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The JET hybrid scenario has been developed from low plasma current carbon wall discharges to the record-breaking Deuterium-Tritium plasmas obtained in 2021 with the ITER-like Be/W wall. The development started in pure Deuterium with refinement of the plasma current, and toroidal magnetic field choices and succeeded in solving the heat load challenges arising from 37 MW of injected power in the ITER like wall environment, keeping the radiation in the edge and core controlled, avoiding MHD instabilities and reaching high neutron rates. The Deuterium hybrid plasmas have been re-run in Tritium and methods have been found to keep the radiation controlled but not at high fusion performance probably due to time constraints. For the first time this scenario has been run in Deuterium-Tritium (50:50). These plasmas were re-optimised to have a radiation-stable H-mode entry phase, good impurity control through edge Ti gradient screening and optimised performance with fusion power exceeding 10 MW for longer than three alpha particle slow down times, 8.3 MW averaged over 5 s and fusion energy of 45.8 MJ. [ABSTRACT FROM AUTHOR]

A new technique has been developed to produce plasmas with improved confinement relative to the H98,y2 scaling law (ITER Physics Expert Groups on Confinement and Transport and Confinement Modelling and Database ITER Physics Basics Editors and ITER EDA 1999 Nucl. Fusion 39 2175) on the JET tokamak. In the mid-size tokamaks ASDEX upgrade and DIII-D heating during the current formation is used to produce a flat q-profile with a minimum close to 1. On JET this technique leads to q-profiles with similar minimum q but opposite to the other tokamaks not to an improved confinement state. By changing the method utilizing a faster current ramp with temporary higher current than in the flattop (current overshoot) plasmas with improved confinement (H98,y2 = 1.35) and good stability (βN " 3) have been produced and extended to many confinement times only limited by technical constraints. The increase in H98,y2-factor is stronger with more heating power as can be seen in a power scan. The q-profile development during the high power phase in JET is reproduced by current diffusion calculated by TRANSP and CRONOS. Therefore the modifications produced by the current overshoot disappear quickly from the edge but the confinement improvement lasts longer, in some cases up to the end of the heating phase. [ABSTRACT FROM AUTHOR]

Hybrid scenarios in present machines are often characterized by improved confinement compared with the IPB98(y,2) empirical scaling law expectations. This work concentrates on isolating the impact of increased s/q at outer radii (where s is the magnetic shear) on core confinement in low-triangularity JET and ASDEX Upgrade (AUG) experiments. This is carried out by predictive heat and particle transport modelling using the integrated modelling code CRONOS coupled to the GLF23 turbulent transport model. For both machines, discharge pairs were analysed displaying similar pedestal confinement yet significant differences in core confinement. From these comparisons, it is found that s/q shaping at outer radii may be responsible for up to ∼50% of the relative core confinement improvement observed in these specific discharges. This relative improvement is independent of the degree of rotational shear turbulence suppression assumed in the GLF23 model. However, employing the full GLF23 rotational shear model leads to an overprediction of the ion temperatures in all discharges analysed. Additional mechanisms for core confinement improvement are discussed and estimated. Further linear threshold analysis with QuaLiKiz is carried out on both pairs of discharges. This work aims to validate recent predictions of the ITER hybrid scenario also employing CRONOS/GLF23, where a high level of confinement and resultant fusion power sensitivity to the s/q profile was found. [ABSTRACT FROM AUTHOR]

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. [Copyright &y& Elsevier]

Experimental campaigns at the Joint European Torus with an ITER-like Be/W wall with pure deuterium (D), tritium (T), and deuterium–tritium (DT) were a unique opportunity to explore various aspects related to the ITER operation. One of the most important challenges in recent years was the development of the hybrid scenario for D-T, based on reference deuterium and tritium plasmas. This kind of scenario, one of the foreseen for ITER, is characterized by a low current plasma and a high normalized beta βN factor compared to the parallel optimized baseline scenario [Hobirk et al., Plasma Phys. Controlled Fusion 54, 095001 (2012)]. As the experiments have shown, controlling the plasma edge in the different phases of the hybrid scenario becomes more difficult with higher isotope mass, and therefore, are also in risk of impurity accumulation [Hobirk et al., Nucl. Fusion 63, 112001 (2023)]. For this reason, investigation of the impurity behavior, as well as their control, constituted the crucial issue. The present contribution aims to compare mid-Z and high-Z impurities behavior within H-mode hybrid discharges in D and T plasmas, as well as D and DT plasmas. Detailed analysis shows that in the H-mode regime in the hybrid scenario, higher impurity radiation is observed for DT in comparison to D plasmas, as well as for T compared to D plasmas. Additionally, it was noticed that the most significant contribution to the plasma radiated power comes from W and to a lesser extent from Ni (∼10%). Moreover, it was found that an earlier transition from small edge localized modes (ELMs) to ELM-free phase can result in the earlier increase in impurities. [ABSTRACT FROM AUTHOR]

