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In the previous divertor campaign, the Wendelstein 7-X (W7-X) device injected 3.6 MW of neutral beam heating power allowing for the achievement of densities approaching 2 × 1020 m−3, and providing the first initial assessment of fast ion confinement in a drift optimized stellarator. The neutral beam injection (NBI) system on W7-X is comprised of two beam boxes with space for four radio frequency sources each. The 3.6 MW of heating reported in this work was achieved with two sources in the NI21 beam box. The effect of combined electron-cyclotron resonance heating (ECRH) and NBI was explored through a series of discharges varying both NBI and ECRH power. Discharges without ECRH saw a linear increase in the line-integrated plasma density, and strong peaking of the core density, over the discharge duration. The presence of 1 MW of ECRH power was found to be sufficient to control a continuous density rise during NBI operation. Simulations of fast ion wall loads were found to be consistent with experimental infrared camera images during operation. In general, NBI discharges were free from the presence of fast ion induced Alfvénic activity, consistent with low beam betas. These experiments provide data for future scenario development and initial assessment of fast-ion confinement in W7-X, a key topic of the project.

During its second operational campaign (OP 1.2a) W7-X operated with an edge island divertor. Much higher densities (1020m−3) than in to the previous, limiter campaign were achieved. In order to asses this improved performance and the nature of transport with regard to the expected higher densities and other parameters such as the electron and ion temperature, the electric field and plasma flow, measurements at the plasma edge were conducted with reciprocating probe, mounted on a Manipulator. The probe contains a set of Langmuir pins, Mach probes, compensation coils and a set of 3D pick up coils for the measurement of the local magnetic field and fluctuations, and an ion sensitive probe. It has been observed, in addition to differences due to the magnetic configurations, that the magnetic topology is sensitive to the plasma conditions, especially to changes due to a evolving toroidal plasma current. Measurements from divertor cameras showed that an increasing toroidal current in the standard configuration broadened the heat flux and shifted the strike line considerably. These down-stream observation can be complemented with the measurements at the mid-plane with the manipulator mounted probes. During the experiments the edge island remnant was identified with the measured edge plasma profiles and compared with the predictions from the field line tracing code. Those electron densities and temperatures have been used as an input for understanding edge transport with the EMC3-EIRENE modeling. The results from EMC3-EIRENE also help improve the caculated density profiles, since the modelling provides a profile of the effective charge Zeff. Keywords: W7-X, Langmuir probes, Magnetic islands, Plasma edge transport

The optimized superconducting stellarator device Wendelstein 7-X (with major radius , minor radius , and plasma volume) restarted operation after the assembly of a graphite heat shield and 10 inertially cooled island divertor modules. This paper reports on the results from the first high-performance plasma operation. Glow discharge conditioning and ECRH conditioning discharges in helium turned out to be important for density and edge radiation control. Plasma densities of with central electron temperatures were routinely achieved with hydrogen gas fueling, frequently terminated by a radiative collapse. In a first stage, plasma densities up to were reached with hydrogen pellet injection and helium gas fueling. Here, the ions are indirectly heated, and at a central density of a temperature of with was transiently accomplished, which corresponds to with a peak diamagnetic energy of and volume-averaged normalized plasma pressure . The routine access to high plasma densities was opened with boronization of the first wall. After boronization, the oxygen impurity content was reduced by a factor of 10, the carbon impurity content by a factor of 5. The reduced (edge) plasma radiation level gives routinely access to higher densities without radiation collapse, e.g. well above line integrated density and central temperatures at moderate ECRH power. Both X2 and O2 mode ECRH schemes were successfully applied. Core turbulence was measured with a phase contrast imaging diagnostic and suppression of turbulence during pellet injection was observed.

