Ricardo Jose Maqueda

Equilibrium properties of spherical torus plasmas in NSTX

Research in NSTX has been conducted to establish spherical torus plasmas to be used for high ?, auxiliary heated experiments. This device has a major radius R0 = 0.86?m and a midplane halfwidth of 0.7?m. It has been operated with toroidal magnetic field B0 ? 0.3?T and Ip ? 1.0?MA. The evolution of the plasma equilibrium is analysed between discharges with an automated version of the EFIT code. Limiter, double null and lower single null diverted configurations have been sustained for several energy confinement times. The plasma stored energy reached 92?kJ (?t = 17.8%) with neutral beam heating. A plasma elongation in the range 1.6 ? ? ? 2.0 and a triangularity in the range 0.25 ? ? ? 0.45 have been sustained, with values of ? = 2.6 and ? = 0.6 being reached transiently. The reconstructed magnetic signals are fitted to the corresponding measured values with low errors. Aspects of the plasma boundary, pressure and safety factor profiles are supported by measurements from non-magnetic diagnostics. Plasma densities have reached 0.8 and 1.2 times the Greenwald limit in deuterium and helium plasmas, respectively, with no clear limit encountered. Instabilities including sawteeth and reconnection events, characterized by Mirnov oscillations, and a perturbation of the Ip, ? and li evolutions, have been observed. A low q limit was observed and is imposed by a low toroidal mode number kink instability.

Gas puff imaging of edge turbulence (invited)

The gas puff imaging (GPI) diagnostic can be used to study the turbulence present at the edge of magnetically confined plasmas. In this diagnostic the instantaneous two-dimensional (2D) radial vs poloidal structure of the turbulence is measured using fast-gated cameras and discrete fast chords. By imaging a controlled neutral gas puff, of typically helium or deuterium, the brightness and contrast of the turbulent emission fluctuations are increased and the structure can be measured independently of natural gas recycling. In addition, recent advances in ultrafast framing cameras allow the turbulence to be followed in time. The gas puff itself does not perturb the edge turbulence and the neutral gas does not introduce fluctuations in the emission that could possibly arise from a nonsmooth (turbulent) neutral gas puff. Results from neutral transport and atomic physics simulations using the DEGAS 2 code are discussed showing that the observed line emission is sensitive to modulations in both the electron density and the electron temperature. The GPI diagnostic implementation in the National Spherical Torus Experiment (NSTX) and Alcator C-Mod tokamak is presented together with example results from these two experiments.

Observations concerning the injection of a lithium aerosol into the edge of TFTR discharges

A new method of actively modifying the plasma-wall interaction was tested on the Tokamak Fusion Test Reactor. A laser was used to introduce a directed lithium aerosol into the discharge scrape-off layer. The lithium introduced in this fashion ablated and migrated preferentially to the limiter contact points. This allowed the plasma-wall interaction to be influenced in situ and in real time by external means. Significant improvement in energy confinement and fusion neutron production rate as well as a reduction in the plasma Zeff have been documented in a neutral beam heated plasma. The introduction of a metallic aerosol into the plasma edge increased the internal inductance of the plasma column and also resulted in prompt heating of core electrons in ohmic plasmas. Preliminary evidence also suggests that the introduction of an aerosol leads to both edge poloidal velocity shear and edge electric field shear.

Edge turbulence measurements in NSTX by gas puff imaging

Turbulent filaments in visible light emission corresponding mainly to density fluctuations at the edge have been observed in large aspect ratio tokamaks: TFTR, ASDEX, Alcator C-Mod, and DIII-D. This article reports on similar turbulent structures observed in the National Spherical Torus Experiment (NSTX) using a fast-framing, intensified, digital visible camera. These filaments were previously detected mainly in high recycling regions, such as at limiters or antennas, where the line emission from neutral atoms was modulated by the fluctuations in local plasma density. However, by introducing controlled edge gas puffs, i.e., gas puff imaging, we have increased the brightness and contrast in the fluctuation images and allowed the turbulent structure to be measured independently of the recycling. A set discrete fiber-optically coupled sight-lines also measured the frequency spectra of these light fluctuations with a 200 kHz bandwidth. Initial results in NSTX show that the turbulent filaments are well aligned...