The present paper is dedicated to the study of the discrepancies encountered in electron temperature (Te) measurements carried out with electron cyclotron emission (ECE) and Thomson scattering (TS) diagnostics in the core of the JET tokamak. A large database of discharges has been collected, including high-performance scenarios performed with deuterium only and deuterium–tritium mixtures. Discrepancies have been found between core Te measurements taken with an X-mode ECE interferometer (TECE) and a LIDAR TS system (TLID) for T e > 5 keV. Depending on the plasma scenario, TECE has been found to be systematically higher or lower than TLID. Discrepancies have also been observed between the peaks of the ECE spectrum in the second (X2) and third (X3) harmonic domains, even in high optical thickness conditions. These discrepancies can be interpreted as evidence of the presence of non-Maxwellian features in the electron energy distribution function (EEDF). In order to investigate the relation between the shape of the EEDF and the measured discrepancies, a model for bipolar perturbations of Maxwellian EEDF has been developed. The model allows analytical calculations of ECE absorption and emission coefficients; hence, the comparison of modeled ECE spectra with experimental data. The different experimental results observed for the various JET scenarios have been found to be qualitatively reproducible by adapting the model parameters, suggesting that bipolar distortions of the bulk EEDF could play a role in giving rise to the reported discrepancies between ECE and TS measurements. [ABSTRACT FROM AUTHOR]

In JET deuterium-tritium (D-T) plasmas, the fusion power is produced through thermonuclear reactions and reactions between thermal ions and fast particles generated by neutral beam injection (NBI) heating or accelerated by electromagnetic wave heating in the ion cyclotron range of frequencies (ICRFs). To complement the experiments with 50/50 D/T mixtures maximizing thermonuclear reactivity, a scenario with dominant non-thermal reactivity has been developed and successfully demonstrated during the second JET deuterium-tritium campaign DTE2, as it was predicted to generate the highest fusion power in JET with a Be/W wall. It was performed in a 15/85 D/T mixture with pure D-NBI heating combined with ICRF heating at the fundamental deuterium resonance. In steady plasma conditions, a record 59 MJ of fusion energy has been achieved in a single pulse, of which 50.5 MJ were produced in a 5 s time window (P fus = 10.1 MW) with average Q = 0.33, confirming predictive modelling in preparation of the experiment. The highest fusion power in these experiments, P fus = 12.5 MW with average Q = 0.38, was achieved over a shorter 2 s time window, with the period of sustainment limited by high-Z impurity accumulation. This scenario provides unique data for the validation of physics-based models used to predict D-T fusion power. [ABSTRACT FROM AUTHOR]

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

Capacitive plasma pickup is a well-known and difficult problem for plasma-facing edge diagnostics. This problem must be addressed to ensure an accurate and robust interpretation of the real signal measurements vs noise. The Faraday cup fast ion loss detector array of the Joint European Torus (JET) is particularly prone to this issue and can be used as a testbed to prototype solutions. The issue of separation and distinction between warranted fast ion signal and electromagnetic plasma noise has traditionally been solved with hardware modifications, but a more versatile post-processing approach is of great interest. This work presents post-processing techniques to characterize the signal noise. While hardware changes and advancements may be limited, the combination with post-processing procedures allows for more rapid and robust analysis of measurements. The characterization of plasma pickup noise is examined for alpha losses in a discharge from JET's tritium campaign. In addition to highlighting the post-processing methodology, the spatial sensitivity of the detector array is also examined, which presents significant advantages for the physical interpretation of fast ion losses. [ABSTRACT FROM AUTHOR]

The aim of this paper is to numerically study the influences of the impurities on the high power hybrid discharges in the JET ITER-like wall (ILW) configuration in the DD and deuteriumâ€"tritium (DT) scenarios. Numerical simulations with the COREDIV code of hybrid discharges with 32 MW auxiliary heating, 2.2 MA plasma current and 2.8 T toroidal magnetic field in the ILW corner configuration are presented. In the simulations five impurity species are used: intrinsic: beryllium (Be) and nickel (Ni) from the side walls, helium (He) from DT reaction, tungsten (W) from divertor and extrinsic neon (Ne) or argon (Ar) by gas puff. The extrapolation of the DD discharges to DT plasmas at the original input power of 32 MW and taking into account only the thermal component of the alpha-power, does not show any significant difference regarding the power to the target with respect to the DD case. Simulations show that sputtering due to D and T is negligible. In contrast, the simulations at auxiliary heating 39 MW show that the power to the target is possibly too high to be sustained for about 5 s by strike-point sweeping alone without any control by Ne seeding. The tungsten is produced mainly by Ni, Be and seeded impurities. [ABSTRACT FROM AUTHOR]