Intrinsic, diffusion-driven toroidal (bootstrap) currents between −7 and 17 kA were measured in the optimised stellarator Wendelstein 7-X (W7-X) for several magnetic configurations at line-integrated plasma densities between 2 × 1019 and 1.6 × 1020 m−2 and heating powers between 0.5 and 6 MW. The sign of the bootstrap current changes with field reversal, and its magnitude decreases with the mirror term and the rotational transform of the magnetic configuration as well as the line-integrated plasma density, and increases with the heating power. Both the absolute values and the dependencies agree with neoclassical calculations within their respective uncertainties, confirming the neoclassical optimisation of W7-X for small bootstrap currents.

A confinement database with mainly electron-heated hydrogen plasmas from ASDEX Upgrade and Wendelstein 7-X was assembled. Stellarator confinement scaling expressions describe both standard discharges in the stellarator and L-mode plasmas in the tokamak similarly well and indicate a similar quality of energy confinement in both devices. While the energy confinement time in ASDEX Upgrade benefits from the smaller aspect ratio of the device, the transport coefficients in Wendelstein 7-X appear to be smaller possibly due to reduced average magnetic field curvature. A physics based confinement scaling is derived from a model that successfully describes transport in tokamaks. The dimensionally correct scaling has very similar parameter dependencies as the stellarator scalings and reproduces also the trends in the data from ITER L- and H-mode databases reasonably well. On the basis of this scaling, which represents the confinement times of the present data base, average tokamak L-mode and H-mode confinement is 7% lower and 76% higher, respectively.

Magnetic reconnection in W7-X is studied by means of global numerical simulations in a series of models of increasing complexity. The magnetic geometry ranges from that of a cylinder to the full three-dimensional field of W7-X, and the equations solved range from ideal magnetohydrodynamics (MHD) to gyrokinetics. We simulate plasmas from the first operation phase with electron cyclotron current drive (ECCD). These are characterized by an equilibrium magnetic field featuring an ECCD-distorted ‘humped’ profile of the rotational transformι, withι= 1 in two radial locations. Such plasmas generally show sawtooth activity, hence motivating the present study. We pay particular attention to the role of equilibrium current density gradients in the destabilization of reconnecting modes. When the equilibrium temperature and density gradients are artificially suppressed (to eliminate the pressure gradient drive), the perturbed electrostatic potential is radially localized between the locations at whichι= 1. This is shown with a purely collisionless gyrokinetic model, in cylindrical geometry. In the real toroidal geometry of W7-X, for a non-ideal MHD model including a uniform resistivity, electron inertia and (numerical) viscosity, the same qualitative behaviour is observed. In particular, even if a resonant (m,n) = (1, −1) perturbation is initialized, the most unstable mode is the (m,n) = (−4, 4), wheremandnare the poloidal and toroidal mode numbers, respectively. Other modes are destabilized due to geometric coupling. The growth rate of this instability scales asη1/3, whereηis the plasma resistivity, thus suggesting that ECCD drives ideal MHD stable W7-X plasmas towards non-ideal marginality. An ideal magnetohydrodynamic analysis confirms the result. A fluid-kinetic hybrid version of theEUTERPEcode shows that gyrokinetic ions have a stabilizing effect on these modes. For W7-X relevant collisionalities, the growth rate scales linearly with the electron skin depth,de. Implications of our results for sawtoothing W7-X operation are discussed.

During the interaction of the scrape off layer (SOL) plasma with the first wall the evolution of both wall and plasma are tightly coupled: The erosion of the first wall leads to an impurity concentration in the plasma which affects the particle and power balance in the plasma. In turn the impurities, when leaving the plasma via transport, can form deposits and mixed materials, far away from their initial source location. These deposits can be eroded, allowing the impurities to stepwise migrate through the fusion device until they end up at a location where the plasma at the wall is cold enough and no further erosion occurs. To describe these processes an integrated model of surface evolution and plasma transport of impurities is needed. The WallDYN code achieves this required coupling of processes by parameterising the output of surface evolution- and plasma impurity-migration-codes by analytical models. For a given fixed background plasma it evolves the surface composition, derives impurity flux into and from the plasma and can from this derive the impurity densities in the plasma. This paper will show the importance of including this recycling of impurities at the wall in impurity migration modelling: The 13CH 4 seeding experiment performed in the Wendelstein 7-X Stellerator is modelled using the recent extension of the WallDYN code to 3D plasma and wall geometries. A comparison with post mortem analysis of the 13C deposition shows both qualitative and quantitative agreement with the WallDYN calculations.

