J.A. Boedo

Fast scanning probe for tokamak plasmas

We describe a fast reciprocating probe drive, which has three main new features: (1) a detachable and modular probe head for easy maintenance, (2) a combination of high heat flux capability, high bandwidth, and low-Z materials construction, and (3) low weight, compact, inexpensive construction. The probe is mounted in a fast pneumatic drive in order to reach plasma regions of interest and remain inserted long enough to obtain good statistics while minimizing the heat flux to the tips and head. The drive is pneumatic and has been designed to be compact and reliable to comply with space and maintenance requirements of tokamaks. The probe described here has five tips which obtain a full spectrum of plasma parameters: electron temperature profile Te(r), electron density profile ne(r), floating potential profile Vf(r), poloidal electric field profile Eθ(r), saturation current profile Isat(r), and their fluctuations up to 3 MHz. We describe the probe show radial profiles of various parameters. We compare the de...

Electric field-induced plasma convection in tokamak divertors

Measurements of the electric fields, E in the DIII-D tokamak divertor region [J. C. Luxon and L. G. Davis, Fusion Technology 8, Part 2A, 441 (1985)] are quantitatively consistent with recent computational modeling establishing that E×BT circulation is the main cause of changes in divertor plasmas with the direction of the toroidal magnetic field, BT. Extensive two-dimensional measurements of plasma potential in the DIII-D tokamak divertor region are reported for the first time. The resulting E×BT/B2 drift particle flux is calculated for standard (ion ∇BT drift toward divertor X-point) and reversed BT direction and for low (L) and high (H) confinement modes. Perpendicular field strengths of up to E∼5 kV/m are observed at the separatrix between the divertor private region and the scrape-off layer (SOL). The E×BT drift, which reverses with reversal of BT, creates a poloidal circulation pattern in the divertor that convects 25%–40% of the total ion flow to the divertor target. The circulation strongly couples...

IMPURITY-INDUCED TURBULENCE SUPPRESSION AND REDUCED TRANSPORT IN THE DIII-D TOKAMAK

Long wavelength turbulence as well as heat and momentum transport are significantly reduced in the DIII-D tokamak [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] as a result of neon seeding of a low confinement mode negative central shear discharge. Correspondingly, the energy confinement time increases by up to 80%. Fully saturated turbulence measurements near ρ=0.7 (ρ=r/a) in the wave number range 0.1⩽k⊥ρs⩽0.6, obtained with beam emission spectroscopy, exhibit a significant reduction of fluctuation power after neon injection. Fluctuation measurements obtained with far infrared scattering also show a reduction of turbulence in the core, while the Langmuir probe array measures reduced particle flux in the edge and scrape-off layer. Gyrokinetic linear stability simulations of these plasmas are qualitatively consistent, showing a reduction in the growth rate of ion temperature gradient driven modes for 0

MEASUREMENTS OF FLOWS IN THE DIII-D DIVERTOR BY MACH PROBES

First measurements of Mach number of background plasma in the DIII-D divertor are presented in conjunction with temperature T{sub e} and density n{sub e} using a fast scanning probe array. To validate the probe measurements, the authors compared the T{sub e}, n{sub e} and J{sub sat} data to Thomson scattering data and find good overall agreement in attached discharges and some discrepancy for T{sub e} and n{sub e} in detached discharges. The discrepancy is mostly due to the effect of large fluctuations present during detached plasmas on the probe characteristic; the particle flux is accurately measured in every case. A composite 2-D map of measured flows is presented for an ELMing H-mode discharge and they focus on some of the details. They have also documented the temperature, density and Mach number in the private flux region of the divertor and the vicinity of the X-point, which are important transition regions that have been little studied or modeled. Background parallel plasma flows and electric fields in the divertor region show a complex structure.

Radiative divertor plasmas with convection in DIII-D

The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features.

Turbulent transport and turbulence in radiative I?mode plasmas in TEXTOR-94

First measurements of turbulence levels and turbulence induced transport in the outer edge of the plasma of TEXTOR-94 during radiative improved mode discharges show a reduction by a factor of 4-7 of the radial particle turbulent transport. The quenching is most evident on the normalized potential fluctuations and is strongest above 100?kHz. Non-linear gyrokinetic particle-in-cell simulations of these discharges show impurity induced suppression of the electrostatic fluctuations associated with the ion temperature gradient driven mode over most of the cross-section, including the edge. Such a mechanism is proposed as the explanation for the improved confinement and turbulence reduction. The reduction in the edge turbulent transport levels is consistent with increased particle confinement time and the reduction of the SOL thickness. Particle and energy fluxes to the limiter are reduced by an order of magnitude. A concomitant increase of the measured energy and particle confinement times ?E and ?p versus radiated fraction suggests a common underlying suppression mechanism.

