Reactor Relevant Regime Research Articles

Design of a multi-configurations divertor for the DTT facility

The standard positive triangularity H-mode tokamak operation with Single Null Divertor (SND) configuration and tungsten monoblocks targets can face some difficulties in providing a solution scalable towards the realization of the fusion reactor. To provide a safer solution, the adoption of alternative divertor magnetic configurations (ADCs) has been considered to provide on the targets a stationary heat load of less than 10 MW/m2 presently considered an upper technological limit for a nuclear fusion reactor and to reduce enough plasma temperature to avoid tungsten sputtering and its possible plasma contamination. Additionally, different plasma scenarios have been considered to avoid the huge transient energy released due to the type-I ELMs of the high confinement H-mode operation. The new high field superconducting divertor tokamak test facility (DTT) [1] is presently under construction to specifically study power exhaust solutions in reactor relevant regimes. The first DTT divertor will use the ITER-like technology based on full tungsten monoblocks bonded on CuCrZr cooling tubes. To test and compare a wide set of different power exhaust solutions, the divertor is being designed to be compatible with the standard single null (SND) configuration but also with some ADCs, like the X divertor (XD) and the hybrid Super-X/long leg SN. Additionally, the l-mode negative triangularity (NT) operation is considered important to explore as a solution to avoid ELMs. Two different closed divertors with the same grazing angle (α = 2°) for the reference SND configuration have been considered: a so called “narrow” one with normalized divertor parameters similar to those presently foreseen for DEMO divertor and the “wide” one with a 50 % wider target aperture which allows testing long legs configurations as well as snowflake divertor configurations (SFD). For the “wide” divertor different dome sizes have been also considered. The comparison between different shapes and their optimization have been done with the 2D edge fluid-kinetic code SOLEDGE2D-EIRENE. The code easily manages all magnetic configurations and can resolve the heat load on all components thanks the mesh filling the whole divertor and vessel volume. In a representative sub-set of all the possible magnetic equilibrium configurations that can be realized in DTT, the edge modelling shows a better performance of the “wide” divertor compared to the “narrow” one. In fact, the “wide” divertor allows to have lower plasma temperature at the targets in pure deuterium at low PSOL power and also lower impurity concentrations is required to achieve plasma detachment by impurity seeding at the reference full power DTT scenario. The results seem to indicate that the improvement due to the longer legs possible with the “wide” divertor can overcome its lower apparent closure compared to the “narrow” divertor solution.

Improved neoclassical plasma confinement and turbulence suppression in a magnetic well in tokamak reactors

It has been known that the width of the trapped particles, i.e. bananas, is reduced or squeezed in a magnetic well in tokamaks. The magnetic squeezing factor S B depends on the energy and pitch angle of the particles. When ρ p/a ∼ 1, S B becomes appreciably larger than unity; it can be of the order of 2 for a parabolic well with B −1 d2 B/dr 2 ∼ 2a −2. Here, ρ p is the poloidal gyro-radius, B is the magnetic field strength, r is the local minor radius, and a is the minor radius. However, the real orbit width measured in terms of the poloidal magnetic flux is still less than a because of the magnetic squeezing; for S B = 2, it is a 30% reduction. The transport consequences of the squeezed bananas in a magnetic well are calculated by solving the drift kinetic equation utilizing the constants of motion in the large aspect ratio limit. Consequently, neoclassical ion heat conductivity χ h in the reactor relevant collisionality regime (i.e. banana regime) is reduced by a factor of ; for S B = 2, it is almost a factor of three reduction for the same temperature. The ion temperature scaling of χ h improves from the conventional, and weakly favorable scaling to the strongly favorable scaling for high temperature fusion grade tokamak plasmas; the scaling however does not take the temperature dependence of the magnetic well itself into account. Here, T i is the ion temperature. In addition, it is shown that plasma turbulence is suppressed in a magnetic well using the reduced decorrelation time argument and anomalous plasma confinement is improved. The combined improvements result in a favorable temperature scaling in the ion energy confinement time, and make burning aneutronic fuels feasible. The implications on a tokamak with a magnetic well as a fusion reactor are discussed.