Experiments at ASDEX Upgrade have been conducted to study the impact of the alignment of external magnetic perturbations (MPs) with n = 2 toroidal mode symmetry on the power threshold of the transition from L- to H-mode ( P LH ). This is interesting for ITER, since its operation will rely on an MP field configuration that securely suppresses edge localized modes (ELMs) while avoiding an increase in the L–H power threshold. P LH can increase up to ≈ 80 % when the external MP field is aligned to maximize the associated plasma response at the edge. This alignment deviates from the equilibrium field pitch and is similar to the one that is required to suppress ELMs by MPs in H-mode. The relative radial magnetic field perturbation δ B r / B T has to exceed a critical value, roughly 1.7 − 2 × 10 − 4 (at the q = 5 surface), to cause an increase in P LH . This value is above the one that is required to sustain ELM suppression at ASDEX Upgrade. Combining plasma response calculations with a critical value of δ B r / B T captures well the measured dependencies of P LH on the alignment of the applied MP field. The increase in P LH is accompanied by a change in the E × B flow velocity ( v E × B ) profile. The application of MPs in L-mode flattens and reverses the edge v E × B profile when at least δ B r / B T ≈ 1.9 × 10 − 4 is applied. This is the same value above which P LH increases. More heating power is needed to roughly achieve the same shear in the v E × B profile at the transition to H-mode as without MPs but with an v E × B profile that is shifted into the ion diamagnetic direction. [ABSTRACT FROM AUTHOR]

Understanding the physics of transport during current ramps in tokamaks is essential in order to prepare adequate ramp-up and ramp-down scenarios for ITER, where constraints on the transformer flux and the internal inductance will be more stringent than in present-day devices. In this paper, five transport models (Coppi-Tang, Neo-Alcator, Bohm/gyro-Bohm, critical gradient model and H98/2 scaling-based model) are used to reproduce the experimental data during the current ramps of ASDEX Upgrade with its metallic wall. The calculated temperature profiles are compared with the experimental temperature profiles under different ramp-up conditions. Our study reveals important differences between boronized and non-boronized walls which are reproduced with variable success by the models. We also investigated ramp-up phases heated centrally by electron cyclotron heating which creates experimental conditions with peaked electron temperature profiles, which are very different from the usual ohmic heated cases. Our study reveals that the models react very differently to this additional heating. [ABSTRACT FROM AUTHOR]

Achieving high neutron yields in today's fusion research relies on high-power auxiliary heating in order to attain required core temperatures. This is usually achieved by means of high neutral beam (NB) and radio frequency (RF) power. Application of NB power is accompanied by production of fast beam ions and associated beam–target (BT) reactions. In standard JET operational conditions, deuterium (D) NBs are injected into D plasmas. The injected beams comprise D atoms at full, one-half and one-third injected energy. Typically, the full energy of the injected D beams is between 90 and 120 keV, providing 1.4–2.0 MW of heating, which is about half of the injected power. Half-energy D beams carry about one-third of the injected power and the rest of the power is carried by the third energy fraction of D beams. Under these conditions, thermal fusion reactions, i.e. those between plasma ions, and BT reactions are of the same order of magnitude. This study addresses important issues regarding the impact of density, central electron and ion temperatures and their ratio, Ti(0)/Te(0), on fusion performance, measured by the total neutron yield and BT neutron counts. NB/RF synergistic effects are discussed as well. It is demonstrated that thermal fusion gain increases linearly with normalised plasma pressure, βN, and confinement, Btτ. The BT neutrons are, however, more difficult to predict and this task in general requires numerical treatment. In this study, BT neutrons in JET's best-performing baseline and hybrid pulses are analysed and the underlying dependencies discussed. Central fast ion densities are found to decrease with increased density and density peaking. This is attributed to poorer beam penetration at high density. The BT reactions however are unchanged and can even increase if operating at higher core temperatures. An increase in the central ion temperature and Ti(0)/Te(0) ratio leads to higher total and BT reaction rates whilst simultaneously the ratio of the BT to total neutron decreases significantly. NB/RF synergistic effects are found to have a negligible impact on total neutron rate. This can be explained by the reduced beam penetration in high-density conditions leading to lower central fast ion density. [ABSTRACT FROM AUTHOR]

This article investigates the triggering of ELMs on JET by injection of frozen pellets of isotopes of Hydrogen. A method is established to determine the probability that a specific pellet triggers a particular ELM. This method allows clear distinction between pellet-ELM pairs that are very likely to represent triggering events and pairs that are very unlikely to represent such an event. Based on this, the pellet parameters that are most likely to affect the ability of pellets to trigger ELMs have been investigated. It has been found that the injection location is very important, with injection from the vertical high field side showing a much higher triggering efficiency than low field side (LFS) injection. The dependence on parameters such as pellet speed and size and the time since the last ELM is also seen to be much stronger for LFS injection. Finally, the paper illustrates how improvements to the pellet injection system by streamlining the pellet flight lines and slightly increasing the pellet size has resulted in a significantly improved ability to deliver pellets to the plasma and trigger ELMs. [ABSTRACT FROM AUTHOR]