Self-consistent plasma-neutral modeling in tokamak plasmas with a large-area toroidal belt limiter

Plasma-neutral phenomena in the edge plasma and scrape-off layer of the Torus Experiment for Technology Oriented Research [G.H. Wolf and the TEXTOR Team, J. Nucl. Mater. 122&123, 1124 (1984)] with the toroidal belt Advanced Limiter Test (ALT-II) [D.M. Goebel et al., J. Nucl. Mater. 162–164, 115 (1989)] are simulated using the code package B2-EIRENE [D. Reiter et al., Plasma Phys. Controlled Fusion 33, 1579 (1991)]. Spatially-constant, anomalous radial transport coefficients (D,V,χ) are used for fitting measured electron temperature and density profiles. Primary neutral fluxes are determined by plasma fluxes to material surfaces, and Dα emissions are predicted from them. Comparison of the predicted Dα emission with measurements indicates a critical need, in predictive modeling, for a self-consistent model of fluxes to material surfaces that are parallel to the magnetic field. Appropriate factors are calculated for deducing D+ source rates from Dα emissions measured in various locations, taking into account...

Plasma Exhaust and Density Control in Tokamak Fusion Experiments with Neutral Beam or ICRF Auxiliary Heating

Particle exhaust studies have been carried out with the pump limiter ALT-II in the TEXTOR tokamak, under ohmic conditions as well as with NBI and with ICRF auxiliary heating, and the pumping effectiveness is shown to meet the requirements for a fusion reactor. Quantitative measurements of Dα emission, made with a CCD camera, have been used to determine the particle efflux from the plasma. Roughly one third of the Dα emission occurs in a diffuse `halo that surrounds the limiter belt. The particle confinement time is less than the energy confinement time by a factor of typically 4. Modelling in 2-D of plasma and neutral flows in the TEXTOR boundary has been performed. The source of D+ ions can be related to the Dα emission by a factor that is found to depend on the location of the emission and on the discharge density. The predicted total Dα emission agrees with the measurements within a factor of about 2. Pumping of ALT-II allows for density control; with NBI, the density can be increased well beyond the ohmic limit without the discharge ending in disruption. The plasma particle efflux and the pumped flux both increase with density as well as with heating power. The exhaust efficiency is typically ~2%, with the highest values observed in high density NBI discharges. Higher exhaust rates are observed with NBI than with ICRF. Plasma and neutral flows in the ALT-II scoops have been simulated, making use of a simple plasma model. The scoop may be viewed as a non-linear amplifier of the plasma particle flux; the amplification is found to range from about 2 to 3 for most cases. Flow reversal in the scoop is found in some of the NBI cases and particularly in the highest density case.

THE TOROIDAL PUMP LIMITER ALT-II IN TEXTOR

Abstract The Advanced Limiter Test (ALT) project is the focus of a fruitful and intense International Energy Agreement collaboration on TEXTOR. The pump limiter is a mechanical boundary that is laid out for taking the full heat load of TEXTOR, namely 8 MW (assuming 2 MW radiated power) for 10 s, and provides a pumping efficiency of at least 5% of the working gas. This layout is adopted from the requirements of a fusion reactor: It is mandatory to remove both the full power that is convected to the limiter or divertor and the helium ash that is generated in the fusion process. In order to obtain pumping for all gases, the ALT-II is equipped with turbomolecular pumps. A short description of ALT-II is given, and the power and particle fluxes to the limiter surface and into the exhaust scoops are discussed. Requirements of the helium removal rate for a reactor and relevant measurements are discussed, and particle removal and the power distribution to the limiters are treated. Related topics of the ALT-II program were hydrogen recycling and the measurement of turbulence-induced anomalous particle transport in the plasma edge.

Chapter 10: First Wall and Operational Diagnostics

Abstract In this chapter we review numerous diagnostics capable of measurements at or near the first wall, many of which contribute information useful for safe operation of a tokamak. Infrared cameras, visible and vacuum ultraviolet cameras, pressure gauges and residual gas analyzers, thermocouples, and erosion and deposition measurements by insertable probes, quartz microbalances, and a rather extensive review of Langmuir probes are discussed. Also discussed are dust measurements by electrostatic detectors, laser scattering, visible and infrared cameras, and manual collection of samples after machine opening. In each case the diagnostic is discussed with a view toward application to a burning plasma machine such as ITER.