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We describe a method of synthesizing a profile of the corrugated surface, which performs equalin-power reflection of E- and H-polarized waves in the case, where a phase difference of the reflection coefficients is a multiple of π. It is demonstrated that realization of such a regime requires a sufficient number of parameters in the profile description. Ambiguity of the obtained solution, i.e., the existence of different profiles with the required properties, is shown. Variants of such profiles are presented, and their diffraction characteristics are compared. The study is based on a rigorous numerical solution of the integral equation obtained by using a special interactive software.

A theory of a diplexer that can operate with an arbitrary polarization of the wave flow is constructed. A model experiment at a frequency of 34 GHz is carried out in good agreement with the theory. Diplexers of the given type are proposed to be used in experiments within the controlled fusion program.

Microwave heating of overdense low-temperature plasmas, whose density exceeds the cutoff density of the injected microwave, is described for the stellarator TJ-K. Three different methods featuring each of their own characteristics are briefly presented. It is shown that TJ-K allows both fusion relevant scenarios such as heating by electron Bernstein waves and fundamental wave-plasma interactions like parametric decay instabilities to be investigated.

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.

Seven gyrotrons are currently in operation in the ASDEX Upgrade Electron Cyclotron Resonance Heating (ECRH) system delivering a total of 3.9 MW to the plasma at 140 GHz. The 3 new gyrotrons are capable of 2-frequency operation and may heat the plasma alternatively with 2.1 MW at 105 GHz.

ASDEX Upgrade (AUG) is an ITER-shaped divertor tokamak with versatile heating, fueling, exhaust, and control systems. All plasma-facing components (PFCs) are coated with tungsten layers. Plasma scenarios have been adopted that avoid central tungsten accumulation, which can lead to an H-L transition due to excessive central radiative losses. Compared to a carbon-PFC tokamak, the AUG operational space is slightly more weighted toward higher densities and collisionalities. Actual and future planned extensions aim toward reducing the core collisionality while maintaining good power and particle exhaust. These extensions include a solid tungsten outer divertor target, improved pumping, higher ECRH power, and modified ICRF antennas that reduce tungsten sources. The newest element for advanced plasma control is the first set of eight magnetic perturbation coils, which already achieved type-I edge localized mode mitigation in multiple plasma scenarios. Another eight coils have been installed in autumn 2011, allowing the production of mode spectra with n >; 2. In parallel to the improved actuator set, an increasing number of diagnostics are brought into real-time state, allowing versatile profile and stability control.

A multi-frequency Electron Cyclotron Heating (ECRH) system is in operation at the ASDEX Upgrade tokamak, currently employing three depressed collector gyrotrons, operating at 105 and 140 GHz. The system offers a high flexibility for a variety of plasma heating and current drive experiments including remotely controlled fast beam steering and modulation.

ASDEX Upgrade is an ITER shaped divertor tokamak with versatile heating, fueling, exhaust and control systems. All plasma facing components are coated with tungsten layers. Plasma scenarios have been adopted to avoid central tungsten accumulation, which can lead to an H-L transition due to excessive central radiative losses. Compared to a carbon-PFC tokamak, the AUG operation space is slightly more weighted towards higher densities and collisionalities. Actual and future planned extensions aim towards reducing the core collisionality while maintaining good power and particle exhaust. These extensions include a solid tungsten outer divertor target, improved pumping, a higher ECRH power and modified ICRF antennas for reduced tungsten sources. The newest element for advanced plasma control is the first set of 8 magnetic perturbation coils, which already achieved type-I ELM mitigation in various plasma scenarios. Another 8 coils will be installed in autumn 2011 allowing to produce mode spectra with n > 2. In parallel to the improved actuator set, an increasing number of diagnostics is brought into real-time state, allowing versatile profile and stability control.

The multi-frequency Electron Cyclotron Heating (ECRH) system at the ASDEX Upgrade tokamak employs depressed collector gyrotrons, step-tunable in the range 105–140 GHz. The system is equipped with a fast steerable launcher allowing for remote steering of the ECRH RF beam during the plasma discharge. The gyrotrons and the mirrors are fully integrated in the discharge control system. The polarization can be controlled in a feed-forward mode. 3 Sniffer probes for millimeter wave stray radiation detection have been installed.

