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Wendelstein 7-X (W7-X) completed its second operation phase in December 2017. A large number of diagnostics were operated in nearly 1000 experiment programs (XP) by an international research team. For the documentation of W7-X XPs, a new electronic logbook software has been developed. The software has been designed for the needs of W7-X researchers and engineers: a web-based logbook application for the entire team. For an effective documentation, large parts of the logbook content comes from automatically generated logs, complemented by the team members via web browser or REST API. The W7-X control software generates log entries for an XP during program execution. This includes automatic extraction of configuration information from the planned program, which is later represented as searchable tags within the logbook. In addition, users can add supplementary tags and rich-text comments to all logs. The logbook allows full-text search and range queries for numeric values: both with millisecond response times. A dedicated web page for each XP log contains overview plots from measurement data, generated on the fly from archived data. Separate component logs are created in the same way for diagnostics and machine sub-systems for different use cases, e.g., XPs, standalone tests, or calibrations. The logbook hosts separate pages for device operation and the W7-X components, providing status information and additional features. In this way it effectively supports the operation of W7-X.

This paper reports on experiences gained during the construction of the large superconducting optimized stellarator device Wendelstein 7-X (W7-X). The goal of the project is to demonstrate the fusion reactor potential of optimized stellarators and to operate for the first time fusion-relevant plasmas under full steady-state conditions. In addition, the validity of the engineering concept is to be proven under operating conditions. The actual construction of the device is now running since more than ten years and will require about three further years of assembly and integration. The current technical status of W7-X is reviewed, the major construction challenges are highlighted, and conclusions are drawn from past experiences that may also be useful for other large-scale science projects.

The Wendelstein 7-X modular advanced stellarator is in the commissioning phase at the Max Planck Institute for Plasma Physics in Greifswald, Germany. The focus of the numerical analysis has been shifted from support of the machine design and assembly towards preparation of commissioning steps, assessment of possible field disturbances under operational loads, and exploration of operational limits.

Helium compression and helium enrichment are important for most future fusion reactors since the He ash produced by the nuclear reactions needs to be pumped out to prevent its accumulation in the core of the plasma. (Helium compression is the ratio of the helium density in the pumping plenum to the helium density in the plasma; helium enrichment is the ratio of the helium compression to the hydrogen or deuterium compression.) In this work the B2–EIRENE code is used to model helium compression and enrichment on ASDEX Upgrade with divertor II under a variety of plasma densities, input powers, transport coefficients and divertor geometries so as to determine the range of acceptable operating conditions for effective He removal. A broad range of acceptable plasma conditions is found, though a lower limit of the upstream separatrix density is noted.

A review is presented of the scenarios and physics of advanced tokamak discharges and the associated technical enhancements leading towards enhanced performance and steady state operation in ASDEX Upgrade. A stationary advanced tokamak scenario with internal transport barriers (ITBs) in combination with an H mode barrier and weak shear (qmin ≈ 1) was maintained for 40 confinement times and several internal skin times with HITERL-89PβN ≈ 5. By raising the triangularity of the plasma shape the performance was increased up to HITERL-89PβN ≈ 7.2, but β is still limited by neoclassical tearing modes. The density was raised to close to 50% of the Greenwald density either by edge gas fuelling, causing an increase of the threshold power to sustain an ITB and a decrease of Zeff below 2, or by improved core particle confinement with more triangular plasma shapes without changing the ITB onset conditions. Sufficient helium pumping and no temporal impurity accumulation was observed despite peaked impurity density profiles. MHD modes contribute to making the shear profile stationary. In the ITB/H mode scenario (1,1) fishbones clamp the q value to the vicinity of 1 and avoid sawteeth, while in ITB scenarios with reversed shear (qmin ≈ 2) (2,1) fishbones can clamp the current profile development near the q = 2 surface without causing energy confinement to deteriorate. Double tearing modes act similarly, but lead to substantial confinement losses. Applying central ECRF heating and current drive to beam heated reversed shear ITB discharges shows a substantial effect on MHD stability, affecting the passage of the q profile through qmin = 2 and degrading or prolonging the reversed shear phase, depending on the current drive direction. Moreover, reactor relevant Te ≈ Ti operation with temperatures in excess of 10 keV was achieved with internal transport barriers for both electrons and ions simultaneously. For current profile control ECCD will be supplemented by on-axis fast wave ICCD and off-axis current drive up to 400 kA using NBI (available in 2001) and ICRF mode conversion. Stationary discharges with reversed shear and qmin > 2 should then be possible at a plasma current of 1 MA according to power deposition, current drive and transport calculations.

