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Electron cyclotron resonance heating (ECRH) can drive large current densities through electron cyclotron current drive (ECCD). ECCD is expected to be crucial for high-performance plasmas in future fusion reactors like ITER and DEMO, making the current drive efficiency of ECCD a critical design parameter for future reactors. In present-day devices, good agreement between measured and predicted current drive efficiency has been found. However, to ensure the reliability in future machines, a direct validation of the electron momentum distribution function is needed. As a first step towards this goal, we present in this paper oblique ECE measurements of a low-density plasma in the ASDEX Upgrade tokamak. Two oblique ECE diagnostics are used to allow the simultaneous measurements of electrons streaming co- and counter-directionally with the plasma current. Predictions for the distribution function are computed with the bounce-averaged Fokker-Planck code RELAX (E. Westerhof et al., Rijnhuizen report,1992). To allow direct comparison with the measurements, synthetic radiation temperatures are computed with the code ECRad (S. Denk et al., Computer Physics Communications, p. 107175, 2020). Good agreement is found if radial transport occurring predominantly at low electron energies is included. We demonstrate that the oblique ECE diagnostics measure the electron distribution function directly at the ECRH deposition site in phase space. Furthermore, they are sensitive to the abundance of pitch-angle scattered electrons that reduce the ECCD efficiency. Limitations and uncertainties of the measurements and the modeling are discussed.

In this paper numerical simulations with COREDIV code of JET discharges with ITER-like wall are presented. We concentrate on the JET H-mode and hybrid scenarios with nitrogen seeding and all simulations have been performed with the same transport model and only the discharge input parameters like auxiliary heating Paux, line average plasma density nline, confinement factor H98, nitrogen input flux ΓpuffN were changed in the calcu lations. The separatrix density nsepe is an input parameter in our model and has been kept equal to 0.4÷0.5nlinein the simulations, with the recycling coefficient adjusted accordingly. It has been shown that COREDIV is able to reproduce basic parameters of nitrogen seeded discharges for both H-mode and hybrid scenarios. We have achieved reasonable agreement with global plasma parameters like radiations levels, Zeff and tungsten concentrations and the plasma profiles, including density, temperature and radiation are in very good agreement with experimental data. The agreement of the code results with the experimental data might be even better, if the simulations are further tuned taking into account uncertainties to the sputtering model, separatrix density or SOL transport. Simulations show that the observed Zeff level is defined mostly by the low Z impurity content, Be and N2 in the considered shots. It has been found that the tungsten radiation plays always very important role and can not be mitigated even by strong influx of nitrogen. (© 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Long term fuel retention experiments have been performed in JET with the ITER- Like Wall (JET-ILW) and compared with reference discharges in the preceding phase with carbon wall (JET-C) in ohmic, L and H-mode plasma scenarios. The long term fuel retention is evaluated through global gas balance with an accuracy of 1.2% for series of repetitive pulses (10–25) carried out over a full day of experiments. Compared to carbon wall, with the JET-ILW, for L mode, the retention exhibits also a drop from 1.26 × 1021 Ds−1 to 4–8 × 1019 Ds−1. Finally for Type III and type I ELMy H-mode, the retention decreases from 1.37 × 1021 Ds−1 to 7.2 × 1019 Ds−1 and from 1.7 × 1021 Ds−1 to 2.7 × 1020 Ds−1 respectively. The retention rates with the JET-ILW exhibit a decrease by a factor of about 10 compared to JET-C attributed to a reduction of carbon impurities and less fuel content in Be codeposition.

Internal diamagnetic flux measurements, with measurement loops and compensation magnetic probes inside the vacuum vessel, are now available on the ASDEX Upgrade tokamak. The measured diamagnetic flux is compared to that predicted by simulations and calculated from equilibrium reconstruction. The diamagnetic flux measured at 2 positions separated toroidally by 180° in the vacuum vessel is compared.

Simulations of motional Stark effect (MSE) spectra for Wendelstein 7-X (W7-X) are reported. While contributing to experimental equilibrium reconstruction in general, the measurements aim at the detection of small deviations of vacuum fields, such as the diamagnetic effect. For the diagnostic beam on W7-X, the expected plasma β implies a diamagnetic effect as low as 3%. A data model for the measurement has been developed for the application to W7-X. The model includes polarisation effects and the beam attenuation to estimate the required signal dynamics. The model has been applied to ASDEX Upgrade data (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Recent studies at ASDEX Upgrade aim to further characterize and understand the physics of the improved H-mode scenario. The main focus is on the influence of the ramp-up phase of the plasma current and heating on energy confinement and MHD-activity during the subsequent flat-top phase. Depending on the ramp-up scenario two different stationary plasmas can be generated, which show different equilibrated current profiles, although external control parameters are the same in the flat-top phase. The difference of the current profiles in the flat-top phase seems to be due to different MHD-modes. These MHD-modes set in during relaxation of the current profile, which itself depends on the ramp-up scenario. Also the stored energy is different in the two cases as is the peaking of the temperature profiles.Three mechanisms seem to play a role in linking the observed changes in MHD-behaviour and current profile to the changes of the kinetic profiles: the increased transport due to the MHD-modes themselves, the variation of the ratio of magnetic shear s to safety factor q, which modifies the critical temperature gradient length for the onset of ion temperature gradients, and effects on the H-mode pedestal pressure.

Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

Ion cyclotron resonance frequency (ICRF) heating has been an essential component in the development of high power H-mode scenarios in JET-ILW. The steps that were taken for the successful use of ICRF heating in terms of enhancing the power capabilities and optimizing the heating performance in view of core impurity mitigation in these experiments will be reviewed.