H-mode plasmas in ASDEX Upgrade (AUG) using different hydrogen isotopes are analysed with respect to their core transport properties. The experimental results are discussed and we present gyrokinetic simulations which are able to reproduce the experimental observations. A novel strategy allows us to disentangle core and pedestal physics by mitigating the isotopic dependence of pedestal properties while keeping the heat and particle sources the same. Matched pedestal profiles are obtained between hydrogen (H) and deuterium (D) plasmas when increasing the triangularity in H plasmas with respect to D plasmas. In the core of these plasmas little isotopic dependence is observed when the fast-ion content is low Wfast/Wth < 1/3. Quasi-linear modelling with TGLF reproduces the experimental trends under these conditions. For larger fast-ion fractions an isotope dependence is observed in the core heat transport. This is related to a difference in fast-ion stabilization of turbulent transport. The fast-ion pressure in H and D plasmas is different due to the mass dependence in the fast-ion slowing down time as well as to operational restrictions when heating with H neutral beam injection (H-NBI) or D-NBI. Typically, Wfast,H < 1/2Wfast,D for comparable NBI heating powers in AUG. The gyrokinetic analysis shows that linear growth rates of ITG modes do not show a pure gyro-Bohm mass dependence, but follow the experimentally observed mass dependence when taking collisions, EM-effects and fast ions into account. Non-linear gyrokinetic simulations reproduce the experimental heat fluxes for different isotopes when fast ions are included. This highlights the role of the fast-ion pressure as a key element to explain the observed differences in the core of H and D plasmas. [ABSTRACT FROM AUTHOR]

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 electron cyclotron emission (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.9 To allow direct comparison with the measurements, synthetic radiation temperatures are computed with the code ECRad.10 Good agreement is found if radial transport occurring predominantly at low electron energies is included. We demonstrate that 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. [ABSTRACT FROM AUTHOR]

The Motional Stark Effect (MSE) diagnostic is a well established technique to infer the local internal magnetic field in fusion plasmas. In this paper, the existing forward model which describes the MSE data is extended by the Zeeman effect, fine-structure, and relativistic corrections in the interpretation of the MSE spectra for different experimental conditions at the tokamak ASDEX Upgrade. The contribution of the non-Local Thermodynamic Equilibrium (non-LTE) populations among the magnetic sub-levels and the Zeeman effect on the derived plasma parameters is different. The obtained pitch angle is changed by 3° ... 4° and by 0.5° ... 1° including the non-LTE and the Zeeman effects into the standard statistical MSE model. The total correction is about 4°. Moreover, the variation of the magnetic field strength is significantly changed by 2.2% due to the Zeeman effect only. While the data on the derived pitch angle still could not be tested against the other diagnostics, the results from an equilibrium reconstruction solver confirm the obtained values for magnetic field strength. [ABSTRACT FROM AUTHOR]

Extensive preparations are now underway for an experiment in the Joint European Torus (JET) using tritium and deuterium–tritium mixtures. The goals of this experiment are described as well as the progress that has been made in developing plasma operational scenarios and physics reference pulses for use in deuterium–tritium and full tritium plasmas. At present, the high performance plasmas to be tested with tritium are based on either a conventional ELMy H-mode at high plasma current and magnetic field (operation at up to 4 MA and 4 T is being prepared) or the so-called improved H-mode or hybrid regime of operation in which high normalised plasma pressure at somewhat reduced plasma current results in enhanced energy confinement. Both of these regimes are being re-developed in conjunction with JET's ITER-like Wall (ILW) of beryllium and tungsten. The influence of the ILW on plasma operation and performance has been substantial. Considerable progress has been made on optimising performance with the all-metal wall. Indeed, operation at the (normalised) ITER reference confinement and pressure has been re-established in JET albeit not yet at high current. In parallel with the physics development, extensive technical preparations are being made to operate JET with tritium. The state and scope of these preparations is reviewed, including the work being done on the safety case for DT operation and on upgrading machine infrastructure and diagnostics. A specific example of the latter is the planned calibration at 14 MeV of JET neutron diagnostics. [ABSTRACT FROM AUTHOR]

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 P aux, line average plasma density n lin e, confinement factor H98, nitrogen input flux Γ puff N were changed in the calcu lations. The separatrix density n sep e is an input parameter in our model and has been kept equal to 0.4÷0.5 n lin ein 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, Z eff 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 Z eff 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) [ABSTRACT FROM AUTHOR]