This paper focuses on the design, installation and test of the first version of the Fast Interlock System (FIS) of W7-X. The structure of the system and the basic components including some safety-relevant plasma diagnostics are described. The very first practical experiences with the FIS have been gained during the operation phase 1.2b in summer 2018. Next to the proof of a satisfactory reaction, scalability and operability, the focus also was on the assessment of the reaction times of the core system and the various safety-relevant plasma diagnostics during several experimental situations. A couple of planned measures to improve and extend the FIS will be presented in this paper.

A significant improvement of plasma parameters in the optimized stellarator W7-X is found after injections of frozen hydrogen pellets. The ion temperature in the post-pellet phase exceeds 3 keV with 5 MW of electron heating and the global energy confinement time surpasses the empirical ISS04-scaling. The plasma parameters realized in such experiments are significantly above those in comparable gas-fuelled discharges. In this paper, we present details of these pellet experiments and discuss the main plasma properties during the enhanced confinement phases. Local power balance is applied to show that the heat transport in post-pellet phases is close to the neoclassical level for the ion channel and is about a factor of two above that level for the combined losses. In comparable gas-fuelled discharges, the heat transport is by about ten times larger than the neoclassical level, and thus is largely anomalous. It is further observed that the improvement in the transport is related to the peaked density profiles that lead to a stabilization of the ion-scale turbulence.

A quasi-coherent mode (QCM) near the plasma edge is characterized experimentally by measuring the density fluctuation via a poloidal correlation reflectometer (PCR) in Wendelstein 7-X. This QCM frequency descends from 25 kHz near the separatrix to 10 kHz as the cutoff position penetrates toward the core. A similar mode structure is also observed in the spectra of electron cyclotron emission edge channels and Mirnov coil signals, which shows a clear correlation with respect to the PCR measurement. The mode evaluation yields a range for the poloidal mode number of , and for the normalized mode scale , which suggest a drift type instability. The QCM structure appears in the standard and narrow-mirror configurations (both with the rotational transform in the boundary), but is absent in the high- configuration (boundary ). The impact of plasma parameters on the QCM is discussed. It is observed that the QCM frequency and amplitude can be affected by the edge magnetic topology and plasma parameters. The shifts upward during the negative and non-stellarator-symmetric control coil current scans, which may be attributed to the effect of the shear, Er, and ne in the plasma edge region.

While the low frequency part of the response of the Rogowski coils mounted in the stellarator Wendelstein 7-X has been calculated, the high frequency part is planned to be calibrated by means of an actively applied current. In a staged approach different conductors have been designed to be mounted inside the plasma vessel and an AC current of 1 kA is foreseen to be driven through them at frequencies up to 1 kHz to simulate the plasma current.

Observations on confinement in the first experimental campaign on the optimized Stellarator Wendelstein 7-X are summarized. In this phase W7-X was equipped with five inboard limiters only and thus the discharge length restricted to avoid local overheating. Stationary plasmas are limited to low densities

For the first time, the optimized stellarator Wendelstein 7-X has operated with an island divertor. An operation regime in hydrogen was found in which the total plasma radiation approached the absorbed heating power without noticeable loss of stored energy. The divertor thermography recorded simultaneously a strong reduction of the heat load on all divertor targets, indicating almost complete power detachment. This operation regime was stably sustained over several energy confinement times until the preprogrammed end of the discharge. The plasma radiation is mainly due to oxygen and is located at the plasma edge. This plasma scenario is reproducible and robust at various heating powers, plasma densities, and gas fueling locations. These experimental results show that the island divertor concept actually works and displays good power dissipation potential, producing a promising exhaust concept for the stellarator reactor line.