Towards understanding reactor relevant tokamak pedestals

The physics of the tokamak pedestal is still not fully understood, for example there is no fully predictive model for the pedestal height and width. However, the pedestal is key in determining the fusion power for a given scenario. If we can improve our understanding of reactor relevant pedestals we will improve our confidence in designing potential fusion power plants. Work has been carried out as part of a collaboration on reactor relevant pedestal physics. We report some of the results in detail here and review some of the wider work which will be reported in full elsewhere. First, we attempt to use a gyrokinetic-based calculation to eliminate the pedestal top density as a model input for Europed/EPED pedestal predictions. We assume power balance at the top of the pedestal, that is, the heat flux crossing the separatrix must be equal to the heat source at the top of the pedestal and investigate the consequences of this assumption. Unfortunately, the transport assumptions of the EPED model mean that this method does not discriminate between different pairs of density and temperature profiles for a given pressure profile. Second, we investigate the effects of non flux surface density on the bootstrap current. Third, type I ELMs will not be tolerable for a reactor relevant regime due to the damage that they are expected to cause to plasma facing components. In recent years various methods of running tokamak plasmas without large ELMs have been developed. These include small and no ELM regimes, the use of resonant magnetic perturbations and the use of vertical kicks. We discuss the quiescent H-mode here. Finally we give a summary and directions for future work.

H-mode confinement in the pellet-enforced high-density regime of the all-metal-wall tokamak ASDEX Upgrade

During recent years a pellet-based system for feedback controlled core density fuelling has been developed at the ASDEX Upgrade tokamak, making the device well suited for the investigation of the reactor relevant high-density regime in an appropriate configuration. Efforts have proven that an adequately applied pellet actuator enables safe and reversible access to core plasma densities, as intended in current reactor designs and even beyond. From the analysis of a wide range of different plasma scenarios it becomes apparent that the beneficial increase of confinement with density saturates when approaching the Greenwald density. When taking this into account properly, it has emerged that for standard configurations the energy confinement can be sustained when the density is increased. However, the surplus confinement by different means, such impurity seeding or strong shaping, cannot be fully preserved. This is presumed to be attributed to a separatrix density which is not yet adequately controlled.

Investigating the thermal stability of highly radiative discharges on JET with a new tomographic method

The next generation of Tokamak devices is expected to work at very high radiated fractions, well above 90%, to preserve the integrity of the plasma facing components in general and the divertor in particular. In addition to maintaining high confinement, these configurations will also have to guarantee a low disruptivity. An accurate determination of the emitted radiation will therefore become increasingly important, not only for the global power balances but also for specific regions of the plasma cross section (for example to properly control detachment). In this perspective, a new tomographic inversion method, based on the maximum likelihood (ML) approach, capable of providing routinely confidence intervals in the estimates of the radiated power, has been applied to the investigation of high radiative discharges on JET with the ITER like wall (ILW). The emission has been increased with injection of extrinsic impurities. Taking into account all the major sources of uncertainties, a systematic analysis of the configurations has shown that it has not been possible to develop stable configurations with radiated fraction higher than 70% of the input power. At higher radiated fractions the discharges always disrupt. Therefore significant work remains to be done to extend JET operation in a reactor relevant regime of sufficient radiation in preparation for ITER and DEMO.

A hybrid fluid-kinetic model for hydrogenic atoms in the plasma edge of tokamaks based on a micro-macro decomposition of the kinetic equation

Monte Carlo (MC) simulations of the full kinetic equation for the neutral particles in the plasma edge become computationally costly for reactor-relevant regimes. To accelerate the simulations, we propose a hybrid fluid-kinetic approach that is based on a micro-macro decomposition of the kinetic equation. This leads to a macro/fluid model with kinetic corrections that follow from an MC simulation of the micro/kinetic part. We distinguish three hybrid models with different underlying fluid equations: (i) a pure pressure-diffusion equation with equal neutral and ion temperatures (Tn=Ti); (ii) a continuity and parallel momentum equation with pressure-diffusion transport retained in the directions perpendicular to the magnetic field lines, with Tn=Ti; and (iii) the same model as (ii), but with a separate neutral energy equation (Tn≠Ti). To facilitate the future integration in more complete plasma edge codes, we neglect some kinetic correction terms. Hence, the hybrid model is not exactly equivalent to the full kinetic equation. We assess the hybrid performance on the basis of the reduction of the CPU time compared to an MC simulation of the full kinetic equation for the same statistical error on a certain plasma source. This is done for a high recycling slab case. Only the models with parallel momentum equation ((ii)-(iii)) are able to significantly reduce the CPU time. However, due to the incomplete kinetic corrections there is a remaining hybrid-kinetic discrepancy that mainly pops up in the ion energy source from the model with energy equation (iii).