Type I ELMy H-mode operation in JET with the ITER-like Be/W wall (JET-ILW) generally occurs at lower pedestal pressures compared to those with the full carbon wall (JET-C). The pedestal density is similar but the pedestal temperature where type I ELMs occur is reduced and below to the so-called critical type I–type III transition temperature reported in JET-C experiments. Furthermore, the confinement factor H98(y,2) in type I ELMy H-mode baseline plasmas is generally lower in JET-ILW compared to JET-C at low power fractions Ploss/Pthr,08 < 2 (where Ploss is (Pin − dW/dt), and Pthr,08 the L–H power threshold from Martin et al 2008 (J. Phys. Conf. Ser.123 012033)). Higher power fractions have thus far not been achieved in the baseline plasmas. At Ploss/Pthr,08 > 2, the confinement in JET-ILW hybrid plasmas is similar to that in JET-C. A reduction in pedestal pressure is the main reason for the reduced confinement in JET-ILW baseline ELMy H-mode plasmas where typically H98(y,2) = 0.8 is obtained, compared to H98(y,2) = 1.0 in JET-C. In JET-ILW hybrid plasmas a similarly reduced pedestal pressure is compensated by an increased peaking of the core pressure profile resulting in H98(y,2) ⩽ 1.25. The pedestal stability has significantly changed in high triangularity baseline plasmas where the confinement loss is also most apparent. Applying the same stability analysis for JET-C and JET-ILW, the measured pedestal in JET-ILW is stable with respect to the calculated peeling–ballooning stability limit and the ELM collapse time has increased to 2 ms from typically 200 µs in JET-C. This indicates that changes in the pedestal stability may have contributed to the reduced pedestal confinement in JET-ILW plasmas. A comparison of EPED1 pedestal pressure prediction with JET-ILW experimental data in over 500 JET-C and JET-ILW baseline and hybrid plasmas shows a good agreement with 0.8 < (measured pped)/(predicted pped,EPED) < 1.2, but that the role of triangularity is generally weaker in the JET-ILW experimental data than in the model predictions. [ABSTRACT FROM AUTHOR]

Since its inception in 2002, the International Tokamak Physics Activity topical group on Integrated Operational Scenarios (IOS) has coordinated experimental and modelling activity on the development of advanced inductive scenarios for applications in the ITER tokamak. The physics basis and the prospects for applications in ITER have been advanced significantly during that time, especially with respect to experimental results. The principal findings of this research activity are as follows. Inductive scenarios capable of higher normalized pressure (βN ⩾ 2.4) than the ITER baseline scenario (βN = 1.8) with normalized confinement at or above the standard H-mode scaling are well established under stationary conditions on the four largest diverted tokamaks (AUG, DIII-D, JET, JT-60U), demonstrated in a database of more than 500 plasmas from these tokamaks analysed here. The parameter range where high performance is achieved is broad in q95 and density normalized to the empirical density limit. MHD modes can play a key role in reaching stationary high performance, but also define the limits to achieved stability and confinement. Projection of performance in ITER from existing experiments uses empirical scalings and theory-based modelling. The status of the experimental validation of both approaches is summarized here. The database shows significant variation in the energy confinement normalized to standard H-mode confinement scalings, indicating the possible influence of additional physics variables absent from the scalings. Tests using the available information on rotation and the ratio of the electron and ion temperatures indicate neither of these variables in isolation can explain the variation in normalized confinement observed. Trends in the normalized confinement with the two dimensionless parameters that vary most from present-day experiments to ITER, gyroradius and collision frequency, are significant. Regression analysis on the multi-tokamak database has been performed, but it appears that the database is not conditioned sufficiently well to yield a new scaling for this type of plasma. Coordinated experiments on size scaling using the dimensionless parameter scaling approach find a weaker scaling with normalized gyroradius than the standard H-mode scaling. Preliminary studies on scaling with collision frequency show a favourable scaling stronger than the standard H-mode scaling. Coordinated modelling activity has resulted in successful benchmarking of modelling codes in the ITER regime. Validation of transport models using these codes on present-day experiments is in progress, but no single model has been shown to capture the variations seen in the experiments. However, projection to ITER using these models is in general agreement with the favourable projections found with the empirical scalings. [ABSTRACT FROM AUTHOR]

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 the strongest affected and the benefit of high shaping to give good normalized 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 normalized confinement H98 ∼ 1 could be achieved in the high triangularity baseline plasmas, albeit at elevated normalized pressure βN > 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. Three mechanisms 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 nonlinear 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 the 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-ILW 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, a widening of the pedestal has also 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. [ABSTRACT FROM AUTHOR]

A spectrally resolved Motional Stark Effect (MSE) diagnostic has been installed at ASDEX Upgrade. The MSE data have been fitted by a forward model providing access to information about the magnetic field in the plasma interior [R. Reimer, A. Dinklage, J. Geiger et al., Contrib. Plasma Phys. 50, 731-735 (2010)]. The forward model for the beam emission spectra comprises also the fast ion Dα signal [W. W. Heidbrink and G. J. Sadler, Nucl. Fusion 34, 535-615 (1994)] and the smearing on the CCD-chip. The calculated magnetic field data as well as the revealed (dia)magnetic effects are consistent with the results from equilibrium reconstruction solver. Measurements of the direction of the magnetic field are affected by unknown and varying polarization effects in the observation. [ABSTRACT FROM AUTHOR]