In 1997 the new `LYRA' divertor went into operation at ASDEX Upgrade and, in parallel, the neutral beam heating power was increased to 20 MW by installation of a second injector leading to a P/R value of 12 MW/m. Experiments have shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. There is an overall reduction of the maximum heat flux in the LYRA divertor by about a factor of 2 compared with the previous open divertor Div I. This reduction is mainly due to increased radiative losses inside the divertor region, which are caused by an effective reflection of hydrogen neutrals into the hot separatrix region. The main channel of radiative loss is carbon radiation, which cools the divertor plasma down to a few electronvolts, where hydrogen radiation losses become significant. The radiative losses preferentially reduce the power flux at the separatrix, leading to early detachment around the strike point position. With increasing density, the detached region extends upwards on the vertical target. The power fraction radiated in the LYRA divertor is around 45% and nearly independent of the heating power. This value is a factor of 2 higher than the typical radiation fraction in Div I. B2-EIRENE modelling of the performed experiments supports the experimental finding and refines the understanding of loss processes in the divertor region.

The change of the divertor plasma behaviour from Div-I to Div-II for ASDEX Upgrade as measured by a set of reconstructed or newly designed divertor diagnostics is presented. The Div-II configuration is characterised – due to the highly inclined target plates – by reflection of neutrals towards the separatrix. Therefore, detachment in Div-II starts rather early localised close to the separatrix. In contrast, the complete global divertor detachment is practically unchanged, because the outer parts of the SOL stay attached even for high densities. The power distribution on the divertor in Div-II is much broader than in Div-I, resulting in a large reduction of peak power loads both in L and H-mode (factor of 2–4) due to larger divertor radiation losses. The larger losses are caused by larger hydrogen losses, enhancement of carbon radiation due to radial transport and convective energy transport into the radiation zone and larger radial energy transport in the divertor. The new Div-II geometry shows larger neutral gas densities in the divertor for the same line averaged density and a much faster helium exhaust rate ( τ ∗ He /τ E ≈4 ) in H-mode. The neon compression is worse compared with helium. No strong effect on impurity compression and overall divertor performance was seen by puff and pump experiments, neither in Div-I nor in Div-II.

Since spring of 1997 ASDEX Upgrade has operated with the new divertor II (LYRA), which is characterised by vertical target plates, tight baffling towards the main chamber, and a dome baffle in the private flux region. The main goal of this divertor modification was to increase the divertor density (of the plasma and of the neutrals) for a given plasma density, whereas the neutral density in the main chamber should be similar or even lower. Pumping occurs now from the private flux region which is linked to the pump chamber by finite conductance below the outer target. A cryo pump (about 100 m 3 /s) has been installed in the outer divertor chamber in addition to the turbo pumps. This divertor scheme is very similar to the one planned for ITER. First experiments with the new divertor indicate that the compression of helium, C He (the ratio of neutral He density in the divertor to the He density at the plasma edge) is much higher in the new divertor, while the neon exhaust efficiency has decreased. The cryo pump can strongly enhance the scrape-off layer flow of deuterium (by a factor of about 3), resulting in a small improvement of the He exhaust rate.