DIII-D Research to Prepare for Steady State Advanced Tokamak Power Plants

We review progress made on the advanced tokamak path to fusion energy by the DIII-D National Fusion Facility (Luxon et al. in Nucl Fusion 42:614, 2002). The advanced tokamak represents a highly attractive approach for a future steady state fusion power plant. In this concept, there is a natural alignment between high pressure operation, favorable stability and transport properties, and a highly self-driven (‘bootstrap’) plasma current to sustain operation efficiently and without disruptions. Research on DIII-D has identified several promising plasma configurations for fully non-inductive operation with potential applications to a range of future devices, from ITER to nuclear science facilities, to compact or large scale fusion power plants. Significant progress has been made toward realizing these scenarios, with the demonstration of high β access, off-axis current drive techniques, model based profile control, and stability and ELM control in reactor relevant physics regimes. Radiative techniques have also been pioneered to develop improved compatibility with divertor requirements, and simultaneous access to high performance pedestals. Research has also developed major advances in physics understanding, validating concepts of kinetic damping of ideal MHD instabilities that enable high β operation, identifying how current profile and β influence plasma turbulence in order to validate and improve turbulent transport models, and understanding the physics of energetic particle redistribution due to Alfvénic and other instabilities. These advances have been partnered with development of a rigorous integrated modeling framework used to interpret and validate individual physics models of the various aspects of plasma behavior, and to guide development of improved regimes and upgrades. These tools are also being used to develop and validate concepts for future reactors directly. Having established these foundations, DIII-D is now undergoing a substantial upgrade to raise power, current drive, electron heating and 3-D field capabilities in order to validate this physics and test conceptual solutions in reactor-relevant physics regimes, with a goal to resolve the key scientific and technology questions to enable a decision on a future steady state fusion power plant.

The Large Helical Device: Entering Deuterium Experiment Phase Toward Steady-State Helical Fusion Reactor Based on Achievements in Hydrogen Experiment Phase

The large helical device (LHD) is one of the world’s largest superconducting helical system fusion-experiment devices. Since the start of experiments in 1998, LHD has extended its parameter regime, aiming at achievement of the reactor-relevant plasma conditions and the exploration of related plasma physics in helical-type magnetic configurations. The LHD has also demonstrated its inherent advantage for steady-state operation. Based on these leading developments of helical plasma research, LHD has progressed to the advanced research phase, that is, the deuterium experiment that started in March 2017. It is expected that plasma parameters should be extended toward more reactor-relevant regime, and the related physics research is allowed in such extended regime. Taking this opportunity, parameter extensions such as density, temperature, and steady-state operation achieved in the hydrogen experiment phase are overviewed, along with the initial highlighted results in the very first deuterium experiment campaign in 2017. The design activity of LHD-type steady-state helical fusion reactor is also briefly introduced.

Prospect Toward Steady-State Helical Fusion Reactor Based on Progress of LHD Project Entering the Deuterium Experiment Phase

Large Helical Device (LHD) is one of the world largest superconducting fusion experiment devices, having demonstrated its inherent advantage for steady-state operation since the start of experiments in 1998. LHD has also demonstrated reliable operation of the large-scale superconducting magnet system for almost two decades. Development of the challenging heating systems, such as negative-ion-based neutral beam injection (NBI), high-power and high-frequency electron cyclotron heating, and steady-state ion cyclotron heating, have led to wide-ranging physics and engineering achievements. LHD has progressed to the next stage, that is, the deuterium experiment starting in March 2017, which should further extend plasma parameters toward reactor-relevant regime. For establishing firm basis for designing steady-state helical fusion reactor, advanced physics research, such as on isotope effect, energetic particle confinement, and plasma–wall interaction, will be intensively performed in the deuterium experiments. In an engineering aspect, the upgrade of NBI system has been carried out in preparation to the deuterium experiment, and it should contribute to future NBI development for fusion reactors including ITER. For enhancement of the particle control, the closed divertor system has been installed with pumping capability. Diagnostics for neutron measurements are newly developed and installed for the deuterium experiment. Aligned with all the progress of LHD project in terms of engineering and physics aspects, the conceptual design activity of the LHD-type helical fusion reactor, FFHR-d1, has been programmatically conducted. In parallel to the design study, engineering research and development for the component development have been performed, including those based on employing challenging ideas such as high-temperature superconductor, liquid metal ergodic divertor, and molten-salt breeder blanket. The present status of LHD project entering the deuterium experiment phase is overviewed with putting emphasis on the engineering aspects, and then the engineering research and development activities toward steady-state helical fusion reactor are described.

