S. Suckewer

Attainment of high confinement in neutral beam heated divertor discharges in the PDX tokamak

Abstract The PDX divertor configuration has recently been converted from an open to a closed geometry to inhibit the return of neutral gas from the divertor region to the main chamber. Since then, operation in a regime with high energy confinement in neutral beam heated discharges (ASDEX H-mode) has been routine over a wide range of operating conditions. These H-mode discharges are characterized by a sudden drop in divertor density and H α emission and a spontaneous rise in main chamber plasma density during neutral beam injection. The confinement time is found to scale nearly linearly with plasma current, but can be degraded due either to the presence of edge instabilities or heavy gas puffing. Detailed Thomson scattering temperature profiles show high values of T c near the plasma edge (∼ 450 eV) with sharp radial gradients (∼ 400 eV/cm) near the separatrix. Density profiles are broad and also exhibit steep gradients close to the separatrix.

Plasma rotation in the PDX tokamak

Toroidal and poloidal rotation has been measured in the Poloidal Divertor Experiment (PDX) tokamak in Ohmic and neutral-beam heated plasmas in a variety of discharge conditions and in both circular and diverted configurations. Rotation velocities were deduced from Doppler shifts of magnetic dipole (M1) lines and lines of optically allowed transitions in the visible and UV regions, from Kα emission, and also from an array of magnetic pickup loops. Poloidal and toroidal rotation velocities in Ohmically heated discharges were usually less than 3 × 105 cms−1. Near the plasma edge, the toroidal rotation velocity varies with poloidal angle both before and during neutral-beam injection. No systematic poloidal rotation was observed during neutral-beam injection centred about or displaced 10 cm from the horizontal midplane, which implies that the poloidal damping time τθ < 0.5 τii consistent with theoretical estimates. The central toroidal rotation velocity during neutral-beam injection scales linearly with the quantity and is independent of plasma current and toroidal magnetic field. The toroidal rotation velocity is higher in deuterium than in hydrogen plasmas, and also in diverted discharges as compared with circular ones. Toroidal rotation decay times after injection range from 80–100 ms at the centre to 160–180 ms at half the minor radius. Modelling of the radial profile of toroidal rotation indicates a central momentum diffusivity of the order of 8 × 103 cm2s−1. This is approximately a factor of three higher than the momentum diffusivity obtained from the decay time. All present theories are inadequate in accounting for the observed damping rate of v.

Toroidal plasma rotation in the PLT tokamak with neutral-beam injection

Toroidal plasma rotation in the Princeton Large Torus, PLT, has been measured for various plasma and neutral-beam injection conditions. Measurements of the plasma rotational velocities were made from Doppler shifts of appropriate spectra lines and include data from both hydrogen and deuterium beams and co- and counter-injection at several electron densities. Without injection, a small but consistent toroidal rotation exists in a direction opposite to the plasma current (counter-direction) in the plasma centre but parallel to the current (co-direction) in the plasma periphery. Using these velocities measured in the absence of injection, and the plasma density and temperature gradients, radial electron fields can be determined from theory, giving Er ≈ 40 V · cm−1 in the plasma centre and Er ≈ 10 V · cm−1 near the plasma edge. Insertion of a local, 2.5% magnetic well produced no observable effect on the beam-driven rotation. Modelling of the time evolution and radial distribution of the rotation allows one to deduce an effective momentum diffusivity of the order of (1–5) × 104 cm2 · s−1.

Low-Z impurity transport in tokamaks

Low-Z impurity transport in tokamaks was simulated with a one-dimensional impurity transport model including both neoclassical and anomalous transports. The neoclassical fluxes are due to collisions between the background plasma and impurity ions as well as to collisions between the various ionization states. The evaluation of the neoclassical fluxes takes into account the different collisionality regimes of the background plasma and the impurity ions. A limiter scrape-off model is used to define the boundary condition for the impurity ions in the plasma periphery. To account for the spectroscopic measurements of power radiated by the lower ionization states, fluxes due to anomalous transport are included. The sensitivities of the results to uncertainties in rate coefficients and plasma parameters in the periphery are investigated. The implications of the transport model for spectroscopic evaluation of impurity concentrations, impurity fluxes, and radiated power from line emission measurements are discussed.

