Table of contents

The 24th European Physical Society Conference on Controlled Fusion and Plasma Physics was held in Berchtesgaden, Germany, 9 - 13 June, 1997. It was organized by the Max-Planck-Institut für Plasmaphysik, Garching, Germany, under the auspices of the Plasma Physics Division of the European Physical Society. Following the guidelines of the Board of the EPS Plasma Physics Division, the 1997 conference included topics from the areas of: tokamaks; stellarators; alternative magnetic confinement schemes; magnetic confinement theory and modelling; plasma edge physics; plasma heating; current drive and profile control; diagnostics; basic collisionless plasma physics; and highly compressed and non-stationary plasmas.

The conference was attended by 541 delegates from 29 countries. The programme of the conference included 8 review lectures, 20 topical lectures and 523 contributed papers (out of 651 submitted abstracts) for presentation as posters in four sessions.

The scientific programme and paper selection were the responsibility of the International Programme Committee appointed by the Board of the EPS Plasma Physics Division. Members of the Programme Committee were:

R Bartiromo (Chairman)

IGI del CNR, Padova, Italy

E Ascasibar

CIEMAT, Madrid, Spain

D Gresillon

LPMI/EP, Palaiseau, France

R D Hazeltine

Texas University, Austin, USA

F Hofmann

CRPP/EPFL, Lausanne, Switzerland

A Litvak

IAP, Nizhny Novgorod, Russia

L Stenflo

Umeå University, Umeå, Sweden

F Sluijter

Eindhoven University, Netherlands

D F H Start

JET Joint Undertaking, Abingdon, UK

U Samm

Forschungszentrum Jülich, Germany

F Serra

Universidade Technica, Lisboa, Portugal

T Todd

UKAEA Fusion, Abingdon, UK

F Wagner

IPP, Garching, Germany

R Weynants

ERM, Brussels, Belgium

The four-page contributed papers are published in the Europhysics Conference Abstracts series (volume 21A, parts I - IV) and distributed to all registered participants by the Local Organizing Committee. These papers are not reviewed. The review and topical Lectures are published in this special issue of Plasma Physics and Controlled Fusion which will also be sent to all conference delegates. These papers were reviewed by the members of the Programme Committee and are published from the final version presented by the authors. The Guest Editors are grateful to the Publishers for their co-operation.

M Schittenhelm, R Bartiromo and F Wagner Guest Editors

All major JET systems have been fully commissioned for D - T and the DTE1 series of experiments has started with the D - T fuel mixture and operating conditions foreseen for ITER. In the area of ITER physics, significant results have been produced in both D - D and D - T. In D-D, the L - H threshold power database has been extended, the bounds on edge-electron temperature and density in ELMy H-modes have been defined and the advantages of Types I and III ELMy discharges have been compared. In D - T plasmas, the isotope effect on H-mode threshold power and transport has been determined so that a more accurate assessment can be made of the ignition margin and heating requirements for ITER. Trace tritium experiments have provided first particle transport measurements and an assessment of the ITER reference ion-cyclotron resonance-frequency heating scenarios has been started. In the area of fusion performance, record D - D neutron yields have been obtained by controlling the plasma and current profiles in hot ion ELM-free H-modes and optimized shear modes. In D - T, internal transport barriers have been readily established in optimized shear discharges and Alfvén eigenmodes have been observed.

Densities achievable in ASDEX Upgrade discharges are restricted by a disruptive limit in the L-mode caused by an edge-power imbalance which is linking divertor detachment, Marfe formation and the separatrix density. The attainable average densities depend then on the internal particle sources and the core transport and can exceed the empirical Greenwald density.

In H-mode an upper density limit is found which represents a non-disruptive H - L back transition, which is preceded by the occurrence of type-III ELMs. Close to the Greenwald limit this H - L transition cannot be avoided at any power flux across the separatrix and - at high external neutral gas fluxes - confinement compared with ITER H-92P scaling degrades even before the back transition. The H-mode operational window is determined by local edge-barrier parameters and their gradients, respectively. The boundaries are represented by the L - H transition-temperature threshold, the ideal ballooning edge-pressure gradient limit, the upper temperature limit for type-III ELMs and an upper H-mode barrier density limitation. The cause for the last limitation is not yet identified; it may be due to resistive ballooning modes or the separatrix density limit.