High confinement and stability are obtained simultaneously in stationary conditions in improved H-mode discharges at ASDEX Upgrade. The improved H-mode discharges are typically composed of two different phases: 'lower heating phase', where H98(y, 2) is similar to standard H-modes (H98(y, 2) ~ 1), and 'fully developed improved H-mode phase', where H98(y, 2) is higher than standard H-modes (H98(y, 2) up to 1.4). In this paper, the confinement physics is studied by comparing these two different phases using ASTRA simulations. The results are compared with experimental observations. Firstly, the time evolution of the q-profile is simulated with ASTRA using experimental kinetic profiles and compared with experimental measurements. Secondly, the two different phases are compared by power balance analyses and predictive modelling using the Weiland model with ASTRA. Ion temperature gradient lengths and turbulence levels measured by reflectometry are compared between the two phases. Lastly, the roles of density peaking and pedestal pressure in confinement improvement are discussed. The transport analyses using the ASTRA code and experimental observations show that the fully developed improved H-mode phase is very similar to the lower heating phase in terms of core heat transport. With respect to the confinement improvement, it is suggested that the increase of the edge pedestal pressure plays an important role. Enhancement of the radial electric field at the edge is believed to be closely linked with confinement improvement at the edge pedestal region in the fully developed phase.

Empirical estimates indicate that the transient divertor power load induced by type-I ELMs in the standard H-mode ITER scenario might result in intolerably short target life times. Significant experimental effort has been put on the exploration of recipes to avoid or at least mitigate the ELM size by external intervention. Experimental ELM mitigation studies (and related modeling) are of interest from different points of views. First, the active intervention gives access to new, externally controlled ELMy discharge scenarios and ELM diagnostic techniques. Second, the ELM response on the type and amplitude of the applied perturbation provides additional hints on the ELM stability. Third, mitigated ELMs, possibly different from spontaneous ones, and their scaling are of interest 'per se', in case that mitigation is to be finally applied in ITER. In this contribution we report on ELM mitigation results in ASDEX Upgrade, where we further restrict the scope to those methods involving 3D, local non-axisymmetric perturbations. Perturbations applied were cryogenic pellet injection, localized supersonic pulsed injection of D gas and laser blow off from targets carrying C or Al micro pellets. The corresponding perturbations differ strongly both in the space and time domain and provide an instrument to probe the local stability. Varying the parameters of the external perturbations by e.g. changing the amount of injected material or the injection speed we mapped out the ELM trigger threshold with high spatial and temporal resolutions.

The 'improved H-mode' regime, realized in ASDEX Upgrade in 1998 and confirmed by other devices, demonstrates the integration of advanced performance beyond the standard H-mode for confinement (confinement enhancement factor H98(y,2) ? 1.6), stability (normalized beta ?N ~ 3?3.5) and, at densities close to the Greenwald density, exhaust in stationary discharges longer than 40 confinement times or up to six resistive diffusion times. The q-profile is characterized by low central magnetic shear and axis safety factor q0 > 1 that is obtained by particular discharge and heating ramp-up scenarios and maintained via fishbones or benign higher (m, n) instabilities without using elaborate current control. Core transport is still governed by drift-wave turbulence with stiff temperature profiles, but density profiles are more strongly peaked and contribute to the increase in global confinement. A further contribution manifests itself by enhanced pressures at the edge barrier pedestal top and at the ? = 0.9 surface both increasing with the input power. (3, 2) NTMs remain small, enabling routine operation up to ?N ~ 3 (limited by (2, 1) NTMs) at ITER relevant collisionalities, for normalized Larmor radii down to four times the ITER value and for a broad range of q95 = 3?5. Tailored heat deposition including central wave heating allows for a compromise in density peaking for enhanced confinement and tolerable high-Z impurity concentrations even with tungsten coated structures.As far as the ITER relevance of this regime is concerned, its compatibility with significant central electron heating, low collisionality and even densities close to the Greenwald density combined with type-II ELMs and ?N ~ 3.5 is of importance. The GLF23 turbulence model still predicts peaked density profiles (R/Ln ~ 3) and sufficient transport to avoid impurity accumulation. At low q95 ~ 3 the fusion performance in terms of is more than doubled compared with the ITER baseline scenario (performance factor ~ 0.2) extrapolating to long Q 10 pulses on ITER. At medium q's bootstrap current fractions up to 50% and performance factors close to 0.2 can be achieved resulting in long inductive pulse lengths of ~1?h allowing ITER 'hybrid' operation at Q ? 9.

The advanced tokamak programme on DIII-D is to develop the scientific basis for steady state, high performance operation in future devices. We report on experiments attempting to demonstrate sustainment of 100% noninductive current for several seconds at high beta, using up to 2.5 MW of off-axis electron cyclotron current drive and up to 15 MW of neutral beam injection with q95 ≈ 5. A 100% noninductive condition was achieved with βT = 3.6%, βN = 3.5, H89 = 2.4 and improved current drive alignment. However, the duration of this phase was limited by the pressure profile evolution leading to magnetohydrodynamic instabilities after about 0.7 s. In a separate discharge, a nearly (~90%) noninductive, stationary condition was maintained for one current relaxation time (1.8 s) only limited by the duration of the hardware system. These experiments have achieved normalized fusion performance with bootstrap current fraction fBS ≈ 60%, consistent with requirements for the ITER Q = 5 steady-state scenarios. The modelling tools that were successfully employed to devise experiments in DIII-D are applied to ITER, indicating that full noninductive operation is plausible for an ITER steady-state scenario with Q ≈ 5.

We present integrated modelling of steady-state and hybrid scenarios for ITER parameters using several predictive transport codes. These employ models for non-inductive current drive sources in conjunction with various theory-based and semi-empirical transport models. In conjunction with the simulation effort, the current drive models are being evaluated in a series of cross-code and code-experiment comparisons under ITER-relevant conditions. New benchmark evaluations of current drive from injection of neutral beams (NBCD), electron cyclotron waves (ECCD) and lower hybrid waves (LHCD) are reported. Simulations using several transport modelling codes self-consistently calculate the heating and current drive sources using ITER design parameters. Operating constraints are also taken into account, although the calculations reported here still require further refinement. The modelling addresses both the final stationary state and dynamic access to it. The simulations indicate that generation and control of internal and edge barriers to access and maintain high confinement will be a major undertaking for future simulations, as well as a challenge for the ITER steady-state and hybrid experimental programme.