In the Wendelstein 7-X stellarator, up to 7MW of power are delivered to the plasma by an electron cyclotron resonance heating system consisting of ten 140 GHz gyrotrons [1]. Due to the flexible front steering mirror of each beam line, the power deposition can be varied over the whole plasma radius and is optionally combinable with additional current drive. This flexibility, together with small toroidal currents in the stellarator, makes W7-X a perfect testbed for electron cyclotron current drive (ECCD) experiments, which have been successfully accomplished during the first two experimental campaigns OP1.1 and OP1.2a. Long discharges (lasting up to 30s) have been performed in OP1.2a, thus allowing the study of the current drive time evolution and the possibility to compensate the bootstrap current. ECCD efficiency has been studied using different power deposition profiles combined with a variation of the injection angles in relation to the magnetic field. During ECCD experiments, saw-tooth-like oscillations have been observed. Depending on the driven current density, ECCD can significantly modify the rotational transform (iota) profile, which can locally reach low order rational, thus triggering plasma instabilities. Different current density profiles have been tested, in order to try to understand the main trigger parameter for the instabilities. In particular, effects caused by current density gradient have been investigated producing both co- and counter-current drive at different radial positions: the total current drive is negligible, but a strong current gradient arises by driving currents in opposite directions. In this work an overview of ECCD operations in OP1.2a is given and first results, comparing different diagnostics, are presented. An initial 1-D model, coupled with the ray tracer TRAVIS, is developed, in order to have an estimation of current diffusion times and the radial position where a low order rational crosses the disturbed iota profile.

Wendelstein 7-X is a highly optimized stellarator that went into operation in 2015. With a 30 cubic meter volume, a superconducting coil system operating at 2.5 T, and steady-state heating capability of eventually up to 10 MW, it was built to demonstrate the benefits of optimized stellarators at parameters approaching those of a fusion power plant. We report here on the first results with the test divertor installed, during the second operation phase, which was performed in the second half of 2017. Operation with a divertor, and the addition of several new fueling systems, allowed higher density operation in hydrogen as well as helium. The effects that higher density operation had on both divertor operation and global confinement will be described. In particular, at high densities detachment was observed, and the highest fusion triple product for a stellarator was achieved.

The control of rotational transform in Wendelstein 7-X (W7-X) is key to both the island divertor operation and safety of plasma facing components. The island divenor concept in W7-X relies on an edge flux surface with rotational transform of t = 1 resonating with an intrinsic n/m = 5/5 resonance to form a five lobed island chain. This island chain intersects with divertor plates to give rise to the island divertor. Changes in the relative position of the rational surface and the divertor plates can result in changes in divenor performance, thus the control of the rotational transform is essential to operation of the W7-X device. During the first divertor campaign electromagnetic loads resulted in elastic deformations of the shaped modular stellarator coils. Such deformations made these coils more planar, reducing the vacuum rotational transform, subsequently shifting the t = 1 resonance outward. Unintended plasma wall interactions provided the first clear evidence of this effect during plasma operation. Flux surface measurements were utilized to estimate the correct level of current in the planar coils for correction of t, and found to be around similar to-100 A. Scans the planar coil current for iota correction were performed during plasma operation. These measurements suggest planar coil currents between -250 and similar to 0 A would place the strike lines at the designed distance from the pumping gaps. Divertor Langmuir and upstream probe measurements confirm these estimates along with measurements of divertor neutral gas pressure.

The paper presents experimental observations and simulations for the effects of toroidal plasma current on divertor power depositions on W7-X. With increasing toroidal current accompanying changes in the island geometry result in a sweep of the strike line and a redistribution of the heat flux footprints. Good agreement between experiments, which partly used electron cyclotron current drive to generate an additional toroidal current contribution, and modelling using field line tracing in vacuum magnetic fields including an ad-hoc toroidal current on the magnetic axis is found for both standard and low-iota magnetic configurations.