Initial results from coaxial helicity injection experiments in NSTX

Coaxial helicity injection has been investigated on the National Spherical Torus Experiment (NSTX). Initial experiments produced 130 kA of toroidal current without the use of the central solenoid. The corresponding injector current was 20 kA. Discharges with pulse lengths up to 130 ms have been produced.

FRX-L: A field-reversed configuration plasma injector for magnetized target fusion

We describe the experiment and technology leading to a target plasma for the magnetized target fusion research effort, an approach to fusion wherein a plasma with embedded magnetic fields is formed and subsequently adiabatically compressed to fusion conditions. The target plasmas under consideration, field-reversed configurations (FRCs), have the required closed-field-line topology and are translatable and compressible. Our goal is to form high-density (1017 cm−3) FRCs on the field-reversed experiment-liner (FRX-L) device, inside a 36 cm long, 6.2 cm radius theta coil, with 5 T peak magnetic field and an azimuthal electric field as high as 1 kV/cm. FRCs have been formed with an equilibrium density ne≈(1 to 2)×1016 cm−3, Te+Ti≈250 eV, and excluded flux ≈2 to 3 mWb.

Non-inductive current generation in NSTX using coaxial helicity injection

Coaxial helicity injection (CHI) on the National Spherical Torus Experiment (NSTX) has produced 240 kA of toroidal current without the use of the central solenoid. Values of the current multiplication ratio (CHI produced toroidal current/injector current) up to 10 were obtained, in agreement with predictions. The discharges, which lasted for up to 200 ms, limited only by the programmed waveform, are more than an order of magnitude longer in duration than any CHI discharges previously produced in a spheromak or a spherical torus.

β-Limiting MHD instabilities in improved-performance NSTX spherical torus plasmas

Global magnetohydrodynamic (MHD) stability limits in the National Spherical Torus Experiment (NSTX) have increased significantly recently due to a combination of device and operational improvements. First, more routine H-mode operation with broadened pressure profiles allows access to higher normalized β and lower internal inductance. Second, the correction of a poloidal field coil induced error-field has largely eliminated locked tearing modes during normal operation and increased the maximum achievable β. As a result of these improvements, peak β values have reached (not simultaneously) βT = 35%, βN = 6.4, βN = 4.5, βN/li = 10, and βP = 1.4. High βP operation with reduced tearing activity has allowed a doubling of discharge pulse-length to just over 1 s with sustained periods of βN≈6 above the ideal no-wall limit and near the with-wall limit. Details of the β-limit scalings and β-limiting instabilities in various operating regimes are described.

H-mode research in NSTX

H-modes are routinely obtained in the National Spherical Torus Experiment (NSTX) and have become a standard operational scenario. L–H transitions triggered by NBI heating have been obtained over a wide parameter range in Ip, Bt, and e in either lower-single-null (LSN) or double-null (DN) diverted discharges. Edge localized modes are observed in both configurations but the characteristics differ between DN and LSN, which also have different triangularities (δ). An H-mode duration of 500 ms was obtained in LSN, with a total pulse length of ~1 s. Preliminary power threshold studies indicate that the L–H threshold is between 600 kW and 1.2 MW, depending on the target parameters. Gas injector fuelling from the centre stack (i.e. the high toroidal field side) has enabled routine H-mode access, and comparisons with low-field side (LFS) fuelled H-mode discharges show that the LFS fuelling delays the L–H transition and alters the pre-transition plasma profiles. Gas puff imaging and reflectometry show that the H-mode edge is usually more quiescent than the L-mode edge. Divertor infrared camera measurements indicate up to 70% of available power flows to the divertor targets in quiescent H-mode discharges.

On the statistics of edge fluctuations: comparative study between various fusion devices

In this paper we present a statistical study of edge fluctuations taken with the gas puffing imaging (GPI) diagnostics. We carry out a comparison of GPI signal from an extensive database including four devices (two tokamaks and two reversed field pinches). The data are analysed in terms of their statistical moments Skewness and Kurtosis, as done in B Labit et al (2007 Phys. Rev. Lett. 98 255002). The data align along parabolic curves, although different from machine to machine, with some spread around the best-fitting curve. A discussion about the meaning of the parabolic trend as well as the departure of real data from it is provided. A phenomenological model is finally provided, attempting to accommodate experimental evidence.