Max Planck Institute for Plasma Physics - Greifswald

Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 Nucl. Fusion 39 2137-2664), is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

Abstract The B2‐Eirene code package was developed to give better insight into the physics in the scrape‐off layer (SOL), which is defined as the region of open field‐lines intersecting walls. The SOL is characterised by the competition of parallel and perpendicular transport defining by this a 2D system. The description of the plasma‐wall interaction due to the existence of walls and atomic processes are necessary ingredients for an understanding of the scrape‐off layer. This paper concentrates on understanding the basic physics by combining the results of the code with experiments and analytical models or estimates. This work will mainly focus on divertor tokamaks, but most of the arguments and principles can be easily adapted also to other concepts like island divertors in stellarators or limiter devices. The paper presents the basic equations for the plasma transport and the basic models for the neutral transport. This defines the basic ingredients for the SOLPS (Scrape‐Off Layer Plasma Simulator) code package. A first level of understanding is approached for pure hydrogenic plasmas based both on simple models and simulations with B2‐Eirene neglecting drifts and currents. The influence of neutral transport on the different operation regimes is here the main topic. This will finish with time‐dependent phenomena for the pure plasma, so‐called Edge Localised Modes (ELMs). Then, the influence of impurities on the SOL plasma is discussed. For the understanding of impurity physics in the SOL one needs a rather complex combination of different aspects. The impurity production process has to be understood, then the effects of impurities in terms of radiation losses have to be included and finally impurity transport is necessary. This will be introduced with rising complexity starting with simple estimates, analysing then the detailed parallel force balance and the flow pattern of impurities. Using this, impurity compression and radiation instabilities will be studied. This part ends, combining all the elements introduced before, with specific, detailed results from different machines. Then, the effect of drifts and currents is introduced and their consequences presented. Finally, some work on deriving scaling laws for the anomalous turbulent transport based on automatic edge transport code fitting procedures will be described. (© 2006 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

The H-mode is a confinement mode of toroidal plasmas, which may make the goals of fusion possible—the development of a clean energy source at competitive electricity costs. The most challenging aspect of the H-mode physics is the sudden disappearance of the edge turbulence whereas its driving forces—the gradients—increase. As the physics behind the H-mode is subtle many features are not yet clarified. There is, however, substantial experimental and theoretical evidence that turbulent flows, which normally limit the confinement, are diminished by sheared poloidal flow residing at the plasma edge. There are many conceivable mechanisms giving rise to sheared flow. The most intriguing of these is that fluctuations themselves induce the flow, which acts back to its generating origin and annihilates the turbulence. This review concentrates mostly on the transition physics, describes one line of understanding the H-mode in more detail, recalls some of the older observations and summarizes the achievements in the H-mode for both tokamaks and stellarators.

Large Type-I edge localized modes (ELMs) are completely eliminated with small n = 3 resonant magnetic perturbations (RMP) in low average triangularity, (delta) over bar = 0.26, plasmas and in ITER similar shaped (ISS) plasmas, delta = 0.53, with ITER relevant collisionalities v(e)* <= 0.2. Significant differences in the RMP requirements and in the properties of the ELM suppressed plasmas are found when comparing the two triangularities. In ISS plasmas, the current required to suppress ELMs is approximately 25% higher than in low average triangularity plasmas. It is also found that the width of the resonant q(95) window required for ELM suppression is smaller in ISS plasmas than in low average triangularity plasmas. An analysis of the positions and widths of resonant magnetic islands across the pedestal region, in the absence of resonant field screening or a self-consistent plasma response, indicates that differences in the shape of the q profile may explain the need for higher RMP coil currents during ELM suppression in ISS plasmas. Changes in the pedestal profiles are compared for each plasma shape as well as with changes in the injected neutral beam power and the RMP amplitude. Implications of these results are discussed in terms of requirements for optimal ELM control coil designs and for establishing the physics basis needed in order to scale this approach to future burning plasma devices such as ITER.