Turbulence in high-beta ASDEX upgrade advanced scenarios

Recent experiments at ASDEX Upgrade achieve non-inductive operation in full tungsten wall conditions by applying electron cyclotron and neutral beam current drive. These discharges are characterised by a well-measured safety factor profile, which does not drop below one, and a good energy confinement. By reproducing the experimental heat fluxes, nonlinear gyrokinetic simulations suggest that the observed strong peaking of the ion temperature in the core is caused by the stabilising impact of a significant beam ion content, as well as strong electromagnetic effects on turbulent transport. Quasilinear transport models are not yet applicable in this interesting and reactor relevant parameter regime, but available simulation data may serve as a testbed for improvements. As the present plasma is close to the kinetic ballooning (KBM) threshold, elevating the safety factor profile under otherwise identical conditions is proposed to clarify, whether profiles are ultimately limited by KBM turbulence, or by global stability constraints.

High flux irradiations of Li coatings on polycrystalline W and ATJ graphite with D, He, and He-seeded D plasmas at Magnum PSI

Lithium wall conditioning on PFCs (Plasma Facing Components) on a variety of substrate platforms (e.g. graphite, Mo, etc.) has resulted in improved plasma performance on multiple magnetic fusion devices. On graphite, this improvement occurs through the control of retention and recycling of hydrogen at the plasma–material interface by the chemical bonding of Li, O, and D at the surface. Moderate fluence (1×1021m−2) studies of Li on W, performed in PRIHSM (Particle Radiation in Soft and Hard Matter), demonstrated that H retention is similar to Li on ATJ graphite but He ions, when mixed in a D beam, can inhibit the retention. To expand these studies closer to reactor relevant regimes like inside ITER, irradiations were carried out in Magnum-PSI at DIFFER up to fluences of ∼1025m−2 with D, He, and He-seeded D plasmas (He 5–10%). Results show that D is still retained at higher fluxes and fluences.

Two steps towards the realization of fusion: New plasma configurations in the TCV tokamak and its ongoing upgrades

To realise the potential of fusion as an abundant energy source, several challenges remain. The TCV tokamak, featuring high shaping capability and a flexible heating system, is strongly contributing to solving these challenges. A fundamental challenge remains in controlling heat exhaust from the plasma. ITER's currently foreseen operational regime implies heat flows to the plasma facing materials that are not compatible with a commercial fusion reactor. TCV has demonstrated alternative plasma divertor configurations, termed “snowflakes”, that have the potential to strongly reduce the heat flux towards the vessel walls. Measurements of particle fluxes, together with IR camera imaging, show a clear reduction of the peak heat fluxes to the walls when the exhaust power is particularly large and a reduction of the heat fluxes most needed. Another challenge lies in the control of plasma instabilities and turbulence in reactor relevant operational regimes. To address this issue, TCV is presently complementing its electron heating system with an ion heating system: a 1MW neutral beam injector. With these ion and electron heating capabilities, TCV will be able to operate high temperature, reactor relevant, plasmas with of ion to electron temperature ratios.

Two steps towards the realization of fusion: New plasma configurations in the TCV tokamak and its ongoing upgrades

To realise the potential of fusion as an abundant energy source, several challenges remain. The TCV tokamak, featuring high shaping capability and a flexible heating system, is strongly contributing to solving these challenges. A fundamental challenge remains in controlling heat exhaust from the plasma. ITER's currently foreseen operational regime implies heat flows to the plasma facing materials that are not compatible with a commercial fusion reactor. TCV has demonstrated alternative plasma divertor configurations, termed “snowflakes”, that have the potential to strongly reduce the heat flux towards the vessel walls. Measurements of particle fluxes, together with IR camera imaging, show a clear reduction of the peak heat fluxes to the walls when the exhaust power is particularly large and a reduction of the heat fluxes most needed. Another challenge lies in the control of plasma instabilities and turbulence in reactor relevant operational regimes. To address this issue, TCV is presently complementing its electron heating system with an ion heating system: a 1MW neutral beam injector. With these ion and electron heating capabilities, TCV will be able to operate high temperature, reactor relevant, plasmas with of ion to electron temperature ratios.