A key element in future toroidal magnetic fusion machines like ITER is the design of a divertor, which allows for safe particle and power exhaust in parallel with high bulk plasma performance. Correspondingly, the definition and design of an optimized divertor is a major task within the ongoing international ITER research and development effort. In order to provide a profound physics basis for such a divertor optimization, different divertor geometries are being tested on major tokamaks. This paper describes the effects of these divertor modifications on plasma performance in ASDEX Upgrade and in JET. In conclusion, increasing closure improves divertor performance without limiting the core plasma performance in ELMy H-modes.

In ASDEX Upgrade the excitation of the fishbone instability has been experimentally investigated. The initial frequency of the fishbone oscillation scales with the toroidal precession frequency of the fast trapped ions. A new feature, the coupling of the fishbone mode with the edge localized mode (ELM) instability during high-β discharges, is found. Fishbones occur in ASDEX Upgrade only for βfast = βtor/(1+τE /τsd) above a certain threshold. It is thus possible to calculate their destabilization regime in terms of global plasma parameters, such as plasma current, plasma density, toroidal magnetic field and central electron temperature.

Assembly of the superconducting stellarator Wendelstein 7-X is well advanced, and commissioning of the device is being prepared. A first draft of the commissioning tasks has been developed and will be discussed in this paper.

The influence of impurity radiation on the and transitions is investigated for highly radiative divertor discharges in the ASDEX Upgrade tokamak. The transitions between H- and L-mode depend on the net heat flow across the separatrix, , calculated from the heating power and bolometric radiation profiles. For typical radiative boundary conditions in ASDEX Upgrade, the radiation distribution is dominated by line emission in the vacuum-ultraviolet region and peaks near the separatrix. For the case of neon used as seed impurity, about 2/3 of the main chamber radiation is emitted inside, but close to the separatrix. Argon seed results in a higher fraction of core radiation, while the line emission is shifted further outside for nitrogen. The radiation-corrected threshold is not affected by gas puffing and is described by [MW, , T, amu]. The threshold power, which is typically lower by a factor of two without strong deuterium puffing, is increased by heavy gas puffing leading to . In the vicinity of the radiation-induced transition, a general alignment of H and L mode is observed with regard to global energy confinement time and edge density and temperature profiles. The transition itself exhibits a smooth evolution in time. Reduction of target plate power load down to about 10% of the total heating power is easily achieved by edge radiation in the CDH-mode for low conditions. However, this reduction is attributed mainly to radiation from inside the separatrix and is connected to relatively high values of the core . These results emphasize the importance of the development of more closed divertor concepts, allowing for higher divertor radiation levels in connection with lower core .

ITER or any other future fusion device will operate in H-mode to take advantage of the improved confinement. Stationary operation in H-mode, however, requires the occurrence of ELMs, and the resulting power load on the divertor target plates is not acceptable from the engineering point of view. Beyond this, the present divertor concepts rely on a cold and dense (possibly detached) divertor plasma. Large type-I ELMs, however, carry too large a power and possibly burn through such cold divertor plasmas. The completely detached H-mode (CDH-mode), recently established in ASDEX Upgrade shows detachment at both divertor target plates and small ELMs (so-called type-III ELMs), that do not burn through the divertor plasma. The extrapolation of this scenario to ITER, however, is an important question.

The radiation and transport characteristics of ASDEX Upgrade discharges with a neon-driven radiative mantle are modelled using a one-dimensional radial impurity transport code that has been coupled to a simple divertor model describing particle recycling and pumping. The code is well suited to describing the measured impurity line radiation, total, soft x-ray and bremsstrahlung radiation in regions of the plasma that are not dominated by two-dimensional effects. The recycling and pumping behaviour of neon as well as the bulk transport of neon for radiative boundary scenarios are discussed.