In order to support the operation of ITER and the planned experimental programme an extensive set of plasma and first wall measurements will be required. The number and type of required measurements will be similar to those made on the present-day large tokamaks while the specification of the measurements-time and spatial resolutions, etc-will in some cases be more stringent. Many of the measurements will be used in the real time control of the plasma driving a requirement for very high reliability in the systems (diagnostics) that provide the measurements.The implementation of diagnostic systems on ITER is a substantial challenge. Because of the harsh environment (high levels of neutron and gamma fluxes, neutron heating, particle bombardment) diagnostic system selection and design has to cope with a range of phenomena not previously encountered in diagnostic design. Extensive design and R&D is needed to prepare the systems. In some cases the environmental difficulties are so severe that new diagnostic techniques are required.The starting point in the development of diagnostics for ITER is to define the measurement requirements and develop their justification. It is necessary to include all the plasma parameters needed to support the basic and advanced operation (including active control) of the device, machine protection and also those needed to support the physics programme. Once the requirements are defined, the appropriate (combination of) diagnostic techniques can be selected and their implementation onto the tokamak can be developed. The selected list of diagnostics is an important guideline for identifying dedicated research and development needs in the area of ITER diagnostics.This paper gives a comprehensive overview of recent progress in the field of ITER diagnostics with emphasis on the implementation issues. After a discussion of the measurement requirements for plasma parameters in ITER and their justifications, recent progress in the field of diagnostics to measure a selected set of plasma parameters is presented. The integration of the various diagnostic systems onto the ITER tokamak is described. Generic research and development in the field of irradiation effects on materials and environmental effects on first mirrors are briefly presented. The paper ends with an assessment of the measurement capability for ITER and a forward of what will be gained from operation of the various diagnostic systems on ITER in preparation for the machines that will follow ITER.

The theory of plasma confinement by non-axisymmetric magnetic fields is reviewed. Such fields are used to confine fusion plasmas in stellarators, where in contrast to tokamaks and reversed-field pinches the magnetic field generally does not possess any continuous symmetry. The discussion is focussed on magnetohydrodynamic equilibrium conditions, collisionless particle orbits, and the kinetic theory of equilbrium and transport. Each of these topics is fundamentally affected by the absence of symmetry in the magnetic field: the field lines need not trace out nested flux surfaces, the particle orbits may not be confined, and the cross-field transport can be very large. Nevertheless, by tailoring the magnetic field appropriately, well-behaved equilibria with good confinement can be constructed, potentially offering an attractive route to magnetic fusion. In this article, the mathematical apparatus to describe stellarator plasmas is developed from first principles and basic elements underlying confinement optimization are introduced.

This paper reviews measurements of edge plasma turbulence in toroidal magnetic fusion devices with an emphasis on recent results in tokamaks. The dominant feature of edge turbulence is a high level of broadband density fluctuations with a relative amplitude δn/n ∼ 5–100%, accompanied by large potential and electron temperature fluctuations. The frequency range of this turbulence is ∼10 kHz–1 MHz, and the size scale is typically ∼0.1–10 cm perpendicular to the magnetic field but many metres along the magnetic field, i.e. the structure is nearly that of 2D 'filaments'. Large intermittent bursts or 'blobs' are usually observed in the scrape-off layer. Diagnostic and data analysis techniques are reviewed and the main experimental results are summarized. Recent comparisons of experimental results with edge turbulence theory are discussed, and some directions for future experiments are suggested.

Abstract This paper is the first in a series of three papers to summarize the recent work of an European‐wide collaborationwhich is ongoing since about one decade using Particle‐in‐Cell (PIC) methods in low temperature plasma physics. In the present first paper the main aspects of this computational technique will be presented. In the second paper, an overview of applications in low‐temperature plasma modelling will be given, whereas the third part will put emphasis on the specific results of modelling ion thrusters. (© 2008 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Abstract High-energy particle accelerators have been crucial in providing a deeper understanding of fundamental particles and the forces that govern their interactions. To increase the energy of the particles or to reduce the size of the accelerator, new acceleration schemes need to be developed. Plasma wakefield acceleration 1–5 , in which the electrons in a plasma are excited, leading to strong electric fields (so called ‘wakefields’), is one such promising acceleration technique. Experiments have shown that an intense laser pulse 6–9 or electron bunch 10,11 traversing a plasma can drive electric fields of tens of gigavolts per metre and above—well beyond those achieved in conventional radio-frequency accelerators (about 0.1 gigavolt per metre). However, the low stored energy of laser pulses and electron bunches means that multiple acceleration stages are needed to reach very high particle energies 5,12 . The use of proton bunches is compelling because they have the potential to drive wakefields and to accelerate electrons to high energy in a single acceleration stage 13 . Long, thin proton bunches can be used because they undergo a process called self-modulation 14–16 , a particle–plasma interaction that splits the bunch longitudinally into a series of high-density microbunches, which then act resonantly to create large wakefields. The Advanced Wakefield (AWAKE) experiment at CERN 17–19 uses high-intensity proton bunches—in which each proton has an energy of 400 gigaelectronvolts, resulting in a total bunch energy of 19 kilojoules—to drive a wakefield in a ten-metre-long plasma. Electron bunches are then injected into this wakefield. Here we present measurements of electrons accelerated up to two gigaelectronvolts at the AWAKE experiment, in a demonstration of proton-driven plasma wakefield acceleration. Measurements were conducted under various plasma conditions and the acceleration was found to be consistent and reliable. The potential for this scheme to produce very high-energy electron bunches in a single accelerating stage 20 means that our results are an important step towards the development of future high-energy particle accelerators 21,22 .