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

In this paper several transport phenomena are described and explained through a (quasi) linear description of the micro-instabilities. This paper deals with the following phenomena: density peaking, electron heat transport, density pump-out, reactor density profiles and the stabilization of the ion temperature gradient (ITG) mode in transport barriers. Density peaking is observed to increase with decreasing collisionality?a phenomenon that can be explained by the influence of collisions on the trapped electron response. The density pump-out due to central electron heating is observed to occur only if the dominant instability is the trapped electron mode (TEM). The proposed explanation involves the thermo-diffusive contribution to the particle flux which is outwards for the TEM while it is inwards for the ITG. The insight in the density profile behaviour can be used to predict moderately peaked density profiles in a reactor. Collisionality is also found to influence the electron heat flux in discharges with dominant electron heating even at relatively small values of the density. The dominant instability under dominant electron heating is found to be the TEM, and the dependence of the electron heat flux on the electron temperature gradient is reasonably described by the quasi-linear results. At higher density (and collisionality) the growth rate of the TEM is reduced and a transition to a dominant ITG is found. This transition is reflected in the speed of the electron heat pulse propagation. Finally, it has been found that the ITG under experimentally relevant conditions is not stabilized by a uniform radial electric field.

A stationary regime with improved confinement (H98(y, 2) > 1) and, simultaneously, improved stability (βN > 2.5) compared to standard H-mode has been investigated on ASDEX Upgrade for many years. This so-called 'improved H-mode' is characterized by a q-profile with low central magnetic shear and q0 ≥ 1 that is obtained by early heating during the current ramp. Studies of this discharge scenario have been continued. New results are presented concerning the existence domain in the q95 range, the dependence on the normalized Larmor radius and initial experiments showing that high performance improved H-modes can be obtained with strong central ion cyclotron resonance heating (ICRH). In addition, the present status of understanding of the improved H-mode is reviewed. Improved H-mode plasmas are documented for 3.2 1 sawteeth as the main trigger of large amplitude neoclassical tearing modes are avoided. The stability is eventually limited by the occurrence of a mode with poloidal mode number m = 2 and toroidal mode number n = 1 at typically βN ~ 3. As far as the reactor relevance of this regime is concerned, its compatibility with significant central electron heating by ICRH, with high edge densities and low amplitude edge localized modes is of importance. The improved H-mode is, therefore, seen as a candidate for a long pulse ITER 'hybrid' operation.

An integrated radiative high performance scenario has been established at ASDEX Upgrade based on simultaneous feedback control of the average divertor neutral particle and power flux in combination with a high, pellet induced frequency of edge localized modes (ELMs). This approach is fully compatible with the present tungsten wall coating covering about 65% of the plasma facing components and is intended for application in the envisaged full-tungsten experiment. In these experiments, divertor recycling and effective divertor temperature (derived from thermoelectric currents) were tuned by acting on fuel gas puff and argon injection rates. The ELM frequency (f(ELM)) was kept high by repetitive injection of small cryogenic deuterium pellets to avoid the radiative instabilities seen at low f(ELM) and high radiated power, and to control the ELM energy. No confinement loss is observed in this radiative type-I ELMy scenario with relatively flat density profiles. In contrast, similar type-III ELM scenarios achieved in hydrogen show a confinement loss of 25% as compared to the type-I phase. In parallel to pellets, alternative ELM trigger techniques have been investigated as well. Fast vertical plasma oscillations are able to synchronize the ELM frequency to values higher and lower than the intrinsic f(ELM), but remain to be tested in the integrated scenario. Supersonic gas injection showed better fuelling efficiencies than usual gas puffing but instantaneous ELM release has not been achieved. A particular experimental challenge for AUG conditions is to obtain a high pace making frequency, to establish scalings of confinement and energy loss as a function of controlled ELM frequency.

The Advanced Limiter Test (ALT) project is the focus of a fruitful and intense International Energy Agreement collaboration on TEXTOR. The pump limiter is a mechanical boundary that is laid out for taking the full heat load of TEXTOR, namely 8 MW (assuming 2 MW radiated power) for 10 s, and provides a pumping efficiency of at least 5% of the working gas. This layout is adopted from the requirements of a fusion reactor: It is mandatory to remove both the full power that is convected to the limiter or divertor and the helium ash that is generated in the fusion process. In order to obtain pumping for all gases, the ALT-II is equipped with turbomolecular pumps. A short description of ALT-II is given, and the power and particle fluxes to the limiter surface and into the exhaust scoops are discussed. Requirements of the helium removal rate for a reactor and relevant measurements are discussed, and particle removal and the power distribution to the limiters are treated. Related topics of the ALT-II program were hydrogen recycling and the measurement of turbulence-induced anomalous particle transport in the plasma edge.

Advanced tokamak (AT) research in DIII-D seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and high poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization by plasma rotation and active feedback with non-axisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining inductively driven current, mostly located near the half radius, with non-inductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining inductive current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with ELMing H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. An advanced plasma control system allows integrated control of these elements. Close coupling between modelling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. This approach has resulted in fully non-inductively driven plasmas with ?N ? 3.5 and ?T ? 3.6% sustained for up to 1?s, which is approximately equal to one current relaxation time. Progress in this area, and its implications for next-step devices, will be illustrated by results of these and other recent experiment and simulation efforts.

Advanced Tokamak (AT) research in DIII-D [K. H. Burrell for the DIII-D Team, in Proceedings of the 19th Fusion Energy Conference, Lyon, France, 2002 (International Atomic Energy Agency, Vienna, 2002) published on CD-ROM] seeks to provide a scientific basis for steady-state high performance operation in future devices. These regimes require high toroidal beta to maximize fusion output and poloidal beta to maximize the self-driven bootstrap current. Achieving these conditions requires integrated, simultaneous control of the current and pressure profiles, and active magnetohydrodynamic stability control. The building blocks for AT operation are in hand. Resistive wall mode stabilization via plasma rotation and active feedback with nonaxisymmetric coils allows routine operation above the no-wall beta limit. Neoclassical tearing modes are stabilized by active feedback control of localized electron cyclotron current drive (ECCD). Plasma shaping and profile control provide further improvements. Under these condi...

