Promising High-Confinement Regime for Steady-State Fusion.

A reproducible stationary high-confinement regime with small "edge-localized modes" (ELMs) has been achieved recently in the Experimental Advanced Superconducting Tokamak, which has a metal wall and low plasma rotation as projected for a fusion reactor. We have uncovered that this small ELM regime is enabled by a wide edge transport barrier (pedestal) with a low density gradient and a high density ratio between the pedestal foot and top. Nonlinear simulations reveal, for the first time, that the underlying mechanism for the observed small ELM crashes is the upper movement of the peeling boundary induced by an initial radially localized collapse in the pedestal, which stops the growth of instabilities and further collapse of the pedestal, thus providing a physics basis for mitigating ELMs in future steady-state fusion reactors.

A multi-species powder dropper for magnetic fusion applications.

We present a device for controlled injection of a variety of materials in powder form. The system implements four independent feeder units, arranged to share a single vertical drop tube. Each unit consists of a 80 ml reservoir, coupled to a horizontal linear trough, where a layer of powder is advanced by piezo-electric agitation at a speed proportional to the applied voltage, until it falls into a drop tube. The dropper has been tested with a number of impurities of low (B, BN, C), intermediate (Si, SiC), and high Z (Sn) and a variety of microscopic structures (flakes, spheres, rocks) and sizes (5-200 µm). For low Z materials, drop rates ∼2-200 mg/s have been obtained showing good repeatability and uniformity. A calibrated light-emitting diode (LED)-based flowmeter allows measuring and monitoring the drop rate during operation. The fast time-response of the four feeders allows combination of steady and pulsed injections, providing a flexible tool for controlled-dose, real-time impurity injection in fusion plasmas.

Upgraded flowing liquid lithium limiter for improving Li coverage uniformity and erosion resistance in EAST device.

We report on design and technology improvements for a flowing liquid lithium (FLiLi) limiter inserted into auxiliary heated discharges in the experimental advanced superconducting tokamak device. In order to enhance Li coverage uniformity and erosion resistance, a new liquid Li distributor with homogenous channels was implemented. In addition, two independent electromagnetic pumps and a new horizontal capillary structure contributed to an improvement in the observed Li flow uniformity (from 30% in the previous FLiLi design to >80% in this FLiLi design). To improve limiter surface erosion resistance, hot isostatic press technology was applied, which improved the thermal contact between thin stainless steel protective layers covering the Cu heat sink. The thickness of the stainless steel layer was increased from 0.1 mm to 0.5 mm, which also helped macroscopic erosion resilience. Despite the high auxiliary heating power up to 4.5 MW, no Li bursts were recorded from FLiLi, underscoring the improved performance of this new design.

New steady-state quiescent high-confinement plasma in an experimental advanced superconducting tokamak.

A critical challenge facing the basic long-pulse high-confinement operation scenario (H mode) for ITER is to control a magnetohydrodynamic (MHD) instability, known as the edge localized mode (ELM), which leads to cyclical high peak heat and particle fluxes at the plasma facing components. A breakthrough is made in the Experimental Advanced Superconducting Tokamak in achieving a new steady-state H mode without the presence of ELMs for a duration exceeding hundreds of energy confinement times, by using a novel technique of continuous real-time injection of a lithium (Li) aerosol into the edge plasma. The steady-state ELM-free H mode is accompanied by a strong edge coherent MHD mode (ECM) at a frequency of 35-40 kHz with a poloidal wavelength of 10.2 cm in the ion diamagnetic drift direction, providing continuous heat and particle exhaust, thus preventing the transient heat deposition on plasma facing components and impurity accumulation in the confined plasma. It is truly remarkable that Li injection appears to promote the growth of the ECM, owing to the increase in Li concentration and hence collisionality at the edge, as predicted by GYRO simulations. This new steady-state ELM-free H-mode regime, enabled by real-time Li injection, may open a new avenue for next-step fusion development.

2D divertor heat flux distribution using a 3D heat conduction solver in National Spherical Torus Experiment.