The 'hybrid' scenario is an attractive operating scenario for ITER since it combines long plasma duration with the reliability of the reference H-mode regime. We review the recent European modelling effort carried out within the Integrated Scenario Modelling group which aims at (i) understanding the underlying physics of the hybrid regime in ASDEX-Upgrade and JET and (ii) extrapolating them towards ITER. JET and ASDEX-Upgrade hybrid scenarios performed under different experimental conditions have been simulated in an interpretative and predictive way in order to address the current profile dynamics and its link with core confinement, the relative importance of magnetic shear, s, and E x B flow shear on the core turbulence, pedestal stability and H-L transition. The correlation of the improved confinement with an increased s/q at outer radii observed in JET and ASDEX-Upgrade discharges is consistent with the predictions based on the GLF23 model applied in the simulations of the ion and electron kinetic profiles. Projections to ITER hybrid scenarios have been carried out focusing on optimization of the heating/current drive schemes to reach and ultimately control the desired plasma equilibrium using ITER actuators. Firstly, access condition to the hybrid-like q -profiles during the current ramp-up phase has been investigated. Secondly, from the interpreted role of the s/q ratio, ITER hybrid scenario flat-top performance has been optimized through tailoring the q-profile shape and pedestal conditions. EPED predictions of pedestal pressure and width have been used as constraints in the interpretative modelling while the core heat transport is predicted by GLF23. Finally, model- based approach for real-time control of advanced tokamak scenarios has been applied to ITER hybrid regime for simultaneous magnetic and kinetic profile control. [ABSTRACT FROM AUTHOR]

One of the key challenges for plasma theory and simulation in view of ITER is to enhance the understanding and predictive capability concerning high-performance discharges. This involves, in particular, questions about high-β operation, ion temperature profile stiffness, and the physics of transport barriers. The goal of this contribution is to shed light on these issues by means of physically comprehensive ab initio simulations with the global gyrokinetic code GENE, applied to discharges in TCV, ASDEX Upgrade, and JET--with direct relevance to ITER. [ABSTRACT FROM AUTHOR]

To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITER's plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30 MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (≈ factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D2/Ar mixture restores levels of radiation and vessel forces similar to those of mitigated disruptions with the C wall. Dedicated L-H transition experiments indicate a 30% power threshold reduction, a distinct minimum density, and a pronounced shape dependence. The L-mode density limit was found to be up to 30% higher than for C allowing stable detached divertor operation over a larger density range. Stable H-modes as well as the hybrid scenario could be re-established only when using gas puff levels of a few 1021 es-1. On average, the confinement is lower with the new PFCs, but nevertheless, H factors up to 1 (H-Mode) and 1.3 (at βN≈3, hybrids) have been achieved with W concentrations well below the maximum acceptable level. [ABSTRACT FROM AUTHOR]

The confinement in JET baseline type I ELMy H-mode plasmas is compared to that in so-called hybrid H-modes in a database study of 112 plasmas in JET with the carbon fibre composite (CFC) wall. The baseline plasmas typically have βN ~ 1.5-2, H98 ~ 1, whereas the hybrid plasmas have βN ~ 2.5-3, H98 < 1.5. The database study contains both low- (δ ~ 0.2-0.25) and high-triangularity (δ ~ 0.4) hybrid and baseline H-mode plasmas from the last JET operational campaigns in the CFC wall from the period 2008-2009. Based on a detailed confinement study of the global as well as the pedestal and core confinement, there is no evidence that the hybrid and baseline plasmas form separate confinement groups; it emerges that the transition between the two scenarios is of a gradual kind rather than demonstrating a bifurcation in the confinement. The elevated confinement enhancement factor H98 in the hybrid plasmas may possibly be explained by the density dependence in the τ98 scaling as n0.41 and the fact that the hybrid plasmas operate at low plasma density compared to the baseline ELMy H-mode plasmas. A separate regression on the confinement data in this study shows a reduction in the density dependence as n0.09±0.08. Furthermore, inclusion of the plasma toroidal rotation in the confinement regression provides a scaling with the toroidal Alfven Mach number as MachA0.41±0.07 and again a reduced density dependence as n0.15±0.08. The differences in pedestal confinement can be explained on the basis of linear MHD stability through a coupling of the total and pedestal poloidal pressure and the pedestal performance can be improved through plasma shaping as well as high β operation. This has been confirmed in a comparison with the EPED1 predictive pedestal code which shows a good agreement between the predicted and measured pedestal pressure within 20-30% for a wide range of βN ~ 1.5-3.5. The core profiles show a strong degree of pressure profile consistency. No beneficial effect of core density peaking on confinement could be identified for the majority of the plasmas presented here as the density peaking is compensated by a temperature de-peaking resulting in no or only a weak variation in the pressure peaking. The core confinement could only be optimized in case the ions and electrons are decoupled, in which case the ion temperature profile peaking can be enhanced, which benefits confinement. In this study, the latter has only been achieved in the low-triangularity hybrid plasmas, and can be attributed to low-density operation. Plasma rotation has been found to reduce core profile stiffness, and can explain an increase in profile peaking at small radius ρtor = 0.3. [ABSTRACT FROM AUTHOR]

