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In the original version of this Article, in the fifth paragraph of the section titled “Qneo for W7-X discharge 20180918.045,” errors appeared in reference to the configurations shown in the curves of Fig. 3a and associated Fig. 3b−e panels. In the sentence originally reading “This comparison is provided in Fig. 3a for the configurations (Fig. 3b) W7-X standard, (Fig. 3c) W7-X high-mirror, (Fig. 3d) LHD R0 = 3.6 m, and (Fig. 3e) LHD R0 = 3.75 m....,” the references to Fig. 3b−e should instead have referred to the lettered curves “b,c,d,e” shown in the graph of Fig. 3a. The text has now been amended to read: “This comparison is provided in Fig. 3a for the configurations (b) W7-X standard, (c) W7-X high-mirror, (d) LHD R0 = 3.6 m, and (e) LHD R0 = 3.75 m….” Additionally, a member of the W7-X Team was incorrectly listed as “M. Sanchez”; their name has been corrected to “E. Sánchez.” The original Article has been corrected online.

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.

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.

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.

We demonstrate favorable stability properties of maximum-J stellarators, exemplified by the Wendelstein 7-X (W7-X) device, in scenarios with low plasma beta. A large number of electrostatic linear gyrokinetic simulations are conducted to scan the relevant parameter space for different configurations, resulting in stability maps that account for the key micro-instabilities thought to drive turbulent transport. These maps exhibit a ‘stability valley’ in the region where the normalized ion temperature gradient is roughly equal to the normalized density gradient. In this valley, the electrostatic instabilities are partly suppressed thanks to the maximum-J property of the W7-X field. This property varies across different W7-X configurations, and this measurable difference is demonstrated to affect the size of the stability valley. Finally, the impact of the isotope effect and collisions on the valley is examined.

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 injection of cryogenic pellets into a magnetically confined plasma is shown to be accompanied by a considerable transfer of thermal energy from the electrons in the background plasma to the ions. The resulting ion heating can be significant, particularly in plasmas with disparate electron and ion temperatures, and can affect the energy balance of the plasma. In recent Wendelstein 7-X experiments, this mechanism can account for a substantial fraction of the ion heating power during pellet injection.

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.

Consistent physics and engineering models are required for many use cases at Wendelstein 7-X. Web service technology is used to provide convenient access to common software components and experiment data. The usage of standard web protocols allows the integration in client software written in any programming language. This leads to a better reusability of expert codes and experiment data by a large international team. At W7-X, web services provide common functions such as Biot–Savart calculations and magnetic field line tracing. Further services provide access to databases for diagnostic data or machine models. A new web service was implemented for the creation of 3D magnetohydrodynamic (MHD) equilibrium calculations and a user-friendly access to the results. Consistent interface definitions allow an orchestration of the W7-X web services by user code, as well as a seamless integration into a data analysis framework. In this paper, we present the current state of the MHD equilibrium web services at Wendelstein 7-X.

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.

The efficiency of electron cyclotron heating is determined by the microwave absorption of the plasma. Good microwave absorption is also crucial for the machine safety. In this paper we present a method of evaluating the microwave absorption coefficient from stray radiation measurements. The discussed method is computationally simple and can be applied potentially in real time. Evolution of the second harmonic extraordinary mode (X2) microwave absorption coefficient in Wendelstein 7-X during the start-up phase is presented, as well as an estimate of the absorption coefficient for the second harmonic ordinary mode (O2) wave.

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.

We present a theoretical/numerical predictive analysis of the performance of the W7-X island divertor under conditions of detachment characterized by intensive radiation. The analysis is based on EMC3-Eirene simulations and the earlier W7-AS experimental and numerical experience. Carbon is employed as a representative radiator. The associated drawbacks, i.e. core contamination and recycling degradation (reduced recycling flux), are evaluated by determining the carbon density at the last closed flux surface (LCFS) and the neutral pressure in the divertor chamber. Optimum conditions are explored in both configuration and plasma parameter space. This study aims to identify the key geometric/magnetic and plasma parameters that affect the performance of detached plasmas in W7-X. Emphasis is placed on what occurs when the islands are enlarged far beyond the maximum size available in W7-AS and whether an island size limit for optimal detachment operation exists, and why. Further issues addressed are the power removal ability of the W7-X edge islands, potentially limiting factors, compatibility between particle and power exhaust, and particle refueling capability of the recycling neutrals.

