes in axisymmetric geometry (e.g. tokamak plasmas). Contributions to the total toroidal momentum from the different plasma species are retained. The contributions of E × B, diamagnetic and poloidal flows are also retained. The different pieces appearing in the expression for the turbulence-driven part of the flux are obtained consistently from standard gyro-kinetic theory.

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. Fusion 52 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.

The ASDEX Upgrade intrinsic rotation database has been expanded to include a large number of measurements from ohmic L-mode discharges covering a wide range of plasma densities, currents and magnetic fields. This database was then used to study the rotation behaviour across the transition from linear to saturated ohmic confinement (LOC/SOC). At low collisionality the plasma is in the LOC regime and the toroidal rotation profile is in the co-current direction. As the collisionality is increased and the plasma transitions from LOC to SOC, the core rotation decreases resulting in a hollow, counter-current profile. Even deeper in the SOC regime, however, a second reversal back towards the co-current direction occurs. Linear gyrokinetic calculations indicate that these reversals cannot be explained by a transition from a trapped electron mode (TEM) to an ion temperature gradient (ITG) dominated regime. Rather, the analysis indicates that the intrinsic normalized rotation gradient, u′, depends strongly on local plasma parameters, in particular on . Taken together with turbulent particle transport theory, these results suggest that the co- to counter-current directed rotation reversal occurs in the TEM regime due to profile changes and not due to transition from TEM to ITG. The second reversal back to the co-current direction is explained by the reduction of observed in ITG dominated plasmas. Lastly, a simple linear model for the residual stress assuming a fixed poloidal tilt angle of the turbulent eddies was applied to the database. This model is able to capture the main parameter dependencies observed in the data and demonstrates that at least two processes contributing to the poloidal tilt of the turbulent eddies are needed to reproduce the experimentally observed u′ values: one proportional to the sign of the turbulence propagation and the second independent of ωr.

Novel flow rotation measurements based on charge exchange recombination spectroscopy at the inboard midplane of the ASDEX Upgrade tokamak reveal the existence of an asymmetric flow structure at the H-mode edge, which is shown to arise due to a poloidal impurity density asymmetry. A quantitative evaluation of the impurity density at the inboard side demonstrates that the impurities redistribute along the flux surface, resulting in a poloidal dependency of the impurity density. The poloidal and toroidal impurity flows measured at the high-field side (HFS) and low-field side (LFS) are compared to theoretical predictions based on the parallel momentum balance, which includes friction, inertia, pressure and electric force. Both a fluid and a kinetic approach are used, showing good agreement with each other. The measured impurity flow structure is described by the model quantitatively when a finite poloidal main ion flow of ~2 km s−1 arises, which is in keeping with the standard neoclassical prediction. The interplay of all terms, in particular the inclusion of the impurity inertia term, is important in reproducing the observed flow structure and results in an impurity accumulation at the HFS. The existence of a poloidal impurity density asymmetry in the edge transport barrier slightly reduces the drift parameter v/D, however, the experimental value is consistent with standard neoclassical theory. This demonstrates that despite the asymmetry in the impurity density, the impurity particle transport is at the neoclassical level.

The modelling of tokamak scenarios requires the simultaneous solution of both the time evolution of the plasma kinetic profiles and of the magnetic equilibrium. Their dynamical coupling involves additional complications, which are not present when the two physical problems are solved separately. Difficulties arise in maintaining consistency in the time evolution among quantities which appear in both the transport and the Grad–Shafranov equations, specifically the poloidal and toroidal magnetic fluxes as a function of each other and of the geometry. The required consistency can be obtained by means of iteration cycles, which are performed outside the equilibrium code and which can have different convergence properties depending on the chosen numerical scheme. When these external iterations are performed, the stability of the coupled system becomes a concern. In contrast, if these iterations are not performed, the coupled system is numerically stable, but can become physically inconsistent. By employing a novel scheme (Fable E et al 2012 Nucl. Fusion submitted), which ensures stability and physical consistency among the same quantities that appear in both the transport and magnetic equilibrium equations, a newly developed version of the ASTRA transport code (Pereverzev G V et al 1991 IPP Report 5/42), which is coupled to the SPIDER equilibrium code (Ivanov A A et al 2005 32nd EPS Conf. on Plasma Physics (Tarragona, 27 June–1 July) vol 29C (ECA) P-5.063), in both prescribed- and free-boundary modes is presented here for the first time. The ASTRA–SPIDER coupled system is then applied to the specific study of the modelling of controlled current ramp-up in ASDEX Upgrade discharges.

Kinetic edge profiles are measured with high spatial and temporal resolution and their behaviour before and after the time of the L–H transition is analysed. It is shown that at low densities electron and ion channels are strongly decoupled and that the ion kinetic profiles determine the transition to H-mode.At medium densities, where electrons and ions are coupled, the H–L back-transition occurs at the same pressure gradients as the L–H transition. The threshold power for the H–L back-transition exhibits the same increase at low densities as the L–H transition.After the L–H transition the density development depends on the available neutral gas reservoir. Modelling of particle transport indicates that the temporal development of the density profiles can be reproduced assuming only a reduction in the edge diffusion coefficient.The influence of magnetic perturbation coils on the L–H transition power threshold strongly increases with density above a Greenwald factor of about 0.6.