In this paper, we present numerical tools convenient for the investigation of tokamak discharges with current holes. These tools allow the computation of fixed- and free-boundary equilibria with current holes, their representation in Boozer co-ordinates, the determination of the corresponding magnetic fields, field line tracing and also tracing of guiding centre orbits inside the hole. The numerical methods are tested by using two example configurations with different hole sizes. Furthermore, an equilibrium reconstruction of an ASDEX Upgrade discharge is performed. The motional Stark effect polarization angles of this discharge, which indicate a small current hole in the plasma centre, are reproduced by a numerical fit of various equilibrium parameters including pressure profile, rotational transform profile and poloidal coil currents. The resulting plasma parameters lie well inside the experimental error margins.

External current drive (CD) is an important prerequisite for the control of the plasma current profile in advanced tokamak scenarios as well as for the development of quasi-stationary, fully noninductivly driven tokamak discharges. On ASDEX Upgrade, three heating systems, neutral beam injection, ion cyclotron resonance heating, and electron cyclotron resonance heating, are available for this purpose. The status of CD modeling and the CD capability of these systems are reviewed, and a brief overview is provided of what has been achieved experimentally with respect to CD in various discharge scenarios.

Advanced scenarios in tokamaks seek to maximize the confinement and stability of thermonuclear plasmas. Key to obtaining these conditions is operation at different current density profiles. Experiments at ASDEX Upgrade are reported with approximately zero magnetic shear in the center or reversed magnetic shear in the center. With zero magnetic shear and q 0 near 1, stationary conditions are obtained in discharges without sawteeth at 800 kA and 1 MA and q95 = 3.3 to 4.5, using a combination of central neutral beam injection (NBI) heating and off-axis NBI heating. In this regime, the temperature profiles are stiff. Central heating with ion cyclotron resonance heating and electron cyclotron resonance heating can be used to prevent excessive density peaking to maximize the stability against neoclassical tearing modes and to prevent impurity accumulation. At a lower plasma current of 400 kA with 10 MW of NBI heating, the bootstrap current fraction in this regime is above 50% giving, with the NBI current drive, nearly fully non inductively driven conditions. Operation at average electron densities of 80 to 90% of the Greenwald density limit is obtained at a triangularity of δ = 0.43 achieving β N = 3.5 in stationary conditions. Moreover, in these plasmas, type II edge-localized modes are observed in configurations close to double null. In plasmas with a reversed magnetic shear in the center, the formation of ion transport barriers with NBI heating was optimized to obtain more reproducible transport barriers with an H-mode edge for maximum stability, achieving, transiently, β N values of 4. With a 1.6 MW counter electron cyclotron current drive in the center and densities in the range =1.3 to 2.0 ×10 19 m -3 , a reversed magnetic shear and electron internal transport barriers are formed and sustained at 600 kA for 1 to 2 s with T e0 > 20 keV. Of the scenarios presented, the stationary plasmas with low magnetic shear in the center and q 95 in the range 3.3 to 4.5 would obtain reactor-relevant values for H X β N /q 2 95, a figure of merit used as a benchmark.

This paper provides an overview on recent experimental results obtained in ASDEX Upgrade using electron cyclotron heating and current drive. The following topics are covered: determination of the power deposition profile, modulated power deposition, studies of the electron heat transport via power balance and heat wave analysis and a comparison with turbulent transport theory, generation of an internal transport barrier for the electron heat flux, impact of electron cyclotron resonance heating (ECRH) on particle and impurity transport, and studies related to neoclassical tearing modes and to sawteeth.

ASDEX Upgrade has a large variety of technical systems, heating and fuelling as well as digital control systems needed for control and optimisation of high performance plasmas. The neutral beam injection (NBI), electron cyclotron RF (ECRF) and ion cyclotron resonance frequency (ICRF) heating systems are able to operate in heating as well as in current drive (CD) mode. One of two NBI systems has recently been redirected to allow more tangential injection to provide larger CD efficiency. With improved control procedures significant success is achieved in the optimisation of the conventional H-mode (ITER reference scenario) as well as of ‘advanced tokamak’ scenarios. The established operational boundaries encounter or exceed those of the reference scenario. H-modes showing type II ELMs with clearly reduced peak power loads on the target plates are produced at densities close to the empirical Greenwald density limit. Discharges with nearly full non-inductively driven plasma current at Greenwald densities are performed. MHD instabilities like sawteeth and neoclassical tearing modes (NTM) can be tailored and/or completely be stabilised via electron cyclotron frequency CD at relatively low injection power levels. In addition, several non-standard feedback control circuits for plasma performance improvements and machine protection circuits are devised.

ASDEX upgrade investigates the integration of confinement, stability and exhaust issues into an operating scenario for ITER and a future fusion reactor. Since commissioned in 1990 the systems used to feedback control plasma position and shape as well as performance have continuously been enhanced. To overcome performance limitations and improve connectivity and steady state capability, a new plasma control system is being implemented. For the new system, adequate and reliable communication mechanisms are essential to integrate the realtime discharge control and data acquisition. We present communication methods and the process organisation of the new system and show that the new concept allows easy performance scaling. We demonstrate how existing periphery and new realtime diagnostics interface to control applications. This facilitates the realisation of novel and sophisticated control tasks combining multiple diagnostics and actuators for common physical goals.

The radial electric field gradient or the E×B flow shear has been pointed out as the underlying mechanism for turbulence suppression, responsible for an internal transport barrier formation in advanced tokamak scenarios. A comprehensive study on these subjects requires a direct measurement of the plasma radial electric field Er. The poloidal component of the magnetic field is assessed by the motional Stark effect (MSE) polarimeter, which is currently a standard diagnostic in fusion devices, allowing a local and nonperturbative measurement of the magnetic pitch angle. A precise measure to the state of polarization of the Stark components gives the information on the direction of the magnetic field. Due to the particular orientation of the Lorentz component, that is nearly perpendicular to Er, the MSE diagnostic is very sensitive to the plasma intrinsic radial electric field. This article describes a technique to measure Er involving the change of the polarization angle of the MSE emission, by using two bea...