The divertor heat flux footprint in tokamaks is often observed to be non-axisymmetric due to intrinsic error fields, applied 3D magnetic fields or during transients such as edge localized modes. Typically, only 1D radial heat flux profiles are analyzed; however, analysis of the full 2D divertor measurements provides opportunities to study the asymmetric nature of the deposited heat flux. To accomplish this an improved 3D Fourier analysis method has been successfully applied in a heat conduction solver (TACO) to determine the 2D heat flux distribution at the lower divertor surface in the National Spherical Torus Experiment (NSTX) tokamak. This advance enables study of helical heat deposition onto the divertor. In order to account for heat transmission through poorly adhered surface layers on the divertor plate, a heat transmission coefficient, defined as the surface layer thermal conductivity divided by the thickness of the layer, was introduced to the solution of heat conduction equation. This coefficient is denoted as α and a range of values were tested in the model to ensure a reliable heat flux calculation until a specific value of α led to the constant total deposited energy in the numerical solution after the end of discharge. A comparison between 1D heat flux profiles from TACO and from a 2D heat flux calculation code, THEODOR, shows good agreement. Advantages of 2D heat flux distribution over the conventional 1D heat flux profile are also discussed, and examples of 2D data analysis in the study of striated heat deposition pattern as well as the toroidal degree of asymmetry of peak heat flux and heat flux width are demonstrated.

Spectral emission measurements of lithium on the lithium tokamak experiment.

There has been a long-standing collaboration between ORNL and PPPL on edge and boundary layer physics. As part of this collaboration, ORNL has a large role in the instrumentation and interpretation of edge physics in the lithium tokamak experiment (LTX). In particular, a charge exchange recombination spectroscopy (CHERS) diagnostic is being designed and undergoing staged testing on LTX. Here we present results of passively measured lithium emission at 5166.89 A in LTX in anticipation of active spectroscopy measurements, which will be enabled by the installation of a neutral beam in 2013. Preliminary measurements are made in transient LTX plasmas with plasma current, I(p) < 70 kA, ohmic heating power, P(oh) ∼ 0.3 MW and discharge lifetimes of 10-15 ms. Measurements are made with a short focal length spectrometer and optics similar to the CHERS diagnostics on NSTX [R. E. Bell, Rev. Sci. Instrum. 68(2), 1273-1280 (1997)]. These preliminary measurements suggest that even without the neutral beam for active spectroscopy, there is sufficient passive lithium emission to allow for line-of-sight profile measurements of ion temperature, T(i); toroidal velocity and v(t). Results show peak T(i) = 70 eV and peak v(t) = 45 km/s were reached 10 ms into the discharge.

High-harmonic fast-wave power flow along magnetic field lines in the scrape-off layer of NSTX.

A significant fraction of high-harmonic fast-wave (HHFW) power applied to NSTX can be lost to the scrape-off layer (SOL) and deposited in bright and hot spirals on the divertor rather than in the core plasma. We show that the HHFW power flows to these spirals along magnetic field lines passing through the SOL in front of the antenna, implying that the HHFW power couples across the entire width of the SOL rather than mostly at the antenna face. This result will help guide future efforts to understand and minimize these edge losses in order to maximize fast-wave heating and current drive.

A dual-band adaptor for infrared imaging.

A novel imaging adaptor providing the capability to extend a standard single-band infrared (IR) camera into a two-color or dual-band device has been developed for application to high-speed IR thermography on the National Spherical Tokamak Experiment (NSTX). Temperature measurement with two-band infrared imaging has the advantage of being mostly independent of surface emissivity, which may vary significantly in the liquid lithium divertor installed on NSTX as compared to that of an all-carbon first wall. In order to take advantage of the high-speed capability of the existing IR camera at NSTX (1.6-6.2 kHz frame rate), a commercial visible-range optical splitter was extensively modified to operate in the medium wavelength and long wavelength IR. This two-band IR adapter utilizes a dichroic beamsplitter, which reflects 4-6 µm wavelengths and transmits 7-10 µm wavelength radiation, each with >95% efficiency and projects each IR channel image side-by-side on the camera's detector. Cutoff filters are used in each IR channel, and ZnSe imaging optics and mirrors optimized for broadband IR use are incorporated into the design. In-situ and ex-situ temperature calibration and preliminary data of the NSTX divertor during plasma discharges are presented, with contrasting results for dual-band vs. single-band IR operation.

Continuous improvement of H-mode discharge performance with progressively increasing lithium coatings in the National Spherical Torus Experiment.

Lithium wall coatings have been shown to reduce recycling, improve energy confinement, and suppress edge localized modes in the National Spherical Torus Experiment. Here, we show that these effects depend continuously on the amount of predischarge lithium evaporation. We observed a nearly monotonic reduction in recycling, decrease in electron transport, and modification of the edge profiles and stability with increasing lithium. These correlations challenge basic expectations, given that even the smallest coatings exceeded that needed for a nominal thickness of the order of the implantation range.