Plasma–neutrals simulations of linear configurations for PSI studies in reactor relevant regimes

Abstract Coupled fluid plasma and kinetic Monte Carlo neutrals simulations in a linear configuration are reported. The configuration mimics the tokamak divertor plasma channel contacting a target surface with nearby wall. We calculate the parameters of the source plasmas, 3–5 m from the target, required to produce high recycling target plasmas recently simulated for ITER. It is shown that the source plasma needs to deliver heat fluxes of 10–20 MW/m 2 , ion fluxes of 10 23 /m 2 /s, densities of 2–6 × 10 19 /m 3 , and electron and ion temperatures of 15–30 eV over a plasma radius of 5–6 cm. The neutral H and H 2 fluxes to the vessel wall are calculated to be comparable to those measured in the divertor regions of today’s tokamaks. These results identify some design features for a prospective plasma material test station and the research required for this plasma source.

Development of Integrated Transport Analysis Suite for LHD Plasmas Towards Transport Model Validation and Increased Predictability

In this study, the integrated transport analysis suite, TASK3D-a, was developed to enhance the physics understanding and accurate discussion of the Large Helical Device (LHD) experiment toward facilitating transport model validation. Steady-state and dynamic (transient) transport analyses of NBI (neutral-beam-injection)-heated LHD plasmas have been greatly facilitated by this suite. This will increase the predictability of the transport properties of LHD plasmas toward reactor-relevant regimes and reactor-scale plasmas.

Production of internal transport barriers via self-generated mean flows in Alcator C-Mod

New results suggest that changes observed in the intrinsic toroidal rotation influence the internal transport barrier (ITB) formation in the Alcator C-Mod tokamak [E. S. Marmar and Alcator C-Mod group, Fusion Sci. Technol. 51, 261 (2007)]. These arise when the resonance for ion cyclotron range of frequencies (ICRF) minority heating is positioned off-axis at or outside of the plasma half-radius. These ITBs form in a reactor relevant regime, without particle or momentum injection, with Ti ≈ Te, and with monotonic q profiles (qmin < 1). C-Mod H-mode plasmas exhibit strong intrinsic co-current rotation that increases with increasing stored energy without external drive. When the resonance position is moved off-axis, the rotation decreases in the center of the plasma resulting in a radial toroidal rotation profile with a central well which deepens and moves farther off-axis when the ICRF resonance location reaches the plasma half-radius. This profile results in strong E × B shear (>1.5 × 105 rad/s) in the region where the ITB foot is observed. Gyrokinetic analyses indicate that this spontaneous shearing rate is comparable to the linear ion temperature gradient (ITG) growth rate at the ITB location and is sufficient to reduce the turbulent particle and energy transport. New and detailed measurement of the ion temperature demonstrates that the radial profile flattens as the ICRF resonance position moves off axis, decreasing the drive for the ITG the instability as well. These results are the first evidence that intrinsic rotation can affect confinement in ITB plasmas.

A scoping study of the application of neutral beam heating on the TCV tokamak

The TCV tokamak contributes to the physics understanding of fusion plasmas, broadening the parameter range of reactor relevant regimes, by investigations based on an extensive use of the existing main experimental tools: flexible shaping and high power real time-controllable electron cyclotron heating (ECH) and current drive (ECCD) systems. A proposed implementation of direct ion heating on the TCV by the installation of a 20–35 keV neutral beam injection (NBI) with a total power of 1–3 MW would permit an extension of the accessible range of ion to electron temperatures ( T i/ T e ∼ 0.1–0.8) to well beyond unity, depending on the NBI/ECH mix and the plasma density. A NBI system would provide TCV with a tool for plasma study at reactor relevant T i/ T e ratios ∼1 and in investigating fast ion and MHD physics together with the effects of plasma rotation and high plasma β scenarios. The feasibility studies for a NBI heating on TCV presented in this paper were undertaken to construct a specification for the neutral beam injectors together with an experimental geometry for possible operational scenarios.

Goal and Achievements of Large Helical Device Project

The Large Helical Device (LHD) is a heliotron-type device employing large-scale superconducting magnets to enable advanced studies of net-current-free plasmas. The major goal of the LHD experiment is to demonstrate the high performance of helical plasmas in a reactor-relevant plasma regime. Engineering achievements and operational experience greatly contribute to the technological basis for a fusion energy reactor. Thorough exploration for scientific and systematic understanding of the physics in the LHD is an important step to a helical fusion reactor. In the 12 years since the initial operation, the physics database as well as operational experience has been accumulated, and the advantages of stable and steady-state features have been demonstrated by the combination of advanced engineering and the intrinsic physical advantages of helical systems in the LHD. The cryogenic system has been operated for 56 000 h in total without any serious trouble and routinely provides a confining magnetic field up to 2.96 T in steady state. The heating capability to date is 23 MW of neutral beam injection, 3 MW of ion cyclotron resonance frequency, and 2.5 MW of electron cyclotron resonance heating. Highlighted physical achievements are high beta (5.1%), high density (1.2 × 1021 m−3), and steady-state operation (3200 s with 490 kW).