During the two most recent experimental campaigns in the advanced stellarator Wendelstein 7-X (W7-X) (Klinger et al 2017 Plasma Phys. Control. Fusion 59 014018; Bosch et al 2017 Nucl. Fusion 57 116015; Wolf et al 2017 Nucl. Fusion 57 102020; Pedersen et al 2017 Phys. Plasmas 24 0555030) hydrogen ice pellet injection was performed for the first time. In order to investigate the potential of pellet fueling in W7-X and to study the particle deposition in a large stellarator, a blower-gun system was installed with 40 pellets capability. The experience gained with this system will be used for the specification of a future steady-state pellet injector system. One important motivation for a pellet injector (Dibon 2014 Master-Thesis Technical University Munich, Max-Planck Institut IPP) on W7-X is the mitigation of hollow density profiles expected in case of predominant neoclassical transport. For long-pulse operation of up to 30 min, only electron cyclotron resonance heating is available on W7-X. Hence, pellet injection will be the only source for deep particle fueling. Deep particle fueling by pellets in tokamaks is supported by a grad-B drift, if the pellets are injected from the magnetic high-field-side. This approach was tested in W7-X, as well. The injection of series of pellets was also tested. Here, deep fueling is supported for later pellets in the series by the plasma cooling following the initial pellets in the same series. As in earlier experiments in the heliotron LHD (Takeiri et al 2017 Nucl. Fusion 57 102023), deep and rapid fueling could be achieved successfully in W7-X.

openaire: EC/H2020/633053/EU//EUROfusion The neutral beam deposition model in the BEAMS3D code is validated against neutral beam attenuation data from Wendelstein 7-X (W7-X). A set of experimental discharges where the neutral beam injection system of W7-X was utilized were reconstructed. These discharges scanned the magnetic configurations and plasma densities of W7-X. The equilibrium reconstructions were performed using STELLOPT which calculates three-dimensional self-consistent ideal magnetohydrodynamic equilibria and kinetic profiles. These reconstructions leveraged new capabilities to incorporate electron cyclotron emission and X-ray imaging diagnostics in the STELLOPT code. The reconstructed equilibria and profiles served as inputs for BEAMS3D calculations of neutral beam deposition in W7-X. It is found that if reconstructed kinetic profiles are utilized, good agreement between measured and simulated beam attenuation is found. As deposition models provide initial conditions for fast-ion slowing down calculations, this work provides a first steptowards validating our ability to predict fast ion confinement in stellarators.

The soft x-ray tomography diagnostic in the stellarator Wendelstein 7-X consists of twenty pinhole cameras, up–down symmetrically arranged in a poloidal, triangular cross-section of the plasma vessel. The x-ray emissivity is measured with 16 bit amplitude resolution at 2 MHz sampling rate along 360 lines-of-sight by silicon photodiode arrays. In the recent operation campaign data acquisition (DAQ) has been working reliable for the conducted plasma pulse lengths

We present a detailed overview and first results of the new laser blow-off system on the stellarator Wendelstein 7-X. The system allows impurity transport studies by the repetitive and controlled injection of different tracer ions into the plasma edge. A Nd:YAG laser is used to ablate a thin metal film, coated on a glass plate, with a repetition rate of up to 20 Hz. A remote-controlled adjustable optical system allows the variation of the laser spot diameter and enables the spot positioning to non-ablated areas on the target between laser pulses. During first experiments, clear spectral lines from higher ionization stages of the tracer ions have been observed in the X-ray to the extreme ultraviolet spectral range. The temporal behavior of the measured emission allows the estimate of transport properties, e.g., impurity transport times in the order of 100 ms. Although the strong injection of impurities is well detectable, the global plasma parameters are barely changed.

Global confinement properties of the limiter plasmas of the first operational campaign of W7-X are investigated with special focus on the energy confinement and possible operational limits. The energy confinement time was found to be close to expectations from the empirical ISS04 scaling for stellarators. Absolute values up to 160 ms were achieved. This can be considered as a great success for the initial operation of the device. However, a clear degradation of the performance was observed when radiative losses became significant, which was typically the case at low heating power. While a significant improvement of the plasma purity is expected during divertor operation, the presented results underline the importance of the impurity dynamics for the development of high-performance steady-state scenarios in W7-X. Furthermore, comparisons of the global performance properties with neoclassical transport modeling are presented. These studies are not yet fully conclusive, which could indicate that anomalous transport may have played an essential role in the low-density OP1.1 plasmas.