System codes are commonly employed for the analysis and conceptual design of fusion reactors. For the helical-axis advanced stellarator (HELIAS) line, a new set of system code models have been developed to account for the stellarator-specific 3-D aspects. The models have recently been implemented in the systems code PROCESS and verified with respect to different test cases. After having established confidence in the stellarator models, system studies were carried out for the five-field-period HELIAS case to define the accessible reactor design window. In the multidimensional physics and engineering parameter space, sensitivity studies are carried out for the reactor regime to ascertain tradeoffs between different parameters and costs. Exemplary design points are analyzed in more detail using the plasma operation contour approach, which, for example, can be used to determine the optimum start-up path to ignition. Finally, with a common set of nondevice-specific models, the PROCESS framework allows a direct comparison of tokamaks and stellarators. Although the five-period HELIAS is a larger machine in terms of major radius, the required mass for both concepts is comparable leading to similar construction costs.

Systems codes are commonly employed for the analysis and conceptual design of next-step burning-plasma devices. For the helical-axis advanced stellarator (HELIAS) line a new set of systems code models have been developed to account for the stellarator-specific 3D aspects. The models have recently been implemented in the systems code PROCESS and verified with respect to different test cases.

The results from the investigation of neoclassical core transport and the role of the radial electric field profile (Er) in the first operational phase of the Wendelstein 7-X (W7-X) stellarator are presented. In stellarator plasmas, the details of the Er profile are expected to have a strong effect on both the particle and heat fluxes. Investigation of the radial electric field is important in understanding neoclassical transport and in validation of neoclassical calculations. The radial electric field is closely related to the perpendicular plasma flow (u⊥) through the force balance equation. This allows the radial electric field to be inferred from measurements of the perpendicular flow velocity, which can be measured using the x-ray imaging crystal spectrometer and correlation reflectometry diagnostics. Large changes in the perpendicular rotation, on the order of Δu⊥∼ 5 km/s (ΔEr ∼ 12 kV/m), have been observed within a set of experiments where the heating power was stepped down from 2 MW to 0.6 MW. The...

The Electron Cyclotron Current Drive (ECCD) efficiency is usually modeled in the collisionless limit. While such models are sufficient for plasmas with rather low collisionality, they might underestimate the current drive in plasmas at low temperatures likely to occur at the initial phase of high density device operation. In this paper, the impact of finite collisionality effects on the wave-induced current drive is studied for a high-mirror configuration of Wendelstein 7-X using a combination of the drift kinetic equation solver NEO-2 and the ray-tracing code TRAVIS for a realistic set of plasma parameter profiles. The generalized Spitzer function, which describes the ECCD efficiency in phase space, is modeled with help of NEO-2, which uses the full linearized Coulomb collision operator, including energy and momentum conservation. Within this approach, the linearized drift kinetic equation is solved by means of the field line integration technique without any simplifications on device geometry. The resul...

To increase flexibility in ECRH experiments on the helically symmetric experiment (HSX), a second gyrotron and transmission line have been installed. The second antenna includes a steerable mirror for off-axis heating, and the launched power may be modulated for use in heat pulse propagation experiments. The extraordinary wave at the second harmonic of the electron gyrofrequency or the ordinary wave at the fundamental resonance are used for plasma start-up and heating on HSX. The tracing visualized ray tracing code (Marushchenko et al 2007 Plasma Fusion Res. 2 S1129) is used to estimate single-pass absorption and to model multi-pass wave damping in the three-dimensional HSX geometry. The single-pass absorption of the ordinary wave at the fundamental resonance is calculated to be as high as 30%, while measurements of the total absorption indicate that 45% of the launched power is absorbed. A multi-pass ray tracing model correctly predicts the experimental absorption and indicates that the launched power is absorbed within the plasma core ().