An advanced high confinement mode scenario with normalized beta (βN) up to 3.5 in stationary conditions (up to 20 energy confinement times) and βN = 3.8 transiently, has been obtained in ASDEX Upgrade. For transport analysis of high βN discharges, two types of simulations have been performed with the automated system for transport analysis code. One is the simulation of the current density profile with experimental data assuming neoclassical electrical conductivity. The other is the simulation of the energy transport and the current density profile using the Weiland transport model (ion temperature gradient and trapped electron mode limit the temperature gradient length). The comparison of the simulations of the q-profiles and temperature profiles with the experimental profiles shows good agreement and demonstrates that for this plasmas regime the temperature profiles are stiff (no internal transport barrier) as in conventional H-mode discharges. The contribution of the Ohmic, bootstrap and neutral beam driven current to the total current density profiles is calculated. The sum of the non-inductively driven plasma current is in the range 50–57% of the total plasma current. Simulations with the Weiland model are performed to investigate the influence of neoclassical tearing modes, which limit confinement and achievable βN in this scenario, by comparing with the measured kinetic data. In addition, the high βN discharges are compared to the improved H-mode discharges in ASDEX Upgrade with respect to the profile stiffness, MHD activity and experimental conditions.

International audience; Ion cyclotron resonance heating (ICRH) in the hydrogen minority scheme provides central ion heating and acts favorably on the core tungsten transport. Full wave modeling shows that, at medium power level (4 MW), after collisional redistribution, the ratio of power transferred to the ions and the electrons vary little with the minority (hydrogen) concentration n H/n e but the high-Z impurity screening provided by the fast ions temperature increases with the concentration. The power radiated by tungsten in the core of the JET discharges has been analyzed on a large database covering the 2013-2014 campaign. In the baseline scenario with moderate plasma current (I p = 2.5 MA) ICRH modifies efficiently tungsten transport to avoid its accumulation in the plasma centre and, when the ICRH power is increased, the tungsten radiation peaking evolves as predicted by the neo-classical theory. At higher current (3-4 MA), tungsten accumulation can be only avoided with 5 MW of ICRH power with high gas injection rate. For discharges in the hybrid scenario, the strong initial peaking of the density leads to strong tungsten accumulation. When this initial density peaking is slightly reduced, with an ICRH power in excess of 4 MW,very low tungsten concentration in the core (˜10-5) is maintained for 3 s. MHD activity plays a key role in tungsten transport and modulation of the tungsten radiation during a sawtooth cycle is correlated to the fishbone activity triggered by the fast ion pressure gradient.

The theoretical model of transport reduction by E × B shear decorrelation is tested experimentally for ASDEX Upgrade discharges. The radial force balance is used to determine the radial electric field from charge exchange recombination spectroscopy measurements. As the effective rate coefficient for photon emission of the charge exchange process depends on the collision energy, the alignment of the lines of sight with respect to the neutral beam gives rise to apparent velocities and temperatures. In addition, the gyro-motion of the observed species along with the finite lifetime of the observed excited state leads to lineshifts in spectra measured in the poloidal direction. Both effects require corrections, which will be discussed. The corrections are tested using the measurements of a discharge with a locked mode. From the profiles of an H mode discharge with improved confinement and a discharge with an internal transport barrier (ITB) the ion heat transport coefficients, E × B shearing rates and the maximum linear growth rates of the instabilities are calculated. Comparison of these results supports the assumption that turbulent transport due to the ion temperature gradient instability is suppressed inside the transport barrier in the ITB discharge. However, due to the dominant influence of the toroidal rotation velocity on the central E × B shear, they do not prove the shear decorrelation model, because Er naturally rises during the phases with improved confinement if unbalanced neutral beam heating is applied.

Various approaches to high performance and steady state operation are presented. Strong neutral beam heating in the current ramp leads to internal transport barriers (ITBs) in conjunction with negative central magnetic shear. To avoid the destabilization of external kink modes this regime is limited to L mode edge operation with βN 1 leads to the formation of an electron ITB.

On the one hand, MHD instabilities in advanced tokamak scenarios are favourable as they can contribute to the stationarity of the current profiles and act as a trigger for the formation of internal transport barriers (ITBs). In particular, fishbone oscillations driven by fast particles arising from NBI are shown to trigger ITBs in low and reversed magnetic shear discharges. During the whistling down period of the fishbone oscillation the transport is reduced around the corresponding rational surface, leading to an increased pressure gradient. This behaviour could be explained by the redistribution of the resonant fast particles resulting in a sheared plasma rotation due to the return current in the bulk plasma, which is equivalent to a radial electric field. On the other hand, MHD instabilities limit the accessible operating regime. Ideal and resistive MHD modes such as double tearing modes, infernal modes and external kinks degrade the confinement or even lead to disruptions in ASDEX Upgrade reversed shear discharges. Localized electron cyclotron heating and current drive are shown to significantly affect the MHD stability of this type of discharge.

A reduction in turbulent density fluctuations is observed in internal transport barrier (ITB) discharges in the ASDEX Upgrade tokamak with an L-mode edge and reversed magnetic shear. The turbulence reduction is localized spatially to around the ITB gradient and the ITB foot-point region, coincident with strong shearing in the toroidal rotation velocity. The radial extent and degree of reduction is related to the strength of the ITB. Strong (narrow and steep) ITBs are well localized and display greater turbulence reduction. Weaker ITBs are less localized with less reduction. The formation of a subsequent H-mode edge barrier moves the velocity shearing region to the edge and weakens the ITB to the point of extinction. During the ITB the turbulence is reduced across a broad frequency range, however a residual level of low-frequency fluctuations (f

Modulated Electron Cyclotron Resonance Heating was applied in ASDEX Upgrade to study the electron heat diffusivity χ e . Its dependence on the isotope mass was found to be opposite to that one known from the power balance analysis. Off-axis ECRH deposition leads to a reduced electron heat transport in the centre which cannot be explained by an inward heat pinch. Electron Cyclotron Current Drive was applied to completely stabilise a (3/2)-neoclassical tearing mode with a consequent recovery of the plasma energy. High electron and ion temperatures have been simultaneously achieved by application of central counter ECCD in a discharge with an internal ion transport barrier.