This paper summarizes the operational experience of the ion cyclotron resonant frequency (ICRF) ITER-like antenna on JET aiming at substantially increasing the power density in the range of the requirements for ITER combined with load resiliency. An in-depth description of its commissioning, operational aspects and achieved performances is presented. [ABSTRACT FROM AUTHOR]

Recent experiments at ASDEX Upgrade demonstrate the compatibility ofELMmitigation by magnetic perturbations with efficient particle fuelling by inboard pellet injection. ELM mitigation persists in a high-density, highcollisionality regime even with the strongest applied pellet perturbations. Pellets injected into mitigation phases trigger no type-I ELM-like events unlike when launched into unmitigated type-I ELMy plasmas. Furthermore, the absence of ELMs results in improved fuelling efficiency and persistent density build-up. Pellet injection is helpful to access the ELM-mitigation regime by raising the edge density beyond the required threshold level, mostly eliminating the need for strong gas puff. Finally, strong pellet fuelling can be applied to access high densities beyond the density limit encountered with pure gas puffing. Core densities of up to 1.6 times the Greenwald density have been reached while maintaining ELM mitigation. No upper density limit for the ELM-mitigated regime has been encountered so far; limitations were set solely by technical restrictions of the pellet launcher. Reliable and reproducible operation at line-averaged densities from 0.75 up to 1.5 times the Greenwald density is demonstrated using pellets. However, in this density range there is no indication of the positive confinement dependence on density implied by the ITERH98P(y,2) scaling. [ABSTRACT FROM AUTHOR]

Stability of high-beta plasmas is studied on discharges from a series of JET experiments on steady-state and hybrid advanced scenarios, with a wide range of safety factor (q) profiles and normalized beta values extending to βN = 4. Bursting and continuous forms of global n = 1 instabilities are encountered that degrade confinement or, in some cases, give rise to disruptions. Mode frequencies are well above the inverse wall time and correspond to plasma rotation at around mid-radius. Stability boundaries in terms of qmin and pressure peaking are examined. For relatively broad pressure profiles the stability limit decreases from βN = 4 at qmin = 1 to βN = 2 at qmin = 3, while at fixed qmin it decreases with increasing pressure peaking. Metastable and unstable regions are identified in the βN-qmin diagram by mode-trigger analysis. Tearing and kink mode structures are found from phase analysis of temperature profile oscillations; for a selection of kink cases, instability conditions and mode structure are compared with ideal stability calculations. [ABSTRACT FROM AUTHOR]

New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation. [ABSTRACT FROM AUTHOR]

In this communication we propose a novel diagnostic technique, which uses the collection optics of the JET Motional Stark Effect (MSE) diagnostic, to perform polarimetry marking of observed MHD in high temperature plasma regimes. To introduce the technique, first we will present measurements of the coherence between MSE polarimeter, electron cyclotron emission, and Mirnov coil signals aiming to show the feasibility of the method. The next step consists of measuring the amplitude fluctuation of the raw MSE polarimeter signals, for each MSE channel, following carefully the MHD frequency on Mirnov coil data spectrograms. A variety of experimental examples in JET ITER-Like Wall (ILW) plasmas are presented, providing an adequate picture and interpretation for the MSE optics polarimeter technique. [ABSTRACT FROM AUTHOR]

Neutral beam current drive (NBCD) experiments in DIII-D using vertically shifted plasmas to move the current drive away from the axis have clearly demonstrated robust off-axis NBCD. Time-dependent measurements of magnetic field pitch angles by the motional Stark effect diagnostic are used to obtain the evolution of the poloidal magnetic flux, which indicates a broad off-axis NBCD profile with a peak at about half the plasma minor radius. In most cases, the measured off-axis NBCD profile is consistent with calculations using an orbit-following Monte Carlo code for the beam ion slowing down including finite-orbit effects provided there is no large-scale magnetohydrodynamic activity such as Alfvén eigenmodes modes or sawteeth. An alternative analysis method shows good agreement between the measured pitch angles and those from simulations using transport-equilibrium codes. Two-dimensional image of Doppler-shifted fast ion Dα light emitted by neutralized energetic ions shows clear evidence for a hollow profile of beam ion density, consistent with classical beam ion slowing down. The magnitude of off-axis NBCD is sensitive to the alignment of the beam injection relative to the helical pitch of the magnetic field lines. If the signs of toroidal magnetic field and plasma current yield the proper helicity, both measurement and calculation indicate that the efficiency is as good as on-axis NBCD because the increased fraction of trapped electrons reduces the electron shielding of the injected ion current, in contrast with electron current drive schemes where the trapping of electrons degrades the efficiency. The measured off-axis NBCD increases approximately linearly with the injection power, although a modest amount of fast ion diffusion is needed to explain an observed difference in the NBCD profile between the measurement and the calculation at high injection power. [ABSTRACT FROM AUTHOR]