Despite the limited edge densities the Greenwald density could be surpassed by a factor of three with pellet refuelling from the low magnetic-field side. Pellet injection from the high-field side gains from the strong increase of fuelling efficiency due to the assisting toroidal outward drift of the formed high- ablatant. Higher densities are achievable in H-mode compared with low-field side injection and diminished convective losses avoid confinement degradation up to the Greenwald density.

In gas-puffed type-I ELMy H-modes the plasma thermal energy and the edge-pressure gradients, which are limited by ballooning stability, are linked via a robust temperature-profile stiffness and the flat density profiles resulting from dominant edge refuelling at high densities. Their confinement does not improve with increasing density (and neutral gas fluxes) and may even slightly degrade. Therefore, the superior confinement of type-I ELMy H-modes compared with type-III ELMy ones at medium densities is actually offset at densities close to the Greenwald density.

In contrast to the temperature-profile resilience density profiles can be changed both by deep refuelling (with pellets) and intrinsic transport improvements connected with density peaking (observed in CDH-modes), which offers the combination of high confinement and high density operation. The possible alliance with radiation cooling, divertor detachment and divertor compatible type-III ELMs could solve the power exhaust problem.

The present status of inertial confinement fusion (ICF) is briefly reviewed, emphasizing the National Ignition Facility (NIF) project in the US and the Megajoule project in France. Critical aspects of target performance such as symmetry and stability of capsule implosions and interaction physics in hohlraum targets are discussed. The advantages of heavy-ion beam drivers and corresponding research programs are pointed out with reference to the long-term prospects for ICF power production. The new concept of the fast ignition of precompressed fuel by petawatt, picosecond laser pulses is also covered. The laser plasma group at the Max-Planck-Institute for Quantum Optics (MPQ) is one of the European institutes funded by EURATOM for an ICF keep-in-touch activity, and we highlight results obtained at MPQ relevant to the recent progress of ICF.

Significant reductions in the size and cost of a fusion power plant core can be realized if simultaneous improvements in the energy replacement time, , and the plasma pressure or beta, can be achieved in steady-state conditions with high self-driven, bootstrap current fraction. Significant recent progress has been made in experimentally achieving these high performance regimes and in developing a theoretical understanding of the underlying physics. Three operational scenarios have demonstrated potential for steady-state high performance, the radiative improved mode, the high internal inductance or high scenario, and the negative central magnetic shear, NCS (or reversed shear) scenario. In a large number of tokamaks, reduced ion thermal transport to near neoclassical values, and reduced particle transport have been observed in the region of negative or very low magnetic shear: the transport reduction is consistent with stabilization of microturbulence by sheared flow. There is strong temporal and spatial correlation between the increased sheared flow, the reduction in the measured turbulence, and the reduction in transport. The DIII-D tokamak, the JET tokamak and the JT-60U tokamak have all observed significant increases in plasma performance in the NCS operational regime. Strong plasma shaping and broad pressure profiles, provided by the H-mode edge, allow high beta operation, consistent with theoretical predictions; and normalized beta values up to simultaneously with confinement enhancement over L-mode scaling, , have been achieved in the DIII-D tokamak. In the JT-60U tokamak, deuterium discharges with negative central magnetic shear have reached equivalent breakeven conditions, .

High-performance experiments with the aim of establishing a physics basis for advanced steady-state tokamak reactors have been carried out in JT-60U using two approaches; high- H-mode and reversed-shear mode. In the high- H-mode, where an internal transport barrier (ITB) formed in the positive-shear region is combined with an edge-transport barrier (H-mode), a quasi-steady state with the ELMy H-mode edge has been obtained through pressure profile control and its beta limit has been improved by increasing the plasma triangularity, . In the reversed-shear mode, a radially localized ITB including a clear electron-temperature pedestal is formed in the negative-shear region and very high confinement is obtained; H factors up to 3.3 have been achieved with an L-mode edge. The location of the ITB was well correlated to the location of . Clear electron- and ion-temperature pedestals were sustained with a small density gradient in the combined heating experiments with ICRF+NBI. Large confinement improvement resulted from the large radius of the ITB and that of in the low region . The performance was limited by disruptive beta collapses with and and no steady-state was attained. The fusion performance was enhanced with the plasma current and the highest performance was achieved at 2.8 MA ; , and keV.