The ASDEX Upgrade programme is directed towards physics input to critical elements of the ITER design and the preparation of ITER operation, as well as addressing physics issues for a future DEMO design. After the finalization of the tungsten coating of the plasma facing components, the re-availability of all flywheel-generators allowed high-power operation with up to 20 MW heating power at I p up to 1.2 MA. Implementation of alternative ECRH schemes (140 GHz O2- and X3-mode) facilitated central heating above n e = 1.2 × 1020 m−3 and low q 95 operation at B t = 1.8 T. Central O2-mode heating was successfully used in high P/R discharges with 20 MW total heating power and divertor load control with nitrogen seeding. Improved energy confinement is obtained with nitrogen seeding both for type-I and type-III ELMy conditions. The main contributor is increased plasma temperature, no significant changes in the density profile have been observed. This behaviour may be explained by higher pedestal temperatures caused by ion dilution in combination with a pressure limited pedestal and hollow nitrogen profiles. Core particle transport simulations with gyrokinetic calculations have been benchmarked by dedicated discharges using variations of the ECRH deposition location. The reaction of normalized electron density gradients to variations of temperature gradients and the T e/T i ratio could be well reproduced. Doppler reflectometry studies at the L–H transition allowed the disentanglement of the interplay between the oscillatory geodesic acoustic modes, turbulent fluctuations and the mean equilibrium E × B flow in the edge negative E r well region just inside the separatrix. Improved pedestal diagnostics revealed also a refined picture of the pedestal transport in the fully developed H-mode type-I ELM cycle. Impurity ion transport turned out to be neoclassical in between ELMs. Electron and energy transport remain anomalous, but exhibit different recovery time scales after an ELM. After recovery of the pre-ELM profiles, strong fluctuations develop in the gradients of n e and T e. The occurrence of the next ELM cannot be explained by the local current diffusion time scale, since this turns out to be too short. Fast ion losses induced by shear Alfvén eigenmodes have been investigated by time-resolved energy and pitch angle measurements. This allowed the separation of the convective and diffusive loss mechanisms.

The multi-frequency Electron Cyclotron Heating (ECRH) system at the ASDEX Upgrade tokamak employs depressed collector gyrotrons, step-tunable in the range 105–140 GHz. The system is equipped with a fast steerable launcher allowing for remote steering of the ECRH beam during the plasma discharge. The polarization can be controlled in a feed-forward mode.

A new ECRH system is currently under construction at the ASDEX Upgrade tokamak. It employs for the first time depressed collector gyrotrons, step-tunable in the range 105–140 GHz. In its final stage it will consist of 4 gyrotrons with a total output power of 4 MW and a pulse length of 10 s. In this paper we describe recent extensions of the system and some experimental results.

In the torsatron TJ-K, low-temperature plasmas are created by means of microwaves at 2.45 GHz and 8.25 GHz. The plasmas are characterized as overdense which means that densities higher than the cut-off density are achieved. Studies in a wide parameter space show that the power deposition occurs predominantly at the upper-hybrid resonance. This can be seen in the electron-energy distribution function and the measured wave electric field. Full wave simulations support these measurements. An array antenna is presented which allows to scan the angle of incidence by variation of the microwave frequency.

The ASDEX Upgrade (AUG) ECRH system now delivers a total of 3.9 MW to the plasma at 140 GHz. Three new units are capable of 2-frequency operation and may heat the plasma alternatively with 2.1 MW at 105 GHz. The system is routinely used with X2, O2, and X3 schemes. For B t = 3.2 T also an ITER-like O1-scheme can be run using 105 GHz. The new launchers are capable of fast poloidal movements necessary for real-time control of the location of power deposition. Here real-time control of NTMs is summarized, which requires a fast analysis of massive data streams (ECE and Mirnov correlation) and extensive calculations (equilibria, ray-tracing). These were implemented at AUG using a modular concept of standardized real-time diagnostics. The new realtime capabilities have also been used during O2 heating to keep the first reflection of the non-absorbed beam fraction on the holographic reflector tile which ensures a well defined second pass of the beam through the central plasma. Sensors for the beam position are fast thermocouples at the edge of the reflector tile. The enhanced ECRH power was used for several physics studies related to the unique feature of pure electron heating without fueling and without momentum input. As an example the effect of the variation of the heating mix in moderately heated H-modes is demonstrated using the three available heating systems, i.e. ECRH, ICRH and NBI. Keeping the total input power constant, strong effects are seen on the rotation, but none on the pedestal parameters. Also global quantities as the stored energy are hardly modified. Still it is found that the central ion temperature drops as the ECRH fraction exceeds a certain threshold.