In tokamaks, turbulent particle and toroidal momentum transport are both characterized by the presence of off-diagonal contributions which play an essential role in establishing the profile shapes of the density and the toroidal rotation under most conditions. In this paper similarities and differences between the two turbulent transport channels are pointed out and, thereby, interesting physical aspects which connect the two channels are identified. The main contributions to off-diagonal particle and toroidal momentum transport are reviewed by means of a rather simplified description, which aims at providing, when possible, a direct connection between theoretical, modelling and experimental research.

Measurements of plasma core impurity ion temperature and toroidal rotation are obtained during the current ramp-up phases of plasma discharges in the ASDEX Upgrade tokamak, by means of charge exchange recombination spectroscopy utilizing short neutral beam injected blips. The phenomenology of these profiles as the current is varied is presented. The observations are reported for both ohmically heated and electron cyclotron heating assisted ramps, with auxiliary heating applied both on-axis and off-axis. The evolution of global parameters is also reported. A comparison between ion and electron kinetic profiles evolution is made. Power balance analysis is performed to establish the local confinement properties and their evolution during the current ramp. Transport modelling of several cases with a theory-based model is carried out and a recipe, which involves edge-relevant parameters, is proposed to account for the transport in the outer plasma region. Finally, linear gyrokinetic calculations of dominant microinstabilities allow us to correlate the observed transport with the background turbulence regime.

A detailed analysis of experimental data from the ASDEX Upgrade tokamak is carried out to shed light on the properties of confinement and transport in the current ramp-up and ramp-down phases of the plasma discharge. The experimental database is used to identify the relevant ranges of parameters explored during the ramp-up and the ramp-down. The energy confinement time observed in the two ramps displays interesting evolution, in many cases attaining different values at the same current level between ramp-up and ramp-down. The possible reasons for this behaviour are investigated. Interpretative transport simulations are used as a tool to clarify the interplay between different parameters, which are coupled in a non-linear way. In addition, a theory-based transport model is used to understand the behaviour of confinement as observed in the experiment, evidencing the role of both turbulent and neoclassical transport. Linear gyrokinetic calculations are performed to identify the relevant turbulence regime, showing that a broad range of frequencies, in the trapped electron modes (TEMs) and in the ion temperature gradient modes (ITGs) regimes, is explored during both the ramp-up and ramp-down. In the same framework, a quasi-linear model is applied to calculate the value of the local logarithmic density gradient and compare it with the experimental value. Finally, first non-linear simulations of heat transport during the current ramps are presented.

Interpretative analysis of experimental data from the ASDEX Upgrade tokamak current ramps-up and ramps-down is carried out to shed light on the properties of confinement and transport in these particular phases of the plasma discharge. It is found that the two ramps are similar in parameters evolution, but are not symmetric with respect to the current level, for several reasons that will be elucidated. Theory-based energy transport modeling allows to understand the underlying transport processes at play, although there are limitations in describing the edge part of the plasma at very low currents, in particular during ramp-up. Possible reasons for this discrepancy are discussed. Finally, the relevant turbulence (linear) regime is identified simultaneously with calculations of quasi-linear particle transport, showing that a broad range of modes, from Trapped Electron Modes (TEM) to Ion Temperature Gradient (ITG) modes, is explored during both the ramp-up and ramp-down.

In the Tokamak a Configuration Variable (TCV) (Hofmann F et al1994 Plasma Phys. Control. Fusion 36 B277), global plasma oscillations have been discovered in fully non-inductively driven plasmas featuring electron internal transport barriers (ITB) with strong ECRH/ECCD. These oscillations are linked to the destabilization and stabilization of MHD modes near the foot of the ITB and can lead to large oscillations of the total plasma current and line-averaged density, among others. They are intrinsically related to the fact that ITBs have large pressure gradients in a region of low magnetic shear. Therefore, the ideal MHD limit is relatively low and infernal modes can be unstable. Depending on the proximity to the ideal limit, small crashes or resistive modes can appear which affect the time evolution of the discharge. Being near marginal stability, the modes can self-stabilize due to the modification of the pressure gradient and local q-profile. The plasma recovers good confinement, reverses shear and the ITB builds up, until a new MHD mode is destabilized. TCV results show that this cycling behaviour can be controlled by modifying the current density or the pressure profiles, either with Ohmic current density perturbation or by modifying the ECH/ECCD power. It is demonstrated that many observations such as q ≥ 2 sawteeth, beta collapses, minor disruptions and oscillation regimes in ITBs can be assigned to the same physics origin: the proximity to the infernal mode stability limit.