Optimization of both regimes will be continued, especially on the non-inductive current drive fraction and particle and heat control in the radiative divertor, using the negative-ion-based NBI and a newly installed W-shaped divertor.

This paper is a review of recent advances in numerical simulations of turbulence in fusion plasmas. Recent improvements of fluid and particle simulations are described. The results are compared with challenging issues such as the L - H transition and scaling laws. Finally, the predictive capability of turbulence simulations is addressed.

The Tokamak Fusion Test Reactor (TFTR) is a large tokamak which has performed experiments with 50:50 deuterium - tritium fuelled plasmas. Since 1993, TFTR has produced about 1090 D - T plasmas using about 100 grams of tritium and producing about 1.6 GJ of D - T fusion energy. These plasmas have significant populations of 3.5 MeV alphas (the charged D - T fusion product). TFTR research has focused on alpha particle confinement, alpha driven modes, and alpha heating studies. Maximum D - T fusion power production has aided these studies, requiring simultaneously operation at high input heating power and large energy confinement time (to produce the highest temperature and density), while maintaining low impurity content. The principal limitation to the TFTR fusion power production was the disruptive stability limit. Secondary limitations were the confinement time, and limiter power handling capability.

The fusion performance of ITER is predicted using three different techniques: statistical analysis of the global energy confinement data, a dimensionless physics parameter similarity method and the full one-dimensional modelling of the plasma profiles. Although the three methods give overlapping predictions for the performance of ITER, the confidence interval of all of the techniques is still quite wide.

Three-dimensional computer codes have been developed to simulate equilibrium, stability and transport in tokamaks and stellarators. Bifurcated solutions of the tokamak problem suggest that three-dimensional effects may be more important than has generally been thought. Extensive calculations have led to the discovery of a stellarator configuration with just two field periods and with aspect ratio 3.2 that has a magnetic field spectrum with toroidal symmetry. Numerical studies of equilibrium, stability and transport for this new device, called the Modular Helias-like Heliac 2 (MHH2), will be presented.

The TCV tokamak (, a < 0.25 m) has produced a wide variety of plasma configurations, both diverted and limited, with elongations ranging from 0.9 to 2.58, triangularities from -0.7 to 1 as well as discharges with nearly rectangular cross sections. Plasma currents of 1 MA have been obtained in elongated discharges . Ohmic discharges with have smaller sawteeth and higher levels of MHD mode activity than plasmas with . The main change in MHD behaviour when elongation is increased beyond two is an increase in the relative importance of modes with m,n < 1 and a reduction of sawtooth amplitudes. Confinement is strongly dependent on plasma shape. In ohmic limiter L-modes energy confinement times improve typically by a factor of two as the plasma triangularity is reduced from 0.5 to 0 at constant . There is also an improvement of confinement as the elongation is increased. In most discharges the changes in confinement are explained by a combination of geometrical effects and power degradation. A global factor of merit (shape enhancement factor) has been introduced to quantify the effect of flux surface geometry. The introduction of into well known confinement scaling expressions such as Neo-Alcator and Rebut - Lallia - Watkins scaling leads to improved descriptions of the effect of shape for a given confinement mode. In some cases with limited ohmic L-modes undergo a slow transition to a confinement regime with an energy confinement improved by a factor of up to 1.5 and higher particle confinement. First experiments to study the effect of shape in ECRH at a frequency of 83 GHz (second harmonic) have been undertaken with 500 kW of additional power.

This paper describes recent experimental investigations of the nonlinear dynamics of collisional current-driven drift waves in a linear low- discharge. It is shown that the bias of an injection grid leads to rigid-body rotation of the cylindrical plasma column that strongly destabilizes the drift waves, thus providing a control parameter for the drift-wave dynamics. In the nonlinear regime, when the control parameter is increased, the transition scenario from stability to weakly developed turbulence is studied. Two successive Hopf bifurcations, a mode-locked state and its gradual destabilization to chaos and finally turbulence follow the classical Ruelle - Takens transition scenario known from neutral fluids. In addition to the temporal dynamics, the spatiotemporal evolution of drift waves is studied by means of circular Langmuir probe arrays with high spatial and temporal resolution. With each Hopf bifurcation, a drift-mode onset is associated and the bifurcation from quasi-periodicity to mode locking corresponds to the transition from non-resonant to resonant mode interaction. The mode-locked state forms a persistent spatiotemporal pattern that is destabilized by the occurrence of defects. In contrast, the turbulent state is a fully disordered, intermittent state.