Normally pellet injection in strongly heated discharges leads at most to a relatively short improvement of the energy and particle confinement times. In contrast to this finding, the radiative improved (RI) mode plasma of TEXTOR-94 is a very well suited target for pellet injection: the interaction of a pellet with the high density plasma together with the radiatively cooled edge leads in the best cases to an improved energy confinement time which lasts to the end of the heating phase. The pellet injection causes at first a density increase; this phase is followed by a quick partial particle loss and later an increase to a quasi-stationary value. Because of the linear density dependence of RI mode confinement this leads to an additional increase in energy confinement time.

The radiative improved (RI) mode is a tokamak regime offering many attractive reactor features. In the article, the RI mode of TEXTOR-94 is shown to follow the same scaling as the linear ohmic confinement regime and is thus identified as one of the most fundamental tokamak operational regimes. The current understanding derived from experiments and modelling of the conditions necessary for sustaining the mode is reviewed, as are the mechanisms leading to L-RI mode transition. The article discusses the compatibility of high impurity seeding with the low central power density of a burning reactor, as well as RI mode properties at and beyond the Greenwald density.

The injection of hydrogen/deuterium pellets into a tokamak leads to a sudden increase in the electron density, and subsequently to a profile peaking of the density and an increase in the stored energy. Immediately after the injection, different types of oscillations are excited. On TEXTOR, the first type immediately follows the injection and the second one is excited with a delay of more than 10 ms. The oscillations show a `snake-like' structure and occur close to the q = 1 surface with a frequency of 0.7 - 2 kHz. The radial location of the second oscillation is slightly shifted with respect to the first one. A fast-cooling phenomenon (`pre-cooling') in the core region of a plasma is often observed at pellet ablation phase. A study on the relation between the `pre-cooling' and sawtooth oscillations suggests that the central value of safety factor of plasmas, q(0), is kept sufficiently below unity even just after the sawtooth crash. During pellet injection, the ablation rate is strongly modulated; these modulations cause so-called `striations' in the ablation cloud. One model relates the striations to the energy reservoir on the plasma flux surfaces and describes the possibility of deriving the q-profile; the question of whether this method provides reliable results cannot yet be answered conclusively. The trajectory of the pellet in the plasma is in general not straight but deflected in the electron drift direction (OH discharges) or in the ion drift direction (CO-NBI discharges). The cloud develops a helically structured tail in the electron flow direction (toroidally) and in the electron diamagnetic drift direction (poloidally). The tail structure is attributed to charge-exchange processes and to plasma rotation.

An important effect observed in ILW discharges is the change of q = 1 MHD activity in hybrid discharges. The q = 1 activity with the JET C-wall had usually shown no effect on plasma confinement, except for NTM triggering, its main experimental signature being a rotating kink activity interleaving sawteeth crashes. In the C-Wall hybrid scenario sawteeth had low amplitude and frequency, and the rotating precursor had often fishbone characteristic. The radial profile of the mode fluctuation has been calculated using a coherence technique between one pick up coil and 96 ECE measurements along the red L.o.S in Figure 1. As explained in ref.[1], the coherence phase radial profile shows a -jump whenever an island is crossed moving along the plasma radius, this effect is due to the radial displacement of magnetic field lines produced by the rotating island, which has opposite directions on the outer side of the island x-point than on the inner side of the x-point. In the C-Wall hybrid pulses the ECE fluctuation at the q = 1 magnetic frequency did not show any radial phase jump, as can be seen in Figure 2, and therefore had kink characteristic.

E. Joffrin, M. Baruzzo, M. Beurskens, Bourdelle, S. Brezinsek, J. Bucalossi, P. Buratti, G. Calabro, C.D. Challis, M. Clever, J. Coenen, E. Delabie, R. Dux, P. Lomas, E. de la Luna, P. de Vries, J. Flanagan, L. Frassinetti, D. Frigione, C. Giroud, M. Groth, N. Hawkes, J. Hobirk, M. Lehnen, G. Maddison, J. Mailloux, C.F. Maggi, G. Matthews, M. Mayoral, A. Meigs, R. Neu, I. Nunes, T. Puetterich, F. Rimini, M. Sertoli, B. Sieglin, A.C.C. Sips, G. van Rooij, I. Voitsekhovitch and JET EFDA contributors In the recent JET experimental campaigns with the new ITER-like Wall (JET-ILW), major progress has been achieved in the characterisation and operation of the H-mode regime in metallic environment: i) plasma breakdown has been achieved at the first attempt and X-point L-mode operation recovered in a few days of operation, ii) stationary and stable type I ELMy Hmodes with bN~1.4 have been achieved in low and high triangularity ITER­like shape plasmas and are showing that their operational domain at H = 1 is significantly reduced with the JET-ILW mainly because of the need to inject large amount of gas (above 1022D/s) to control core radiation. iii) in contrast the hybrid H-mode scenario has reached H factor of 1.2 to 1.3 at bN of 3 for 2 to 3s. iv) in comparison to carbon equivalent discharges, total radiation is similar but the edge radiation is lower and Zeff of the order of 1.3-1.4. At low gas fuelling rate (below 0.5 1022D/s) and low ELM frequency (typically less than 10Hz), strong core radiation peaking is observed in H-mode discharges even when tungsten influx from the divertor is constant. High Z impurity transport from the plasma edge to the core appears to be the dominant factor to explain is paper reviews the major physics and operational achievements and challenges that an ITERlike wall configuration has to face to produce stable plasma scenarios with maximised performance.

