The fast camera (Fastcam) imaging diagnostic systems on the DIII-D tokamak.

Two camera systems are installed on the DIII-D tokamak at the toroidal positions of 90° (90° system) and 225° (225° system), respectively. The cameras have two types of relay optics, namely, a coherent optical fiber bundle and a periscope system. The periscope system provides absolute intensity calibration stability while sacrificing resolution (10 lp/mm), while the fiber system provides high resolution (16 lp/mm) while sacrificing calibration stability. The periscope is available only for the 90° system. The optics of the 225° system were designed for view stability, repeatability, and easy maintenance. The cameras are located inside optimized neutron, x ray and magnetic shielding in order to reduce electronics damage, reboots, and magnetic and neutron interference, increasing the overall system reliability. An automated filter wheel, providing remote filter change, allows for remote wavelength selection. A software suite automates camera acquisition and data storage, allowing for remote operation and reduced operator involvement. System metadata is used to streamline the data analysis workflow, particularly for intensity calibration. The spatial calibration uses multiple observable wall features, resulting in a reconstruction accuracy ≤2 cm.

Accounting for Debye sheath expansion for proud Langmuir probes in magnetic confinement fusion plasmas.

A Child-Langmuir law-based method for accounting for Debye sheath expansion while fitting the current-voltage I-V characteristic of proud Langmuir probes (electrodes that extend into the volume of the plasma) is described. For Langmuir probes of a typical size used in tokamak plasmas, these new estimates of electron temperature and ion saturation current density values decreased by up to 60% compared to methods that did not account for sheath expansion. Changes to the collection area are modeled using the Child-Langmuir law and effective expansion perimeter lp, and the model is thus referred to as the "perimeter sheath expansion method." lp is determined solely from electrode geometry, so the method may be employed without prior measurement of the magnitude of the sheath expansion effects for a given Langmuir probe and can be used for electrodes of different geometries. This method correctly predicts the non-saturating ΔI/ΔV slope for cold, low-density plasmas where sheath-expansion effects are strong, as well as for hot plasmas where ΔI/ΔV ∼ 0, though it is shown that the sheath can still significantly affect the collection area in these hot conditions. The perimeter sheath expansion method has several advantages compared to methods where the non-saturating current is fitted: (1) It is more resilient to scatter in the I-V characteristics observed in turbulent plasmas. (2) It is able to separate the contributions to the ΔI/ΔV slope from sheath expansion to that of the high energy electron tail in high Te conditions. (3) It calculates the change in the collection area due to the Debye sheath for conditions where ΔI/ΔV ∼ 0 and for V = Vf.

Estimation of plasma ion saturation current and reduced tip arcing using Langmuir probe harmonics.

We present a method to calculate the ion saturation current, Isat, for Langmuir probes at high frequency (>100 kHz) using the harmonics technique and we compare that to a direct measurement of Isat. It is noted that the Isat estimation can be made directly by the ratio of harmonic amplitudes, without explicitly calculating Te. We also demonstrate that since the probe tips using the harmonic method are oscillating near the floating potential, drawing little power, this method reduces tip heating and arcing and allows plasma density measurements at a plasma power flux that would cause continuously biased tips to arc. A multi-probe array is used, with two spatially separated tips employing the harmonics technique and measuring the amplitude of at least two harmonics per tip. A third tip, located between the other two, measures the ion saturation current directly. We compare the measured and calculated ion saturation currents for a variety of plasma conditions and demonstrate the validity of the technique and its use in reducing arcs.

Main-Ion Intrinsic Toroidal Rotation Profile Driven by Residual Stress Torque from Ion Temperature Gradient Turbulence in the DIII-D Tokamak.

Intrinsic toroidal rotation of the deuterium main ions in the core of the DIII-D tokamak is observed to transition from flat to hollow, forming an off-axis peak, above a threshold level of direct electron heating. Nonlinear gyrokinetic simulations show that the residual stress associated with electrostatic ion temperature gradient turbulence possesses the correct radial location and stress structure to cause the observed hollow rotation profile. Residual stress momentum flux in the gyrokinetic simulations is balanced by turbulent momentum diffusion, with negligible contributions from turbulent pinch. The prediction of the velocity profile by integrating the momentum balance equation produces a rotation profile that qualitatively and quantitatively agrees with the measured main-ion profile, demonstrating that fluctuation-induced residual stress can drive the observed intrinsic velocity profile.

Experimental investigation of the role of fluid turbulent stresses and edge plasma flows for intrinsic rotation generation in DIII-D H-mode plasmas.