Divertor plasmas with strong external gas puffing in ASDEX Upgrade have shown very efficient impurity retention, increasing with the divertor neutral gas density. The experiments presented here use feedback-controlled gas puffs in discharges with different pumping speed to keep the divertor neutral gas flux density the same. This allows for the first time a decoupling of the divertor neutral gas flux density and the external gas flow. The resulting plasmas are almost identical and show identical impurity retention, clearly demonstrating the importance of the divertor neutral gas density over the externally induced flow. {copyright} {ital 1996 The American Physical Society.}

The next step in the Wendelstein stellarator line is the large superconducting device Wendelstein 7-X, currently under construction in Greifswald, Germany. Steady-state operation is an intrinsic feature of stellarators, and one key element of the Wendelstein 7-X mission is to demonstrate steady-state operation under plasma conditions relevant for a fusion power plant. Steady-state operation of a fusion device, on the one hand, requires the implementation of special technologies, giving rise to technical challenges during the design, fabrication and assembly of such a device. On the other hand, also the physics development of steady-state operation at high plasma performance poses a challenge and careful preparation. The electron cyclotron resonance heating system, diagnostics, experiment control and data acquisition are prepared for plasma operation lasting 30 min. This requires many new technological approaches for plasma heating and diagnostics as well as new concepts for experiment control and data acquisition.

The completely detached high confinement (CDH) regime established recently in high power ASDEX-Upgrade single null divertor discharges is described. The standard CDH mode scenario with feedback controlled external deuterium and neon puffing is analysed in detail with respect to its divertor detachment and confinement properties. Impurity transport and pumping effects and the density operation windows of the CDH mode are outlined. First results with other impurity radiators (N,Ar) are given. Finally the CDH mode operation space is discussed in terms of main chamber and scrape-off layer plus divertor radiation and possible consequences for future fusion experiments are outlined.

Neutron yield measurements are performed at the Julich Tokamak experiment TEXTOR with different detectors located around the Tokamak and on center axis above the shielding roof. Due to the large structure of the machine and the extended diagnostic equipment, both virgin and scattered neutrons are detected. To prepare a Bonner sphere of appropriate response, the shape of the degraded neutron spectrum was determined with a spherical ionization chamber in situ. These investigations have been enlarged under quasi‐real conditions at the accelerator facility of the Physikalisch‐Technische Bundesanstalt in Braunschweig with a set of Bonner spheres used as a spectrometer. For the calibration the plasma neutron source is simulated by two point sources, 238Pu/B and 252Cf. The sources are moved around the TEXTOR vessel along the position of the magnetic axis.

Simultaneous operation of lower hybrid current drive and pellet injection could be successfully achieved. With peripheral ablation of the pellets by suprathermal electrons, the same net inward flux of particles is found as with deep penetration of pellets into Ohmically heated plasmas. The density profile ne(r) peaks with the same increment of the peaking factor Qn=ne0/(ne) in both cases. The global energy confinement time rises with density, tau E approximately ne, in the combined operation.

There exist two classes of discharges on ASDEX which overcome the severe confinement deterioration at high densities or heating power. The first class is characterized by unusually peaked density profiles and the second one by exceptionally broad ones. These profiles and the concomitant confinement enhancement imply a change in both the particle and heat transport. Whereas the changes in particle transport are not fully understood, transport analyses point out that the improved heat transport in the first profile class can be explained by reduced ion contribution losses coming close to the neoclassical ones. The different results for the ion transport with flat and peaked density profiles are qualitatively consistent with that expected from eta i-modes. This conclusion is supported by momentum transport studies. The analyses, however, cannot yet explain the electron heat transport which is anomalously high in all regimes.

Feedback-controlled puffing of neon and deuterium has been applied to control the edge-localized-mode behavior and the target plate power deposition during high-power H -mode discharges in ASDEX Upgrade. A regime has been found in which more than 90% of the heating power is lost through radiation and divertor detachment occurs, without deterioration of the energy confinement. The plasma remains in the H mode, exhibiting small-amplitude, high-frequency ELM's, which do not penetrate to the target plates in the strike zone region.