The first priority for the 2013 JET campaigns will be to provide relevant data for the ITER decision on the day one divertor material selection. In particular, a tungsten melt experiment using a specially modified divertor module will study the characteristics and impact of ELM induced melting on plasma operation ­ only JET has large enough ELMs to perform such a test. The second priority is to better understand why the change to an all metal wall has restricted access to scenarios with high normalised energy confinement (H-factor). This unexpected result may be partly due to the impact of the higher fuelling needed to control tungsten levels but in some scenarios there is also an effect related to the loss of a beneficial effect of carbon as a plasma impurity. In parallel with pushing JET's core plasma parameters as close as possible to those in ITER, we also intend to develop ITER demonstration pulses including as many of ITER's requirements as possible but not focusing on core performance. Another high priority item is the study of the generation and mitigation of runaway electrons which are so far not created in JET disruptions but are potentially very challenging for JET and ITER. Finally, there is still important work to be continued on key plasma-wall interaction issues ranging from material migration to fuel removal studies.

High confinement improvement (H-factor) relative to the IPB98(y,2) scaling law, stability and bN are achieved in JET hybrid scenario pulses due to formation of specific safety factor (q) profiles with broad flat region around q~1 in the plasma core and steep dq/dr at the edge. The replacement of the C-wall with Be/W-wall caused changes in the impurity content with effect on the q-profile, MHD, radiation and tE. An interaction between these factors in hybrid pulses with ILW has been analysed using transport modelling and important statistical tendencies identified.

The reestablishment of the hybrid scenario in JET after installation of the ITER-like wall (ILW) was successful in reaching similar global performances as in CFC wall (H98 ~1.3, bN ~ 3). However, these performances have only been maintained for 2­3s, after which W accumulation in the core compromises the plasma performance, both due to radiative cooling and to the onset of MHD activity with negative effects on confinement. This paper aims at studying the dynamics of core W accumulation in hybrid plasmas, understanding its origin and possible ways to counteract it, to allow time extension of the improved confinement phase. The focus of the paper will not be on W sources from the edge, which can be regulated by e.g. gas puffing, ELMs pacing, SOL and divertor density control, which are the object of other studies. The focus here will be on core impurity transport and MHD, and how they evolve during the hybrid discharge eventually leading to accumulation.

In this paper numerical simulations with COREDIV code of JET discharges with ITER-like wall are presented. We concentrate on the JET H-mode and hybrid scenarios with nitrogen seeding and all simulations have been performed with the same transport model and only the discharge input parameters like auxiliary heating Paux, line average plasma density nelin, confinement factor H98, nitrogen input flux GNpuff were changed in the calculations. The separatrix density nesep is an input parameter in our model and has been kept equal to 0.4 0.5elin in the simulations, with the recycling coefficient adjusted accordingly. It has been shown that COREDIV is able to reproduce basic parameters of nitrogen seeded discharges for both H-mode and hybrid scenarios. We have achieved reasonable agreement with global plasma parameters like radiations levels, Zeff and tungsten concentrations and the plasma profiles, including density, temperature and radiation are in very good agreement with experimental data. The agreement of the code results with the experimental data might be even better, if the simulations are further tuned taking into account uncertainties to the sputtering model, separatrix density or SOL transport. Simulations show that the observed Zeff level is defined mostly by the low Z impurity content, Be and N2 in the considered shots. It has been found that the tungsten radiation plays always very important role and can not be mitigated even by strong influx of nitrogen.

The behaviour of tungsten in the core of hybrid scenario plasmas in JET with the ITER-like wall is analyzed and modelled with a combination of neoclassical and gyrokinetic codes. In these discharges, good confinement conditions can be maintained only for the first 2 - 3 seconds of the high power phase. Later W accumulation is regularly observed, often accompanied by the onset of magnetohydrodynamical activity, in particular neoclassical tearing modes (NTMs), both of which have detrimental effects on the global energy confinement. The dynamics of the accumulation process is examined, taking into consideration the concurrent evolution of the background plasma profiles, and the possible onset of NTMs. Two time slices of a representative discharge, before and during the accumulation process, are analysed with two indipendent methods, in order to reconstruct the W density distribution over the poloidal cross-section. The same time slices are modelled, computing both neoclassical and turbulent transport components and consistently including the impact of centrifugal effects, which can be significant in these plasmas, and strongly enhanceW neoclassical transport. The modelling closely reproduces the observations and identifies inward neoclassical convection due to the density peaking of the bulk plasma in the central region as the main cause of the accumulation. The analysis of a large set of discharges provides clear indications that this effect is generic in this scenario. The unfavorable impact of the onset of NTMs on the W behaviour, observed in several discharges, is suggested to be a consequence of a detrimental combination of the effects of neoclassical transport and of the appearance of an island.

The construction of the future tokamaks as ITER or DEMO highly depends on the prediction capability of the performance of the main operation scenarios. For this purpose, the validation of the main models available for the plasma simulation is mandatory. QuaLiKiz and TGLF are two of the newest and more sophisticated quasi-linear transport models derived from first principles. QuaLiKiz is based on an electrostatic gyrokinetic eigenvalue code, in the s-a geometry. In the version here used QuaLiKiz does not take plasma rotation effects into account. TGLF is a gyro-Landau fluid model. It contains two models for the E¥B rotation shear: the quench rule and the spectral shift model. It can run using s-a or Miller geometry. Both models take into account passing and trapped particles. These models contain many physics effects, and are fast enough to be inserted in integrated modeling codes (QuaLiKiz is however slower than TGLF and it is parallelized). They then can play a fundamental role in understanding and predicting the transport of plasma particle and heat in present and future machines. QuaLiKiz and TGLF have been first compared in their stand-alone versions. They show good agreement. Then they are used coupled in the CRONOS suite of codes to study heat transport in H-mode and hybrid plasmas of JET. The aim is to validate the models with the experimental data, to compare the results with the well known and faster even more approximated transport model GLF23 and to investigate the possible physical reasons of the resulting discrepancies.