This paper presents the results of the latest analysis of disruption in tokamaks and the prediction of disruptions in ITER. The emphasis is on predictions of the halo current and its toroidal asymmetry in ITER, a study of runaway electron formation and an analysis of the concept for fast plasma shutdown by impurity injection. The concept of the plasma shutdown based on a massive injection of deuterium is also discussed.

Until a few years ago, most transient transport studies observed primarily diffusive plasma transport responses to fast, localized perturbations. Recently, several experiments have, in addition, observed `non-local' electron heat responses. Most remarkably, in `cold-pulse' experiments the abrupt edge cooling via radiative processes can induce both a diffusive cooling response moving in from the edge and, simultaneously, a rising electron temperature in the central core of tokamak plasmas - an opposite response even before the diffusive cooling from the edge reaches the centre! These and other non-local electron heat transport conundra from recent experiments are reviewed. The models and physical processes advanced to explain these puzzling phenomena are also discussed. The importance of resolving this transport enigma is emphasized.

The radiative improved mode obtained on the limiter tokamak TEXTOR-94 combines the possibility of power exhaust by a radiating plasma boundary (with a fraction of the radiated power with respect to the total input power up to 90% with neon or argon cooling) with improved energy confinement (as good as in the ELM-free H-mode in divertor tokamaks) at high plasma densities (line-averaged central-electron density equal to or even above the Greenwald density limit ) in quasi-stationary discharges. An overview is given of the substantial changes in plasma-edge properties occurring at high radiated power levels . These changes are characterized by a reduction of the plasma-edge density and temperature, a reduction of particle transport out of the confined plasma volume and an increase of the penetration depth of deuterium and impurity atoms. As a consequence, the particle confinement time increases and the electron-density profiles steepen. The transition to improved confinement takes place as soon as the density peaking reaches a critical threshold. An internal transport barrier is observed in the bulk of RI-mode plasmas (at ) characterized by an increase of the pressure gradient and of the shear of the toroidal velocity compared to discharges without additional impurity seeding. The dilution at the plasma boundary is strongly increased by the seeded impurities whereas the central dilution is only weakly affected.

The improved ergodic divertor, with a power handling capability of 9 MW over 30 s, opens the way to steady-state operation. Bootstrap current generation is investigated with fast wave electron heating and is achieved with . Lower hybrid current drive up to 75% of the total current of 1.3 MA has been achieved with MW. Detachment in this open divertor configuration occurs at 44% of the Greenwald density in ohmic shots. Target-plate plasma temperatures drop below 10 eV at detachment and then reach . The energy flux to the target plate measured by IR and Langmuir probe vanishes in this detached regime. Enhancement of radiation , with respect to the JET scaling, is achieved for some ED shots, . Self-healing of the confinement loss in the boundary layer is numerically investigated. In qualitative agreement with experimental results, an intrinsic transport barrier is found to build up in the vicinity of the separatrix.

The edge-region properties of a reverse-field pinch (RFP) configuration are reviewed and a comparison with those of tokamaks is attempted. Despite the different magnetic configurations, many similarities have been observed. Among the similarities is the electrostatic nature of the particle transport and the structure of the plasma potential. In particular the radial electric field changes sign across the last closed flux surface and gives rise to an velocity shear layer. The anomalous transport driven by electrostatic and magnetic fluctuations is addressed and the regimes of improved confinement recently observed in RFPs are reviewed. The role of the velocity shear on turbulence stabilization in these regimes is discussed.

Neoclassical tearing modes are shown to be the MHD phenomenon that limits the achievable in long pulse discharges both in ASDEX Upgrade and in COMPASS-D. The experimental data are consistent with a generalized Rutherford equation including the effect of polarization currents induced by the motion of the island through the plasma. Possible strategies to avoid or reduce the effect of neoclassical tearing modes are discussed.

Using additional heating provided by neutral-beam injection, the START spherical tokamak at UKAEA Fusion Culham has achieved high- (ratio of volume average plasma pressure to vacuum magnetic-field pressure) values of , more than twice the value previously obtained in a tokamak. These plasmas reach normalized beta values of at values of auxiliary heating power comparable to the ohmic power. Operation at high normalized current is observed, so that the plasma current exceeds the central rod toroidal-field current for the first time in a hot tokamak.