Neutral beam (NI), ion cyclotron resonance (ICRH) and lower hybrid resonance (LHRH) heating on ASDEX are discussed with regard to their effect on plasma confinement. Comparison of NI and ICRH shows that the L and H-regimes are universal confinement modes of auxiliary-heated tokamak plasmas (i.e. independent of the heating method), and that the edge electron temperature (or a related parameter) dictates which mode prevails. In this connection it is noted that carbonization of the vessel walls impedes transition to the H-mode in the case of NI heating. Studies of energy confinement in the intermediate regime from ohmic to NI heating reveal a gradual transition from ohmic ( approximately ne) to neutral injection L-mode ( approximately I) scaling. At the same time a remarkable invariance of electron temperature profile shape with increasing heating power is observed. Changing the NI power deposition profiles from central to off-axis leaves gross energy confinement times unchanged while central confinement is substantially improved.

Both in ohmically and beam-heated L-mode discharges of ASDEX, the electron-temperature (T/sub e/) profile shape can be varied over a wide range by the choice of the safety factor q/sub a/. The power-deposition profile, on the contrary, has no effect on the T/sub e/-profile shape. In current-free WVII-A stellarator plasmas, no such invariance property is found. An independent constraint seems to fix the current distribution j(r) of the tokamak, which defines the conditions of electron heat transport.

The H mode in ion cyclotron-resonance-heated plasmas has been investigated with and without additional neutral beam injection. Ion cyclotron-resonance heating can cause the transition into a high-confinement regime (H mode) in combination with beam heating. The H mode, however, has also been realized: for the first time: with ion cyclotron-resonance heating alone in the D (H)-hydrogen minority scheme at an absorbed rf power of 1.1 MW.

Interaction of lower hybrid waves at 1.3 GHz with ions and electrons was studied in the density range 0.2-5*1013 cm-3 in the ASDEX tokamak. At high densities, ne>or approximately=4*1013 cm-3, fast ions with mainly perpendicular velocities are produced by the RF power at the plasma periphery. They are not well confined and do not lead to any bulk plasma heating. At lower densities, 2*1013

Discharges fuelled by stationary pellet injection (PI), gas puffing (GP) or a combination of the two methods are compared with respect to recycling in the divertor and particle confinement. Fuelling by PI yields much better global particle confinement than by GP. This has been found for both low and high recycling. In the low-recycling case this improvement is due to the deeper particle deposition for PI than for GP since the transport in the inner plasma is not reduced. For high recycling the improvement results from both the deeper deposition and a reduction in the transport. The best global particle confinement was found for phases with low or no GP. This, however, can be reached for short times only. Since with PI alone it is impossible to keep the recycling on a high level, GP is unavoidable for sustaining the favourable high-recycling condition.

First ICRF experiments on ASDEX have been performed at 67 MHz, corresponding to 2 Omega CH-heating of a hydrogen plasma at B0=2.2 T. Despite divertor operation ICRH is accompanied by a significant increase of impurity production which can drastically be reduced by means of wall carbonisation. RF power up to 2.3 MW is routinely coupled to the plasma for pulse lengths of up to 1 sec. The RF heating is found to depend strongly on plasma preheating. In combination with neutral beam injection the ICRF heating efficiency is even higher than the one of NI. Confinement degrades with ICRH to values in between NI-L-type and OH confinement.

Neutral particle fluxes onto the walls of ASDEX have been investigated using a time-of-flight (TOF) method. The energy distributions of the neutrals could be determined in the range of 10–1000 eV/amu. Ohmic divertor and limiter discharges with equal plasma currents and densities have been compared for He and D2. The He0 outflux at ∼2000 eV from He discharges is 110 of the corresponding D0 flux in D2 discharges. At lower energies this difference is much smaller. In all cases many more He neutrals were observed than was anticipated from the CX rate-coefficients for He2+. The impurity fluxes due to sputtering by the CX-neutrals show no significant difference for He and D2 discharges. For divertor discharges CX-sputtering can fully account for the Fe impurity content determined spectroscopically.