The ExB shear stabilisation of anomalous transport in JET hybrid discharges is studied via selfconsistent predictive modelling of electron and ion temperature, ion density and toroidal rotation velocity performed with the GLF23 model. The ExB shear stabilisation factor (parameter aE in the GLF23 model) is adjusted to predict accurately the four simulated quantities under different experimental conditions, and the uncertainty in aE determined by 15% deviation between simulated and measured quantities is estimated. A correlation of aE with toroidal rotation velocity and ExB shearing rate is found in the low density plasmas, suggesting that the turbulence quench rule may be more complicated than assumed in the GLF23 model with constant aE. For the selected discharges the best predictive accuracy is obtained by using weak/no ExB shear stabilisation (i.e. aE 0) at low toroidal angular frequency (W < 60krad/s), even in the scenarios with the current overshoot, and aE = 0.9 at high frequency (W > 100krad/s). Interestingly, a weak ExB shear stabilisation of anomalous transport is found in the medium density strongly rotating discharge. An importance of linear be stabilisation in this discharge is estimated and compared to the low density discharge with equally high be. The toroidal rotation velocity is well predicted here by assuming that the momentum diffusion coefficient is a fraction of thermal ion diffusivity. Taking into account the aE and Prandtl number with their uncertainties determined in the modelling of JET hybrid discharges, the performance of ITER hybrid scenario with optimised heat mix (33MW of NBI and 20MW of ECCD) is estimated showing the importance of toroidal rotation for achieving Q > 5.

The physical understanding of the transport of heavy impurities, like W, is important for the achievement of practical fusion energy. Among the plasma parameters which affect impurity transport, rotation has only recently received the deserved consideration. In addition to the off-diagonal contribution related to the presence of a radial gradient of the toroidal rotation (usually dubbed roto­diffusion), even in moderately rotating plasmas, centrifugal effects of heavy impurities become non-negligible. Plasmas obtained in the hybrid scenario in JET with the ITER-like wall (JETILW) can be used to study these effects on W, since the deuterium toroidal rotation reaches central thermal Mach numbers around 0.4 (Mach of W up to 3.8). At the same time, this study allows us to explore the transport mechanisms which are responsible for the central W accumulation which is often observed in this type of discharges.

In both JET and ASDEX Upgrade (AUG) the plasma energy confinement has been affected by the presence of a metal wall by the requirement of increased gas-fuelling to avoid tungsten pollution of the plasma. In JET with a beryllium/tungsten wall the high triangularity baseline H-mode scenario (i.e. similar to the ITER reference scenario) has been strongest affected and the benefit of high shaping to give good normalised confinement of H98~1 at high Greenwald density fraction of fGW ~0.8 has not been recovered to date. In AUG with a full tungsten wall a good normalised confinement H98~1 could be achieved in the high triangularity baseline plasmas, albeit at elevated normalised pressure bN>2. The confinement lost with respect to the carbon devices can be largely recovered by the seeding of nitrogen in both JET and AUG. This suggests that the absence of carbon in JET and AUG with a metal wall may have affected the achievable confinement. Threemechanisms have been tested that could explain the effect of carbon or nitrogen (and the absence thereof) on the plasma confinement. First it has been seen in experiments and by means of non-linear gyrokinetic simulations (with the GENE code), that nitrogen seeding does not significantly change the core temperature profile peaking and does not affect the critical ion temperature gradient. Secondly, the dilution of the edge ion density by the injection of nitrogen is not sufficient to explain the plasma temperature and pressure rise. For this latter mechanism to explain the confinement improvement with nitrogen seeding, strongly hollow Zeff profiles would be required, which is not supported by experimental observations. The confinement improvement with nitrogen seeding cannot be explained with these two mechanisms. Thirdly, detailed pedestal structure analysis in JET high triangularity baseline plasmas have shown that fuelling of either deuterium or nitrogen widens the pressure pedestal. However, in JET-ILW this only leads to a confinement benefit in the case of nitrogen seeding where as the pedestal widens, the obtained pedestal pressure gradient is conserved. In the case of deuterium fuelling in JET-ILW the pressure gradient is strongly degraded in the fuelling scan leading to no net confinement gain due to the pedestal widening. The pedestal code EPED correctly predicts the pedestal pressure of the unseeded plasmas in JET-IW within +/­5%, however it does not capture the complex variation of pedestal width and gradient with fuelling and impurity seeding. Also it does not predict the observed increase of pedestal pressure by nitrogen seeding in JET-ILW. Ideal peeling ballooning MHD stability analysis shows that the widening of the pedestal leads to a down shift of the marginal stability boundary by only 10-20%. However, the variations in the pressure gradient observed in the JET-ILW fuelling experiment is much larger and spans a factor of more than two. As a result the experimental points move from deeply unstable to deeply stable on the stability diagram in a deuterium fuelling scan. In AUG-W nitrogen seeded plasmas, also a widening of the pedestal has been observed, consistent with the JET observations. The absence of carbon can thus affect the pedestal structure, and mainly the achieved pedestal gradient, which can be recovered by seeding nitrogen. The underlying physics mechanism is still under investigation and requires further understanding of the role of impurities on the pedestal stability and pedestal structure formation.

Detailed experimental studies of ion heat transport have been carried out in JET exploiting the upgrade of active charge exchange spectroscopy and the availability of multi-frequency ion cyclotron resonance heating with 3He minority. The determination of ion temperature gradient (ITG) threshold and ion stiffness offers unique opportunities for validation of the well-established theory of ITG driven modes. Ion stiffness is observed to decrease strongly in the presence of toroidal rotation when the magnetic shear is sufficiently low. This effect is dominant with respect to the well-known ω E×B threshold up-shift and plays a major role in enhancing core confinement in hybrid regimes and ion internal transport barriers. The effects of T e/T i and s/q on ion threshold are found rather weak in the domain explored. Quasi-linear fluid/gyro-fluid and linear/non-linear gyro-kinetic simulations have been carried out. Whilst threshold predictions show good match with experimental observations, some significant discrepancies are found on the stiffness behaviour.