G. A. Wurden

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...

Energetic electron measurements in the edge of a reversed-field pinch

The edge plasma of the ZT‐40M [Fusion Technol. 8, 1571 (1985)] reversed‐field pinch has been studied using a combination of three different plasma probes: a double‐swept Langmuir probe, an electrostatic energy analyzer, and a calorimeter–Langmuir probe. The edge plasma has been measured both with and without a movable graphite tile limiter present nearby in the plasma. Without a limiter a fast nonthermal tail of electrons (T≂350 eV) is detected in the edge plasma with nearly unidirectional flow along B and having a density between 2% and 10% of the cold edge plasma (T≂20 eV). The toroidal sense of this fast electron flow is against the force of the applied electric field. A large power flux along B is measured flowing in the same direction as the fast electrons and is apparently carried by the fast electrons. With the limiter present the fast electrons are still detected in the plasma, but are strongly attenuated in the shadow of the limiter. The measured scrape‐off lengths for both the fast electrons and...

A high density field reversed configuration (FRC) target for magnetized target fusion: First internal profile measurements of a high density FRC

Magnetized target fusion (MTF) is a potentially low cost path to fusion, intermediate in plasma regime between magnetic and inertial fusion energy. It requires compression of a magnetized target plasma and consequent heating to fusion relevant conditions inside a converging flux conserver. To demonstrate the physics basis for MTF, a field reversed configuration (FRC) target plasma has been chosen that will ultimately be compressed within an imploding metal liner. The required FRC will need large density, and this regime is being explored by the FRX–L (FRC-Liner) experiment. All theta pinch formed FRCs have some shock heating during formation, but FRX–L depends further on large ohmic heating from magnetic flux annihilation to heat the high density (2–5×1022 m−3), plasma to a temperature of Te+Ti≈500 eV. At the field null, anomalous resistivity is typically invoked to characterize the resistive like flux dissipation process. The first resistivity estimate for a high density collisional FRC is shown here. Th...

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.

Scaling of the power exhaust channel in Alcator C-Mod

Parametric dependences of the heat flux footprint on the outer divertor target plate are explored in EDA H-mode and ohmic L-mode plasmas over a wide range of parameters with attached plasma conditions. Heat flux profile shapes are found to be independent of toroidal field strength, independent of power flow along magnetic field lines and insensitive to x-point topology (single-null versus double-null). The magnitudes and widths closely follow that of the “upstream” pressure profile, which are correlated to plasma thermal energy content and plasma current. Heat flux decay lengths near the strike-point in H- and L-mode plasmas scale approximately with the inverse of plasma current, with a diminished dependence at high collisionality in L-mode. Consistent with previous studies, pressure gradients in the boundary scale with plasma current squared, holding the magnetohydrodynamic ballooning parameter approximately invariant at fixed collisionality—strong evidence that critical-gradient transport physics plays ...

Gas jet disruption mitigation studies on Alcator C-Mod and DIII-D

High-pressure noble gas jet injection is a mitigation technique which potentially satisfies the requirements of fast response time and reliability, without degrading subsequent discharges. Previously reported gas jet experiments on DIII-D showed good success at reducing deleterious disruption effects. In this paper, results of recent gas jet disruption mitigation experiments on Alcator C-Mod and DIII-D are reported. Jointly, these experiments have greatly improved the understanding of gas jet dynamics and the processes involved in mitigating disruption effects. In both machines, the sequence of events following gas injection is observed to be quite similar: the jet neutrals stop near the plasma edge, the edge temperature collapses and large MHD modes are quickly destabilized, mixing the hot plasma core with the edge impurity ions and radiating away the plasma thermal energy. High radiated power fractions are achieved, thus reducing the conducted heat loads to the chamber walls and divertor. A significant (2 × or more) reduction in halo current is also observed. Runaway electron generation is small or absent. These similar results in two quite different tokamaks are encouraging for the applicability of this disruption mitigation technique to ITER.

Reconnection scaling experiment: A new device for three-dimensional magnetic reconnection studies

The reconnection scaling experiment (RSX), a linear device for studying three-dimensional magnetic reconnection in both collisional and collisionless laboratory plasmas, has been constructed at Los Alamos National Laboratory. Advanced experimental features of the RSX that lead to scientific advantages include the use of simple technology (commercial plasma guns) to create plasma and current channels. Physics motivations, design and construction features of the RSX, are presented. Basic plasma parameters that characterize the RSX are shown together with preliminary measurements of visible light emission during the merging of two parallel current channels.