The sawtooth oscillations in tokamak discharges with Ohmic and neutral-beam heating could be suppressed when a large part of the plasma current was driven by lower-hybrid waves (I/sub HF//I/sub p/approx. =0.5). The stabilization is due to a flattening of the current profile j(r) and an increase of q(0) above 1. Higher central electron temperatures are obtained with neutral-beam heating if the sawteeth are stabilized. The increase in total energy content in this case was 30% higher than in the presence of sawteeth.

This letter reports on two experiments undertaken to evaluate the retention of gaseous and target produced impurities in the ASDEX divertor. The retention for gaseous impurities was determined by puffing Ar into the main chamber and simulating the time behaviour of the Ar XVI line intensity with a time dependent impurity transport code including a simple divertor model. During Ohmic heating a factor of 3 and 4.5 enhancement of impurity retention if found relative to the vacuum time constant (90 ms) of the divertor chamber, for ne = 2 × 1013 cm−3 and ne = 3.5 × 1013 cm−3, respectively, while a drastic breakdown of the retention occurs during high power NI heating. – To deduce the retention of impurities generated at the divertor plates, a segment (3.5%) of the plates was covered with copper, a metal previously not used in ASDEX. By measuring the Cu influx at the target plates and the line intensity of the Cu XX line (11.38 A) in the core plasma and by using the transport code, it is found that during NI heating (ne ≤ 2 × 1013 cm−3) Cu atoms originating from the target plates have a ≤ 3.5 times higher probability to penetrate into the core plasma than if they had when originating from the main chamber walls.

The stellarator device Wendelstein 7-X (W7-X) is now in the final stage of assembly. All the machine components have been built and in mid 2014 the commissioning activities should start. The first objective of W7-X is to prove the stellarator optimization principles, i.e., to reach at least the same confinement quality as a similar sized tokamak. The second objective is to demonstrate stable high-power steady-state operation. In the present paper, after a short description of the device, the ongoing activities to complete the construction will be summarized, focusing on the optimization of the layout of the torus hall and on the completion and assembly of the in-vessel components. The main lessons learned during the assembly will be presented as well as the application of the quality control, the handling of non-conformities and the design change procedures. The planning of the commissioning sequences to perform the integral tests of the major W7-X systems is discussed in order to prepare the W7-X device to start plasma operation.

Lower hybrid current drive can be applied in a tokamak in a number of different schemes, namely for start-up of the discharge, for plasma current ramp-up, for driving a stationary plasma current or for recharging the Ohmic heating transformer in a quasistationary tokamak operation. Recent experimental work related to these topics can be found in Refs. /1–10/. In this paper we report experiments performed in ASDEX.

During the second operation phase (OP1.2) of the most advanced stellarator Wendelstein 7-X (W7-X), some events of fast plasma current decay with a time constant of about 1 ms were observed (for both toroidal and diamagnetic plasma currents). The events were triggered by Electron Cyclotron Current Drive (ECCD) and were much faster than the assumptions considered during the design phase of W7-X components: 100 kA and 2.7 MA plasma current discharge with the time constants of 140 and 50 ms for toroidal and diamagnetic currents respectively. During the observed 1 ms discharge of maximum expected plasma currents, significant eddy currents are expected to be induced in some components with resulting electro-magnetic (EM) forces. In order to avoid unnecessary changes on as-built components, evaluation of the eddy currents / EM forces with high accuracy is required, if the conservative simplified estimation shows criticality. As a result, a sophisticated EM model with plasma currents, superconducting (sc) coils, plasma vessel (PV) and multiple ports is developed in ANSYS® using the new element type SOLID236 / 237 (edge-based formulation), which is much more accurate than the legacy elements SOLID97 (vector potential formulation). The specific features of using SOLID236 elements for eddy current analysis are studied and presented in this paper. This includes the application of boundary conditions on sector EM model, modelling of excitation currents, field and eddy current accuracy, skin effect, shielding effect, voltage in conducting components, etc. Most of the issues are studied in comparison with SOLID97 elements. Finally, the pros and cons of SOLID236 and SOLID97 elements are summarized, and the experiences during the W7-X EM analyses are presented.