In a plasma interacting with ultra-short, high-intensity laser pulses, the magnetic part of the Lorentz force on the electrons can become as important as the electric part. In this case, we can expect the magnetic field to change the pattern of the nonlinear laser-pulse - plasma interaction drastically. The relativistic nonlinearities introduced by the magnetic interaction are of general interest in relation to the field of ultra-strong electromagnetic waves propagating in media, high-energy space plasmas and laser - plasma interaction under laboratory conditions. We present a summary of analytical and numerical results concerning the generation of quasi-static magnetic fields by high-intensity laser pulses in underdense plasmas and in thin plasma foils, and discuss the dynamical effects of these fields on the plasma motion and on the pulse propagation. This analysis indicates that phenomena such as current pinching, reconnection of magnetic-field lines and vortex propagation etc, that have been previously discussed in the case of space and laboratory plasmas, are also important for laser - plasma interactions.

Improved neoclassical electron confinement in the centre of low-density ECRH plasmas has been observed in the presence of a strong positive radial electric field, which resembles the electron root solution of the neoclassical ambipolarity condition but is obviously driven by the loss of ECRH-generated suprathermal electrons. At higher densities and with NBI heating, a high confinement regime substantially above the ISS95-scaling and different from the H-mode is established with a strongly sheared negative radial electric field at the boundary. The application of plasma-current induced magnetic shear reveals that confinement in W7-AS is essentially determined by perturbations at high-order rational surfaces. For optimum confinement, these resonances have either to be avoided in the boundary region or magnetic shear must be sufficiently large. Independent of its sign, magnetic shear can reduce electron energy transport which is enhanced in the presence of such resonances to the neoclassical level.

In tokamaks and stellarators, measurements of electromagnetic fluctuations in the presence of resonant particle drive, including fusion-produced , reveal the excitation of Alfvén eigenmodes (AE), related under certain conditions to a degradation in the fast-particle confinement. The balance between the drive and the background damping is investigated using active diagnostic systems to excite and measure the AE spectrum in terms of frequencies and damping rates. At JET, saddle-coil antennae drive low toroidal mode number (n<4) AE in the range 30 - 500 kHz, including toroidal AE, kinetic AE, elliptical AE and global AE. Conditions for weak damping are identified. Low-n AE appear to be strongly damped during the creation of the magnetic X-point. In the presence of resonant fast particles, information on the instability drive is obtained: low-n modes are found to be stable in the presence of NBI with . Fast ions generated by ICRH are observed to produce a drive for , with ; under these conditions, intrinsically driven TAE and EAE are clearly observed in the magnetic fluctuation spectra, with no measurable effect on the plasma performance.

Experiments with strong localized electron cyclotron heating (ECH) in the RTP tokamak show that electron heat transport is governed by alternating layers of good and bad thermal conduction. For central deposition hot filaments are observed inside the q = 1 radius. Moving the ECH resonance from the centre to the edge of the plasma results in discrete steps of the central electron temperature. The transitions occur when the minimum q value crosses q = 1,2,5/2 or 3, and correspond to the loss of a transport barrier situated close to the rational q value. Close to the transitions a new type of sawtooth activity is observed, characterized by the formation of sharp off-axis maxima on the profile, which collapse abruptly. The formation of the off-axis maxima is attributed to heat deposition precisely `on top of' a transport barrier.

Alfvén waves have long been known to be a major component of the turbulence measured in situ in the interplanetary medium. Until recently, however, observations had been limited to the ecliptic plane, where the solar wind structure is complicated by the interaction of fast and slow solar wind streams, the Alfvénic turbulence being essentially limited to high-speed streams in well defined magnetic sectors. The Ulysses spacecraft has shown how this structure disappears with increasing latitude, leading to a relatively constant high-speed stream originating from polar coronal holes. Within this region the radial magnetic field appears to be relatively constant with latitude, and the fluctuations are everywhere dominated by large-amplitude Alfvén waves propagating away from the sun, covering a broad band of wavelengths. Here we discuss the origin and evolution of solar wind Alfvén waves; the possible role played by such fluctuations in the heating of the corona and acceleration of the high-speed wind is explored in the light of both analytical models and numerical simulations.

