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Edge localized mode (ELM) measurements in many tokamaks, including ASDEX-Upgrade, DIII-D, JET, JT-60U and MAST, are reviewed, which includes progress in experimental observations at the plasma edge region by means of fast-time resolved diagnostics with high precision, such as scanning probe, radial interferometer chord, BES and tangentially viewing fast-gated camera at the midplane. ELM dynamics data show that the majority of the ELM particle and energy transport should be dominated by ion convection physics and associated timescales. Furthermore, recent diagnostic upgrades on many tokamaks reveal the ELM filament structure and their complex motion towards radial, poloidal and toroidal directions. Approaches to control the Type-I ELMs, in addition to the alternative scenarios to Type-I ELMy H-mode operation (so-called, small/no ELM regimes) are also a key area of research for current tokamaks, which demonstrated a high confinement (being comparable to that of Type-I ELMy H-mode plasmas at similar parameters) in the absence of large, ELM induced, transient heat/particle fluxes to the divertor targets. Although tolerable ELM regimes are obtained in existing devices, their application to ITER is uncertain. Issues of these regimes towards further experiments and power deposition on divertor targets and main chamber wall are discussed.

The magnitude of plasma contact with main-wall surfaces is examined on the DIII-D poloidal divertor tokamak. A 'window-frame' technique has been developed for axisymmetric surfaces to provide measurements of total plasma flux (ions s−1) to the walls, Iwall. Despite the use of a separatrix-wall gap that is 2–3 times the radial e-folding length of the plasma parameters near the separatrix, increasing e-folding lengths away from the separatrix result in an Iwall of similar magnitude to the ion flux received by the divertor plate, Idiv. The Iwall/Idiv ratio increases strongly with line-averaged density and ranges from ~0.1–0.2 with attached outer divertor plasmas to ~1 with detached divertor plasmas. These observations hold during core density scans in both low (L-mode) and high (H-mode) confinement energy confinement regimes, and their importance to core fuelling and impurities is discussed. It is found that the magnitude of Iwall cannot be accurately measured by arbitrary main chamber D-α views due to the strong poloidal and toroidal asymmetry of the plasma contact. However D-α measurements reflect the relative trends of main-chamber recycling. Based on scrape-off layer (SOL) profiles in the shadow of the main-wall baffle, the far SOL cross-field particle transport is best described as convective with an effective velocity ~100 m s−1.

Edge density profile modifications associated with edge localized modes (ELMs) and edge harmonic oscillations (EHOs) have been measured using fast profile reflectometry in DIII-D H-mode plasmas. Specifically, high time (down to 10 µs) and spatial (~4 mm) resolution density profile measurements from 0 × 1019 to 3.1 × 1019 m−3 provide us with new tools to study the dynamics of pedestal perturbations. During ELMs, it has been observed that the SOL density profile expands outwards to the vessel wall while the pedestal density reduces. These measurements provide direct evidence for enhanced particle radial transport to the vessel wall at the onset of ELMs. A large SOL density profile radial expansion velocity is observed during ELMs. A comparison of the velocities for different pedestal densities is made. The density scale length, density at the wall, and density fluctuations during ELMs are determined as a function of pedestal density, as well as particle transport during ELMs. Measurements in quiescent H-mode plasmas indicate that the pedestal density profile is modulated at the fundamental frequency of the EHO. A comparison between reflectometer and beam emission spectroscopy measurements regarding the EHO density fluctuation profile is presented.

Research on the spherical torus (or spherical tokamak) (ST) is being pursued to explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MA-class National Spherical Torus Experiment (NSTX) at Princeton, and three medium sized ST research facilities: PEGASUS at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (?), non-inductive sustainment, Ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values ?T of up to 35% with a near unity central ?T have been obtained. NSTX will be exploring advanced regimes where ?T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta poloidal regime, where discharges with a high non-inductive fraction (~60% bootstrap current+NBI current drive) were sustained over the resistive skin time. Research on radio-frequency (RF) based heating and current drive utilizing high harmonic fast wave and electron Bernstein wave is also pursued on NSTX, PEGASUS, and CDX-U. For non-inductive start-up, the coaxial helicity injection, developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500?kA. In parallel, start-up using a RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.

This paper summarizes results from a two-dimensional (2D) physics analysis of the transition to and stable operation of the partially detached divertor (PDD) regime induced by deuterium injection in DIII-D. The analysis [1] shows that PDD operation is characterized by a radiation zone near the X-point at -15 eV which reduces the energy flux into the divertor and thereby also reduces the target plate heat flux, an ionization zone below the X-point which provides a deuterium ion source to fuel parallel flow down the outer divertor leg, an ion-neutral interaction zone in the outer leg which removes momentum and energy from the flow and finally a volume recombination zone above the target plate which reduces the particle flux to the low levels measured on the plates and thereby also contributes to reduction in target plate heat flux.

Edge transport and confinement changes induced by radial electric fields, externally imposed in TEXTOR-94 by means of electrode biasing, were investigated. The edge profiles of the electric field were measured continuously by a specially developed nine-tip Langmuir probe, thus allowing the study of the relative spatial and temporal dynamics of and its radial gradient and possible ensuing transport changes. A particle transport barrier is found to be built up as the electric field gradient increases, thus strengthening the conjecture that stabilization is a viable mechanism for improved confinement in tokamaks.

Measurements of edge profiles, turbulence, and turbulent-driven transport were made inside the last-closed flux surface (LCFS) and in the scrape-off layer (SOL) of PBX-M L-mode and H-mode plasmas using a fast reciprocating Langmuir probe diagnostic. Direct measurements of the potential profile confirm the generation of a strong inward radial electric field (Er approximately -100 V/cm) just inside the LCFS in H-mode. Density and potential fluctuation levels are reduced at the L-H transition, resulting in significantly lower turbulent transport. The edge density gradient then steepens in response to the reduced cross-field transport. After a stable H-mode edge equilibrium is formed, the fluctuation amplitudes recover back to nearly L-mode values. The poloidal wavenumber spectrum is narrowed in the transport barrier, and the density and potential fluctuations are decorrelated. The reduction in turbulent transport occurs across the LCFS and SOL regions and is not localized to the region of strong radial electric field.

Links between turbulent fluctuations and transport processes in tokamaks, particularly in TEXT, are discussed. In the plasma edge probes allow detailed measurements of the quantities required to directly determine the fluctuation driven fluxes. The total flux of particles is well explained by the measured electrostatic fluctuation driven flux. However, a satisfactory model to explain the origin of the fluctuations has not been identified. The processes responsible for determining the edge electron heat flux are less clear. In the confinement region experimental observations are restricted to measurements of density and potential fluctuations and their correlations. Three distinct features have been identified: (1) the ubiquitous broadband turbulence and (2) a quasi-coherent feature, both of which propagate in the electron diamagnetic drift direction, and (3) an ion feature which by definition propagates in the ion drift direction. The characteristics of the measured fluctuations are discussed and compared with the predictions of various models. Comparisons between measured particle, electron heat and ion heat fluxes, and those fluxes predicted to result from the measured fluctuations, are made.