International collaboration on development of a stellarator confinement database has progressed. More than 3000 data points from nine major stellarator experiments have been compiled. Robust dependences of the energy confinement time on the density and the heating power have been confirmed. Dependences on other operational parameters, i.e. the major and minor radii, magnetic field and the rotational transform , have been evaluated using inter-machine analyses. In order to express the energy confinement in a unified scaling law, systematic differences in each subgroup are quantified. An a posteriori approach using a confinement enhancement factor on ISS95 as a renormalizing configuration- dependent parameter yields a new scaling expression ISS04; . Gyro-Bohm characteristic similar to ISS95 has been confirmed for the extended database with a wider range of plasma parameters and magnetic configurations than in the study of ISS95. It has also been discovered that there is a systematic offset of energy confinement between magnetic configurations, and its measure correlates with the effective helical ripple of the external stellarator field. Full documentation of the International Stellarator Confinement Database is available at http://iscdb.nifs.ac.jp/ and http://www.ipp.mpg.de/ISS.

This review covers basic physics ingredients of the runaway phenomenon and the ongoing efforts (experimental and theoretical) aimed at runaway electron taming in the next generation tokamaks. We emphasize the prevailing physics themes of the last 20 years: the hot-tail mechanism of runaway production, runaway electron interaction with impurity ions, the role of synchrotron radiation in runaway kinetics, runaway electron transport in presence of magnetic fluctuations, micro-instabilities of runaway electrons in magnetized plasmas, and vertical stability of the plasma with runaway electrons. Here, the review also discusses the implications of the runaway phenomenon for ITER and the current strategy of runaway electron mitigation.

Abstract 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.

Abstract The paper presents an overview on the state of the art of 3D fluid transport modeling in the boundaries of 3D toroidal confinement devices and on applications to island divertor physics. Typically, such edge configurations are characterized by the coexistence of closed magnetic surfaces, islands and open stochastic regions, e.g. in helical devices like W7‐AS, W7‐X, LHD and in tokamaks like TEXTOR‐DED. Two main approach branches falling within the current numeric catalogue of the 3D modeling are the finite volume andMonte Carlo methods. They differ essentially in the elementary treatment of the local transport. While in a finite volume method interpolation of the fluid fluxes through the interfaces by appropriate choice of a shape function is essential for the discretization process, the full fluid dynamics are, in a Monte Carlo approach, simulated by means of a local stochastic process, with the fluxes passing through cell boundary surfaces being a net result of the random process. In this paper, we present the numerics and strategies proposed in different models. Concerning the practical applications to a realistic 3D experiment, W7‐AS provides not only a practical fully 3D island divertor configuration but also sufficient experimental data for code validation. We present the main simulation results from the 3D edge Monte Carlo code EMC3/EIRENE and discuss the island divertor physics with respect to tokamak divertors. (© 2004 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Wendelstein 7-AS was the first modular stellarator device to test some basic elements of stellarator optimization: a reduced Shafranov shift and improved stability properties resulted in β-values up to 3.4% (at 0.9 T). This operational limit was determined by power balance and impurity radiation without noticeable degradation of stability or a violent collapse. The partial reduction of neoclassical transport could be verified in agreement with calculations indicating the feasibility of the concept of drift optimization. A full neoclassical optimization, in particular a minimization of the bootstrap current was beyond the scope of this project. A variety of non-ohmic heating and current drive scenarios by ICRH, NBI and in particular, ECRH were tested and compared successfully with their theoretical predictions. Besides, new heating schemes of overdense plasmas were developed such as RF mode conversion heating—Ordinary mode, Extraordinary mode, Bernstein-wave (OXB) heating—or 2nd harmonic O-mode (O2) heating. The energy confinement was about a factor of 2 above ISS95 without degradation near operational boundaries. A number of improved confinement regimes such as core electron-root confinement with central Te ⩽ 7 keV and regimes with strongly sheared radial electric field at the plasma edge resulting in Ti ⩽ 1.7 keV were obtained. As the first non-tokamak device, W7-AS achieved the H-mode and moreover developed a high density H-mode regime (HDH) with strongly reduced impurity confinement that allowed quasi-steady-state operation (τ ≈ 65 · τE) at densities (at 2.5 T). The first island divertor was tested successfully and operated with stable partial detachment in agreement with numerical simulations. With these results W7-AS laid the physics background for operation of an optimized low-shear steady-state stellarator.