The “improved H-mode,” realized in ASDEX Upgrade [A. Herrmann and O. Gruber, Fusion Sci. Technol. 44, 569 (2003)] in 1998, demonstrates that advanced requirements beyond the standard H-mode for confinement [confinement enhancement factor H98(y,2)>1], stability (normalized beta βN∼3–3.5) and, at densities close to Greenwald density, exhaust can be simultaneously met and maintained stationary for several resistive diffusion times. The q profile is characterized by low central magnetic shear and axis safety factor q0>1 that is obtained by particular heating and current ramp-up scenarios and maintained via benign instabilities. 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 confinement. Neoclassical tearing modes remain small, enabling routine operation up to βN∼3 at international thermonuclear experimental reactor (ITER) relevant collisionalities, for normalized Lamor radii down to four times the ITER value and for a broad range of q95=3.2–4.5. Using tailored heat deposition including central wave heating a compromise was found in density peaking for enhanced confinement and limiting the high-Z impurity concentrations even with a tungsten-coated first wall and divertor. As far as the ITER [ITER EDA Documentation Series No. 24, 2002] relevance of this regime is concerned, its compatibility with significant central electron heating, high edge densities, and type-II edge localized modes is of importance. The GLF23 turbulence model predicts still peaked density profiles and sufficient transport to avoid impurity accumulation. The fusion performance in terms of βNH98(y,2)/q952 is nearly doubled compared with the ITER base-line scenario at low-q values, while at medium q’s bootstrap current fractions up to 50% and long inductive pulse lengths allow ITER “hybrid” operation. [ABSTRACT FROM AUTHOR]

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 conditions, bootstrap supplies most of the current. Steady-state operation requires replacing the remaining Ohmic current, mostly located near the half radius, with noninductive external sources. In DIII-D this current is provided by ECCD, and nearly stationary AT discharges have been sustained with little remaining Ohmic current. Fast wave current drive is being developed to control the central magnetic shear. Density control, with divertor cryopumps, of AT discharges with edge localized moding H-mode edges facilitates high current drive efficiency at reactor relevant collisionalities. A sophisticated plasma control system allows integrated control of these elements. Close coupling between modeling and experiment is key to understanding the separate elements, their complex nonlinear interactions, and their integration into self-consistent high performance scenarios. [ABSTRACT FROM AUTHOR]

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. [Copyright &y& Elsevier]

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 E[sub r]. 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 E[sub r], the MSE diagnostic is very sensitive to the plasma intrinsic radial electric field. This article describes a technique to measure E[sub r] involving the change of the polarization angle of the MSE emission, by using two beam injectors at different energies, firing sequentially. Experimental results for the low E[sub r] case, i.e., with very little plasma rotation, showing the ability of the MSE to perform this measurement, will be presented. This is the first time that evidence of a direct measurement of the plasma E[sub r] is reported from the Joint European Torus. [ABSTRACT FROM AUTHOR]

The behaviour of tungsten in the core of hybrid scenario plasmas in JET with the ITER-like wall is analysed 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 s of the high power phase. Later W accumulation is regularly observed, often accompanied by the onset of magneto-hydrodynamical 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 independent 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 enhance W 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 change in W neoclassical convection is directly produced by the transient behaviour of the main plasma density profile, which is hollow in the central region in the initial part of the high power phase of the discharge, but which develops a significant density peaking very close to the magnetic axis in the later phase. The analysis of a large set of discharges provides clear indications that this effect is generic in this scenario. The unfavourable 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. [ABSTRACT FROM AUTHOR]

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. [ABSTRACT FROM AUTHOR]

Abstract: Operation with all tungsten plasma facing components has become routine in ASDEX Upgrade. The conditioning of the device is strongly simplified and short glow discharges are used only on a daily basis. The long term fuel retention was reduced by more than a factor of 5 as demonstrated in gas balance as well as in post mortem analyses. Injecting nitrogen for radiative cooling, discharges with additional heating power up to 23MW have been achieved, providing good confinement (H98y2 =1), divertor power loads around 5MWm−2 and divertor temperatures below 10eV. ELM mitigation by pellet ELM pacemaking or magnetic perturbation coils reduces the deposited energy during ELMs, but also keeps the W density at the pedestal low. As a recipe to keep the central W concentration sufficiently low, central (wave) heating is well established and low density H-Modes could be re-established with the newly available ECRH power of up to 4MW. The ICRH induced W sources could be strongly reduced by applying boron coatings to the poloidal guard limiters. [Copyright &y& Elsevier]