The multi-channel HCN-polarimeter on TEXTOR-94 has been used to measure the plasma current density for a wide range of operation/confinement conditions. The characteristic features of the safety factor profiles in discharges with edge-radiation cooling and the application of different heating scenarios together with a strong impurity influx will be described. The stabilization of sawteeth and changes in the MHD behaviour of the plasma is related to modifications of the current density profile.

Collective Thomson scattering (CTS) of electromagnetic radiation from thermal plasma fluctuations in principle allows the velocity distribution of plasma ions and its composition in the plasma to be measured. The use of powerful microwave radiation from gyrotrons opens new perspectives for the application of CTS, which is considered to be a promising candidate for alpha-particle diagnostics in reactor-size tokamaks with D/T operation.

We have performed the first experiments at W7-AS with different scattering geometries to prove the applicability of gyrotrons for CTS. The experiments were performed with a 140 GHz gyrotron which is routinely used for ECRH, delivering a power of 0.45 MW. The receiver antenna and detection system for the registration of CTS spectra were especially designed for the scattering experiment. In backscattering experiments, which have inherently no spatial resolution, we have measured a transversely propagating, non-thermal lower-hybrid turbulence, which is driven by perpendicularly injected fast particles from a diagnostic neutral beam. The instability is excited by the beam ions under double-resonance conditions, where the LH frequency coincides with some harmonic of the beam ion gyrofrequency. For scattering geometries with the scattering wavevector not perpendicular to the magnetic field, thermal density fluctuations in the plasma were experimentally detected. The ion temperatures derived from these thermal spectra agree well with other diagnostics.

A modified scattering geometry ( scattering) allows local measurements of the ion temperature and is considered a prototype for the design of a routine diagnostic for ion-temperature measurements.

A record performance on JET has been obtained with shear optimization scenarios. A neutron yield of in deuterium discharges, and a global energy confinement improvement above the ITER-89 L-mode scaling with in L-mode and in H-mode have been achieved. The tailoring of plasma current, density and heating power waveforms and current profile control with lower hybrid current drive and ICRF phasing have been essential. Internal energy, particle and momentum transport barriers develop spontaneously upon heating above a threshold power of about 15 MW with neutral beams and ICRH into a low-density target plasma, with a wide central region of slightly negative or flat magnetic shear with q > 1 everywhere. An additional H-mode transition can also raise the pressure in the region between internal and edge transport barriers. The ion heat conductivity falls to the neoclassical level in the improved core confinement region. Pressure profile control through power deposition feedback control makes it possible to work close to the marginal stability boundary for pressure-driven MHD modes. First experiments in deuterium/tritium plasmas, with up to 75% tritium target concentration, have established internal transport barriers already with heating powers at the lowest threshold of pure deuterium plasmas, resulting in a fusion power output of .

This review addresses transitions into regimes with improved confinement. The paradigm is tested on the example of L - H transitions. It is demonstrated that L - H transitions may emerge either due to an amplification of the diamagnetic drift term caused by enhanced pressure gradient or due to an increased poloidal rotation velocity at the separatrix. In general, it is asserted that the emergence and dynamics of transitions are very sensitive to fine details of prelude plasma profiles. The L - H transition occurs provided the radial electric field changes dramatically within a few poloidal gyro-radii from the separatrix. The adoption of a significant power flux to the separatrix as the trigger of the L - H transition appears from the theoretical point of view to be a simplistic `brute force' solution to the issue of confinement control. The propagation of the front of the barrier is very fast. The dynamics of interest consists of threshold conditions for barrier formation, barrier propagation speed and profile steepening rates, barrier limits and mechanisms for their relaxation and termination.

The MHD activity of plasma configurations with reversed magnetic shear has been investigated on the FTU tokamak. In the presence of pairs of surfaces with the same rational value q = m/n of the safety factor, double-tearing modes are excited which give rise in most cases to bursts of sawtooth-like profile rearrangements. More stable regimes have also been found, in which the activity is dominated by rotating saturated modes. In a particular case with and a discharge without any detectable MHD activity during the current flat-top has been obtained. In high-temperature regimes ( at ), an irregular activity has been detected near the plasma centre which could be due to the excitation of resistive interchange modes.