During the first operational phase (OP 1.1) of Wendelstein 7-X (W7-X) electron cyclotron resonance heating (ECRH) was the exclusive heating method and provided plasma start-up, wall conditioning, heating and current drive. Six gyrotrons were commissioned for OP1.1 and used in parallel for plasma operation with a power of up to 4.3 MW. During standard X2-heating the spatially localized power deposition with high power density allowed controlling the radial profiles of the electron temperature and the rotational transform. Even though W7-X was not fully equipped with first wall tiles and operated with a graphite limiter instead of a divertor, electron densities of n e > 3·1019 m-3 could be achieved at electron temperatures of several keV and ion temperatures above 2 keV. These plasma parameters allowed the first demonstration of a multipath O2-heating scenario, which is envisaged for safe operation near the X-cutoff-density of 1.2·1020 m-3 after full commissioning of the ECRH system in the next operation phase OP1.2.

Classical particle drifts are known to have substantial impacts on fluxes of particles and heat through the edge plasmas in both tokamaks and stellarators. Here we present results from the first dedicated investigation of drift effects in the W7-X stellarator. By comparing similar plasma discharges conducted with a forward- and reverse-directed magnetic field, the impacts of drifts could be isolated through the observation of up-down asymmetries in flux profiles on the divertor targets. In low-density plasmas, the radial locations of the strike lines (i.e. peaks in the target heat flux profiles) exhibited discrepancies of up to 3 cm that reversed upon magnetic field reversal. In addition, asymmetric heat loads were observed in regions of the target that are shadowed by other targets from parallel flux from the core plasma. A comparison of these asymmetric features with the footprints of key topological regions of the edge magnetic field on the divertor suggests that the main driver of the asymmetries at low density is poloidal E × B drift due to radial electric fields in the scrape-off layer and private flux region. In higher-density plasmas, upper and lower targets collected non-ambipolar currents with opposite signs that also inverted upon field reversal. Overall, in these experiments, almost all up-down asymmetry could be attributed to the field reversal and, therefore, field-dependent drifts.

Indirectly heated emissive probes have advantages compared to conventional emissive wire probes. We have developed a laser-heated emissive probe consisting of a 1 mm diameter and 2 mm long pin of LaB6, heated by a focussed laser beam of 808 nm wavelength with a power up to 50 W. This probe is smaller and simpler than electrically heated emissive wire probes. Materials of low work function, high temperature stability and longer lifetime such as LaB6 can be used. There is no deformation in a magnetic field and no voltage drop along the probe wire. In this contribution we show the good time resolution of such a probe. Whereas emissive wire probes need two cables and an electric power supply or battery, for our probe one connection to an oscilloscope with high input impedance suffices. Therefore the probe system has a much lower stray capacitance than conventional emissive wire probes (© 2013 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Wendelstein 7-X, a superconducting optimized stellarator built in Greifswald/Germany, started its first plasmas with the last closed flux surface (LCFS) defined by 5 uncooled graphite limiters in December 2015. At the end of the 10 weeks long experimental campaign (OP1.1) more than 20 independent diagnostic systems were in operation, allowing detailed studies of many interesting plasma phenomena. For example, fast neutral gas manometers supported by video cameras (including one fast-frame camera with frame rates of tens of kHz) as well as visible cameras with different interference filters, with field of views covering all ten half-modules of the stellarator, discovered a MARFE-like radiation zone on the inboard side of machine module 4. This structure is presumably triggered by an inadvertent plasma-wall interaction in module 4 resulting in a high impurity influx that terminates some discharges by radiation cooling. The main plasma parameters achieved in OP1.1 exceeded predicted values in discharges of a length reaching 6 s. Although OP1.1 is characterized by short pulses, many of the diagnostics are already designed for quasi-steady state operation of 30 min discharges heated at 10 MW of ECRH. An overview of diagnostic performance for OP1.1 is given, including some highlights from the physics campaigns.