Impurity levels and power loading in the pdx tokamak with high power neutral beam injection

Abstract The PDX tokamak provides an experimental facility for the direct comparison of various impurity control techniques under reactor-like conditions. Four neutral beam lines inject > 6 MW for 300 ms. Carbon rail limiter discharges have been used to test the effectiveness of perpendicular injection, but non-disruptive full power operation for > 100 ms is difficult without extensive conditioning. Initial tests of a toroidal bumper limiter indicate reduced power loading and roughly similar impurity levels compared to the carbon rail limiter discharges. Poloidal divertor discharges with up to 5 MW of injected power are cleaner than similar circular discharges, and the power is deposited in a remote divertor chamber. High density divertor operation indicates a reduction of impurity flow velocity in the divertor and enhanced recycling in the divertor region during neutral injection.

Vertical poloidal asymmetries of low-Z element radiation in the PDX tokamak

Vertical poloidal asymmetries of hydrogen isotopes and low-Z impurity radiation in the PDX tokamak may be caused by poloidally asymmetric sources of these elements at gas inlet valves, limiters or vacuum vessel walls, asymmetric magnetic-field geometry in the region beyond the plasma boundary, or by ion curvature drifts. Low ionization states of carbon (C II to C IV) are more easily affected by edge conditions than is C V. Vertical poloidal asymmetries of C V are correlated with the direction of the toroidal field. The magnitude of the asymmetry agrees with the predictions of a quasi-fluid neoclassical model. Experimental data and numerical simulations are presented to investigate different models of impurity poloidal asymmetries.

The effect of current profile evolution on plasma-limiter interaction and the energy confinement time

Experiments conducted on the PLT tokamak have shown that both plasma-limiter interaction and the gross energy confinement time are functions of the gas influx during the discharge. By suitably controlling the gas influx, it is possible to contract the current channel, decrease impurity radiation from the core of the discharge, and increase the gross energy confinement time, whether the aperture limiters are of tungsten, stainless steel or carbon.

Ion energy balance in Ohmically heated PLT discharges

Radial profiles of the ion temperature have been measured on PLT by using neutron flux Doppler broadening of oxygen, carbon and helium line radiations, and charge exchange. The ion temperature ( 1.2 keV) is consistent with Hinton-Rosenbluth neoclassical heat conduction in the plateau regime as the dominant energy loss from the plasma with central ion energy confinement times 0.1 s.

Plasma-materials interactions during RF experiments in tokamaks

Plasma-materials interactions studied in recent ICRF heating and lower hybrid current drive experiments are reviewed. The microscopic processes responsible for impurity generation are discussed. In ICRF experiments, improvements in machine operation and in antenna and feedthrough design have allowed efficient plasma heating at RF powers up to 3 MW. No significant loss of energy from the plasma core due to impurity radiation occurs. Lower hybrid current drive results in the generation and maintenance of hundreds of kiloamperes of plasma current carried by suprathermal electrons. The loss of these electrons and their role in impurity generation are assessed. Methods to avoid this problem are evaluated.

Radiation losses in PLT during neutral-beam and ICRF heating experiments

Radiation and charge-exchange losses in the PLT tokamak are compared for discharges with Ohmic heating only (OH), and with additional heating by neutral beams (NB) or RF in the ion cyclotron frequency range (ICRF). Spectroscopic, bolometric and soft-X-ray diagnostics were used. The effects of discharge cleaning, vacuum wall gettering, and rate of gas inlet on radiation losses from OH plasmas and the correlation between radiation from plasma core and edge temperatures are discussed. – For discharges with neutral-beam injection the radiation dependence on type of injection (e.g. co-injection versus counter- and co- plus counter-injection) was investigated. Radial profiles of radiation loss were compared with profiles of power deposition. Although total radiation was in the range of 30–60% of total input power into relatively clean plasma, nevertheless only 10–20% of the total central input power to ions and electrons was radiated from the plasma core. The radiated power was increased mainly by increased influx of impurities, however, a fraction of this radiation was due to the change in charge-state distribution associated with charge-exchange recombination. – During ICRF heating radiation losses were higher than or comparable to those experienced during co- plus counter-injection at similar power levels. At these low power levels of ICRF heating the total radiated power was ~ 80% of auxiliary-heating power. Radiation losses changed somewhat less rapidly than linearly with ICRF power input up to the maximum available at the time of these measurements (0.65 MW).