The Fusion Advanced Studies Torus (FAST): a proposal for an ITER satellite facility in support of the development of fusion energy**This article was originally submitted as a paper relating to the 2008 Fusion Energy Conference (Geneva). The Cumulative Index of these papers can be found in issue 3 2010 of Nuclear Fusion http://iopscience.iop.org/0029-5515/50/3/039901.

FAST is a new machine proposed to support ITER experimental exploitation as well as to anticipate DEMO relevant physics and technology. FAST is aimed at studying, under burning plasma relevant conditions, fast particle (FP) physics, plasma operations and plasma wall interaction in an integrated way. FAST has the capability to approach all the ITER scenarios significantly closer than the present day experiments using deuterium plasmas. The necessity of achieving ITER relevant performance with a moderate cost has led to conceiving a compact tokamak (R = 1.82 m, a = 0.64 m) with high toroidal field (BT up to 8.5 T) and plasma current (Ip up to 8 MA). In order to study FP behaviours under conditions similar to those of ITER, the project has been provided with a dominant ion cyclotron resonance heating system (ICRH; 30 MW on the plasma). Moreover, the experiment foresees the use of 6 MW of lower hybrid (LHCD), essentially for plasma control and for non-inductive current drive, and of electron cyclotron resonance heating (ECRH, 4 MW) for localized electron heating and plasma control. The ports have been designed to accommodate up to 10 MW of negative neutral beams (NNBI) in the energy range 0.5–1 MeV. The total power input will be in the 30–40 MW range under different plasma scenarios with a wall power load comparable to that of ITER (P/R ∼ 22 MW m−1). All the ITER scenarios will be studied: from the reference H mode, with plasma edge and ELMs characteristics similar to the ITER ones (Q up to ≈1.5), to a full current drive scenario, lasting around 170 s. The first wall (FW) as well as the divertor plates will be of tungsten in order to ensure reactor relevant operation regimes. The divertor itself is designed to be completely removable by remote handling. This will allow us to study (in view of DEMO) the behaviour of innovative divertor concepts, such as those based on liquid lithium.FAST is capable of operating with very long pulses, up to 170 s, despite being a copper machine. The magnets initial operation temperature is 30 K, with cooling provided by helium gas. The in vessel components, namely FW and divertor, are actively cooled by pressurized water above 80 °C. The same water is also used to bake the vacuum vessel. FAST is equipped with ferromagnetic inserts to keep the toroidal field magnet ripple down to 0.3%.

Overview of physics research on the TCV tokamak

The Tokamak à Configuration Variable (TCV) tokamak is equipped with high-power (4.5 MW), real-time-controllable EC systems and flexible shaping, and plays an important role in fusion research by broadening the parameter range of reactor relevant regimes, by investigating tokamak physics questions and by developing new control tools. Steady-state discharges are achieved, in which the current is entirely self-generated through the bootstrap mechanism, a fundamental ingredient for ITER steady-state operation. The discharge remains quiescent over several current redistribution times, demonstrating that a self-consistent, ‘bootstrap-aligned’ equilibrium state is possible. Electron internal transport barrier regimes sustained by EC current drive have also been explored. MHD activity is shown to be crucial in scenarios characterized by large and slow oscillations in plasma confinement, which in turn can be modified by small Ohmic current perturbations altering the barrier strength. In studies of the relation between anomalous transport and plasma shape, the observed dependences of the electron thermal diffusivity on triangularity (direct) and collisionality (inverse) are qualitatively reproduced by non-linear gyro-kinetic simulations and shown to be governed by TEM turbulence. Parallel SOL flows are studied for their importance for material migration. Flow profiles are measured using a reciprocating Mach probe by changing from lower to upper single-null diverted equilibria and shifting the plasmas vertically. The dominant, field-direction-dependent Pfirsch–Schlüter component is found to be in good agreement with theoretical predictions. A field-direction-independent component is identified and is consistent with flows generated by transient over-pressure due to ballooning-like interchange turbulence. Initial high-resolution infrared images confirm that ELMs have a filamentary structure, while fast, localized radiation measurements reveal that ELM activity first appears in the X-point region. Real time control techniques are currently being applied to EC multiple independent power supplies and beam launchers, e.g. to control the plasma current in fully non-inductive conditions, and the plasma elongation through current broadening by far-off-axis heating at constant shaping field.