Wendelstein 7-X (W7-X) is a large optimized stellarator (B=2.5T, V=30m3) aiming at demonstrating the reactor relevance of the optimized stellarators. In 2015 W7-X will begin its first operation phase (OP1.1) with five inertially cooled inboard limiters made of graphite. Assuming the heat loads can be spread out evenly between the limiters, 1 second discharges at 2 MW of ECRH heating power could be run in OP1.1. The expected plasma parameters will be sufficient to demonstrate the readiness of the installed diagnostics and even to run a first physics program. The diagnostics available for this first operation phase, including some special limiter diagnostics, and their capabilities are being presented. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

The optimized, superconducting stellarator Wendelstein 7-X went into operation and delivered first measurement data after 15 years of construction and one year commissioning. Errors in the magnet assembly were confirmend to be small. Plasma operation was started with 5 MW electron cyclotron resonance heating (ECRH) power and five inboard limiters. Core plasma values of ${{T}_{\text{e}}}>8$ keV, ${{T}_{\text{i}}}>2$ keV at line-integrated densities $n\approx 3\centerdot {{10}^{19}}~{{\text{m}}^{-2}}$ were achieved, exceeding the original expectations by about a factor of two. Indications for a core-electron-root were found. The energy confinement times are in line with the international stellarator scaling, despite unfavourable wall conditions, i.e. large areas of metal surfaces and particle sources from the limiter close to the plasma volume. Well controlled shorter hydrogen discharges at higher power (4 MW ECRH power for 1 s) and longer discharges at lower power (0.7 MW ECRH power for 6 s) could be routinely established after proper wall conditioning. The fairly large set of diagnostic systems running in the end of the 10 weeks operation campaign provided first insights into expected and unexpected physics of optimized stellarators.

Dissipation range cascades in plasma turbulence are described and spectra are formulated from the scaled attenuation in wavenumber space of the spectral energy transfer rate. This yields spectra characterized by the product of a power law and exponential fall-off, applicable to all scales. Spectral indices of the power law and exponential fall-off depend on the scaling of the dissipation, the strength of the nonlinearity, and nonlocal effects when dissipation rates of multiple fluctuation fields are different. The theory is used to derive spectra for MHD turbulence with magnetic Prandtl number greater than unity, extending previous work. The theory is also applied to generic plasma turbulence by considering the spectrum from damping with arbitrary wavenumber scaling. The latter is relevant to ion temperature gradient turbulence modeled by gyrokinetics. The spectrum in this case has an exponential component that becomes weaker at small scale, giving a power law asymptotically. Results from the theory are c...

A linear experiment dedicated to the study of driven magnetic reconnection is presented. The new device (VINETA II) is suitable for investigating both collisional and near collisionless reconnection. Reconnection is achieved by externally driving magnetic field lines towards an X-point, inducing a current in the background plasma which consequently modifies the magnetic field topology. Owing to the open field line configuration of the experiment, the current is limited by the axial sheath boundary conditions. A plasma gun is used as an additional electron source in order to counterbalance the charge separation effects and supply the required current. Two drive methods are used in the device. First, an oscillating current through two parallel conductors drive the reconnection. Second, a stationary X-point topology is formed by the parallel conductors, and the drive is achieved by an oscillating current through a third conductor. In the first setup, the magnetic field of the axial plasma current dominates the field topology near the X-point throughout most of the drive. The second setup allows for the amplitude of the plasma current as well as the motion of the flux to be set independently of the X-point topology of the parallel conductors.