The first measurements of turbulent stresses and flows inside the separatrix of a tokamak H-mode plasma are reported, using a reciprocating multitip Langmuir probe at the DIII-D tokamak. A strong co-current rotation layer at the separatrix is found to precede intrinsic rotation development in the core. The measured fluid turbulent stresses transport toroidal momentum outward against the velocity gradient and thus try to sustain the edge layer. However, large kinetic stresses must exist to explain the net inward momentum transport leading to co-current core plasma rotation. The importance of such kinetic stresses is corroborated by the success of a simple orbit loss model, representing a purely kinetic mechanism, in the prediction of features of the edge corotation layer.

Fast scanning probe for the NSTX spherical tokamak.

We describe a fast reciprocating Langmuir probe and drive system, which has four main new features: (1) use of high-temperature, vacuum, circuit boards instead of cables to reduce weight and increase to 21 the number of possible connections, (2) rotatable and removable shaft, (3) 10 tip construction with designed hardware bandwidth up to 10 MHz, and (4) a detachable and modular tip assembly for easy maintenance. The probe is mounted in a fast pneumatic drive capable of speeds approximately 7 m/s and approximately 20g's acceleration in order to reach the scrape-off layer (SOL) and pedestal regions and remain inserted long enough to obtain good statistics while minimizing the heat deposition to the tips and head in a power density environment of 1-10 MW/m2. The National Spherical Torus Experiment SOL features electron temperature, T(e) approximately 10-30 eV, and electron density, n(e) approximately 0.1-5x10(12) cm(-3) while the pedestal features n(e) approximately 0.5-1.5x10(13) cm(-3) and T(e) approximately 30-150 eV. The probe described here has ten tips which obtain a wide spectrum of plasma parameters: electron temperature profile T(e)(r), electron density profile n(e)(r) and Mach number profile M(r), floating potential V(f)(r), poloidal and radial electric field profiles E(theta)(r) and E(rho)(r), saturation current profile I(sat)(r), and their fluctuations up to 3 MHz. We describe the probe and show representative radial profiles of various parameters.

Dust measurements in tokamaks (invited).

Dust production and accumulation present potential safety and operational issues for the ITER. Dust diagnostics can be divided into two groups: diagnostics of dust on surfaces and diagnostics of dust in plasma. Diagnostics from both groups are employed in contemporary tokamaks; new diagnostics suitable for ITER are also being developed and tested. Dust accumulation in ITER is likely to occur in hidden areas, e.g., between tiles and under divertor baffles. A novel electrostatic dust detector for monitoring dust in these regions has been developed and tested at PPPL. In the DIII-D tokamak dust diagnostics include Mie scattering from Nd:YAG lasers, visible imaging, and spectroscopy. Laser scattering is able to resolve particles between 0.16 and 1.6 microm in diameter; using these data the total dust content in the edge plasmas and trends in the dust production rates within this size range have been established. Individual dust particles are observed by visible imaging using fast framing cameras, detecting dust particles of a few microns in diameter and larger. Dust velocities and trajectories can be determined in two-dimension with a single camera or three-dimension using multiple cameras, but determination of particle size is challenging. In order to calibrate diagnostics and benchmark dust dynamics modeling, precharacterized carbon dust has been injected into the lower divertor of DIII-D. Injected dust is seen by cameras, and spectroscopic diagnostics observe an increase in carbon line (CI, CII, C(2) dimer) and thermal continuum emissions from the injected dust. The latter observation can be used in the design of novel dust survey diagnostics.

Suppression of large edge-localized modes in high-confinement DIII-D plasmas with a stochastic magnetic boundary.

A stochastic magnetic boundary, produced by an applied edge resonant magnetic perturbation, is used to suppress most large edge-localized modes (ELMs) in high confinement (H-mode) plasmas. The resulting H mode displays rapid, small oscillations with a bursty character modulated by a coherent 130 Hz envelope. The H mode transport barrier and core confinement are unaffected by the stochastic boundary, despite a threefold drop in the toroidal rotation. These results demonstrate that stochastic boundaries are compatible with H modes and may be attractive for ELM control in next-step fusion tokamaks.

Experimental evidence of intermittent convection in the edge of magnetic confinement devices.

Probe measurements in the PISCES linear device indicate the presence of plasma radially far from where it is produced. We show that this is mainly caused by large-scale structures of plasma with high radial velocity. Data from the Tore Supra tokamak show striking similarities in the shape of these intermittent events as well as the fluctuation density probability distribution and frequency spectrum. The fact that intermittent, large-scale events are so similar in linear devices and tokamaks indicates the universality of convective transport in magnetically confined plasmas.