Compact optimized stellarators offer novel solutions for confining high-beta plasmas and developing magnetic confinement fusion. The three-dimensional plasma shape can be designed to enhance the magnetohydrodynamic (MHD) stability without feedback or nearby conducting structures and provide drift-orbit confinement similar to tokamaks. These configurations offer the possibility of combining the steady-state low-recirculating power, external control, and disruption resilience of previous stellarators with the low aspect ratio, high beta limit, and good confinement of advanced tokamaks. Quasi-axisymmetric equilibria have been developed for the proposed National Compact Stellarator Experiment (NCSX) with average aspect ratio 4-4.4 and average elongation similar to 1.8. Even with bootstrap-current consistent profiles, they are passively stable to the ballooning, kink, vertical, Mercier, and neoclassical-tearing modes for beta > 4%, without the need for external feedback or conducting walls. The bootstrap current generates only 1/4 of the magnetic rotational transform at beta = 4% (the rest is from the coils); thus the equilibrium is much less non-linear and is more controllable than similar advanced tokamaks. The enhanced stability is a result of 'reversed' global shear, the spatial distribution of local shear, and the large fraction of externally generated transform. Transport simulations show adequate fast-ion confinement and thermal neoclassical transport similar to equivalent tokamaks. Modular coils have been designed which reproduce the physics properties, provide good flux surfaces, and allow flexible variation of the plasma shape to control the predicted MHD stability and transport properties.

A complete system of transport equations with all the important perpendicular currents is derived for the simulation of tokamak edge plasma. These transport equations are implemented in the B2.5 code and solved for the parameters of the ASDEX Upgrade tokamak. The relative roles of different mechanisms of transverse conductivity in the formation of the potential profile are studied. It is demonstrated that a reasonable potential distribution in the tokamak edge plasma can be obtained without an ad hoc assumption of the existence of the anomalous perpendicular conductivity. The role of E × B drifts in the redistribution of edge plasma and closing of the currents in the plasma is analysed.

The Wendelstein 7X (W7-X) stellarator (R = 5.5 m, a = 0.55 m, B < 3.0 T), which at present is being built at Max-Planck-Institut für Plasmaphysik, Greifswald, aims at demonstrating the inherent steady-state capability of stellarators at reactor-relevant plasma parameters. A 10-MW electron cyclotron resonance heating (ECRH) plant with continuous-wave (cw) capability is under construction to meet the scientific objectives. The physics background of the different heating and current drive scenarios is presented. The expected plasma parameters are calculated for different transport assumptions. A newly developed ray-tracing code is used to calculate selected reference scenarios and optimize the electron cyclotron launcher and in-vessel structure. Examples are discussed, and the technological solutions for optimum wave coupling are presented. The ECRH plant consists of ten radio-frequency (rf) modules with 1 MW of power each at 140 GHz. The rf beams are transmitted to the W7-X torus (typically 60 m) via two open multibeam mirror lines with a power-handling capability, which would already satisfy the ITER requirements (24 MW). Integrated full-power, cw tests of two rf modules (gyrotrons and the related transmission line sections) are reported, and the key features of the gyrotron and transmission line technology are presented. As the physics and technology of ECRH for both W7-X and ITER have many similarities, test results from the W7-X ECRH may provide valuable input for the ITER-ECRH plant.

A series of different discharge configurations suitable for surface treatment at atmospheric pressure is discussed, including a non-thermal modular radio frequency (13.56, 27.12 or 40.78 MHz) jet plasma.

Documento escrito por un elevado número de autores/as, sólo se referencia el/la que aparece en primer lugar y los/as autores/as pertenecientes a la UC3M.

Interatomic or intermolecular Coulombic decay (ICD) is a nonlocal electronic decay mechanism occurring in weakly bound matter. In an ICD process, energy released by electronic relaxation of an excited atom or molecule leads to ionization of a neighboring one via Coulombic electron interactions. ICD has been predicted theoretically in the mid nineties of the last century, and its existence has been confirmed experimentally approximately ten years later. Since then, a number of fundamental and applied aspects have been studied in this quickly growing field of research. This review provides an introduction to ICD and draws the connection to related energy transfer and ionization processes. The theoretical approaches for the description of ICD as well as the experimental techniques developed and employed for its investigation are described. The existing body of literature on experimental and theoretical studies of ICD processes in different atomic and molecular systems is reviewed.