Wall conditioning is essential in tokamak and stellarator research to achieve plasma performance and reproducibility. This paper presents an overview of recent conditioning results, both from experiments in present devices and modelling, in view of devices with superconducting coils, with focus on W7-X, JT-60SA and ITER. In these devices, the coils stay energised throughout an experimental day or week which demands for new conditioning techniques that work in presence of the nominal field, in addition to the proven conditioning methods such as baking, glow discharge conditioning (GDC) and low-Z wall coating through GDC-plasma, which do not work under such condition. The discussed techniques are RF conditioning without plasma current, both in the ion cyclotron and electron cyclotron range of frequencies, and diverted conditioning plasmas with nested magnetic flux surfaces. Similarities and differences between tokamaks and stellarators are highlighted. Finally a conditional tritium recovery strategy for ITER is proposed based on Ion Cyclotron Wall Conditioning and L-mode plasma results from JET, equipped with an ITER-like wall (beryllium main chamber wall and tungsten divertor).

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3 × 1019 m−3 , central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre. EURATOM 633053

This paper reports on the effect of on- and off-axis heating power deposition on the impurity confinement in purely electron cyclotron resonance heated He plasmas on the stellarator Wendelstein 7-X. Therefore, impurity transport times τ I have been determined after Fe impurity injections by laser ablations and monitoring the temporal impurity emissivities by the x-ray imaging spectrometer HR-XIS. A significant increase of τ I has been observed when changing the power deposition from on- to off-axis heating with energy confinement times τ E being mainly unaffected. In addition, the scaling of impurity transport properties with respect to a variation of heating power P ECRH and electron density n e has been investigated by keeping the heating power deposition on-axis. The observed τ I scaling compares well to known τ I scaling laws observed in other machines. A comparison of τ I and τ E yields an averaged ratio of τ E /τ I = 1.3 and transport times in the range of τ I = 40–130 ms and τ E = 40–190 ms. Comparing those absolute values to neoclassical predictions supports the recently observed nature of anomalous transport in Wendelstein 7-X, given within the up to now investigated operational parameters.

The first physics operation phase on the stellarator experiment Wendelstein 7-X was successfully completed in March 2016 after about 10 weeks of operation. Experiments in this phase were conducted with five graphite limiters as the primary plasma-facing components. Overall, the results were beyond the expectations published shortly before the start of operation [Sunn Pedersen et al., Nucl. Fusion 55, 126001 (2015)] both with respect to parameters reached and with respect to physics themes addressed. We report here on some of the most important plasma experiments that were conducted. The importance of electric fields on global confinement will be discussed, and the obtained results will be compared and contrasted with results from other devices, quantified in terms of the fusion triple product. Expected values for the triple product in future operation phases will also be described and put into a broader fusion perspective.

A 1-2 kHz electromagnetic mode is present in a large fraction of W7-X discharges in the standard magnetic field configurations with 5/5 edge island structure. The power in the mode is significantly reduced in other magnetic configurations. The mode can appear spontaneously and persist for the full duration of a discharge. Its full width at half maximum is typically 100-300 Hz. The 1-2 kHz mode was first observed in visible light emissions from the divertor [1] . Matching spectral peaks were found in measurements of both global and edge plasma parameters, including line-integrated density, diamagnetic energy, and edge floating potential and ion saturation current [2] . An infrared camera recording shows a temperature modulation of ∼2.5 K due to this mode on a divertor tile where the average temperature was 600 °C. Separate correlation analyses of the high-speed camera recordings, Langmuir probe signals, and Mirnov coil signals all indicate a poloidal velocity of 0.5-10 km/s and m=2 as a likely poloidal mode number. Langmuir probes mounted on the Multi-Purpose Manipulator show that the fluctuation is present in one of the magnetic islands over a radial distance of at least 5 millimeters. A database of various diagnostic measurements and dB/dt signals from a Mirnov coil was created using a majority of the discharges from the 2018 campaign on W7-X, in order to shed light on the plasma parameters that are correlated with the existence of the 1-2 kHz mode. The power in the mode is positively correlated to the toroidal current, total external heating power, and core electron temperature. It is negatively correlated with plasma density.