Simulation of a tangential soft x-ray imaging system.

Tangentially viewing soft x-ray (SXR) cameras are capable of detecting nonaxisymmetric plasma structures in magnetically confined plasmas. They are particularly useful for studying stationary perturbations or phenomenon that occur on a timescale faster than the plasma rotation period. Tangential SXR camera diagnostics are planned for the DIII-D and NSTX tokamaks to elucidate the static edge magnetic structure during the application of 3D perturbations. To support the design of the proposed diagnostics, a synthetic diagnostic model was developed using the CHIANTI database to estimate the SXR emission. The model is shown to be in good agreement with the measurements from an existing tangential SXR camera diagnostic on NSTX.

2D soft x-ray system on DIII-D for imaging the magnetic topology in the pedestal region.

A new tangential two-dimensional soft x-ray imaging system (SXRIS) is being designed to examine the edge island structure in the lower X-point region of DIII-D. Plasma shielding and/or amplification of the calculated vacuum islands may play a role in the suppression of edge-localized modes via resonant magnetic perturbations (RMPs). The SXRIS is intended to improve the understanding of three-dimensional (3D) phenomena associated with RMPs. This system utilizes a tangential view with a pinhole imaging system and spectral filtering with beryllium foils. SXR emission is chosen to avoid line radiation and allows suitable signal at the top of a H-mode pedestal where T(e)∼1-2 keV. A synthetic diagnostic calculation based on 3D SXR emissivity estimates is used to help assess signal levels and resolution of the design. A signal-to-noise ratio of 10 at 1 cm resolution is expected for the perturbed signals, which are sufficient to resolve most of the predicted vacuum island sizes.

On demand triggering of edge localized instabilities using external nonaxisymmetric magnetic perturbations in toroidal plasmas.

The application of nonaxisymmetric magnetic fields is shown to destabilize edge-localized modes (ELMs) during otherwise ELM-free periods of discharges in the National Spherical Torus Experiment (NSTX). Profile analysis shows the applied fields increased the temperature and pressure gradients, decreasing edge stability. This robust effect was exploited for a new form of ELM control: the triggering of ELMs at will in high performance H mode plasmas enabled by lithium conditioning, yielding high time-averaged energy confinement with reduced core impurity density and radiated power.

Demonstration of Tokamak ohmic flux saving by transient coaxial helicity injection in the national spherical torus experiment.

Transient coaxial helicity injection (CHI) started discharges in the National Spherical Torus Experiment (NSTX) have attained peak currents up to 300 kA and when coupled to induction, it has produced up to 200 kA additional current over inductive-only operation. CHI in NSTX has shown to be energetically quite efficient, producing a plasma current of about 10 A/J of capacitor bank energy. In addition, for the first time, the CHI-produced toroidal current that couples to induction continues to increase with the energy supplied by the CHI power supply at otherwise similar values of the injector flux, indicating the potential for substantial current generation capability by CHI in NSTX and in future toroidal devices.

High speed infrared camera diagnostic for heat flux measurement in NSTX.

A new high speed infrared camera has been successfully implemented and produced first set of heat flux measurements on the lower divertor tiles in the NSTX tokamak. High spatial and temporal resolutions, 6.4 mm and 1.6-6.3 kHz, respectively, enable us to investigate detailed structure of heat flux deposition pattern caused by transient events such as edge localized modes. A comparison of the data with a slow infrared camera viewing the same region of interest shows good agreement between the two independent measurements. Data analysis for various plasma conditions is in progress.

Triggered confinement enhancement and pedestal expansion in high-confinement-mode discharges in the national spherical torus experiment.

We report observation of a new high performance regime in discharges in the National Spherical Torus Experiment, where the H mode edge "pedestal" temperature doubles and the energy confinement increases by 50%. The spontaneous transition is triggered by a large edge-localized mode, either natural or externally triggered by 3D fields. The transport barrier grows inward from the edge, with a doubling of both the pedestal pressure width and the spatial extent of steep radial electric field shear. The dynamics suggest that 3D fields could be applied to reduce edge transport in fusion devices.

Edge-localized-mode suppression through density-profile modification with lithium-wall coatings in the National Spherical Torus Experiment.

Reduction or elimination of edge localized modes (ELMs) while maintaining high confinement is essential for future fusion devices, e.g., the ITER. An ELM-free regime was recently obtained in the National Spherical Torus Experiment, following lithium (Li) evaporation onto the plasma-facing components. Edge stability calculations indicate that the pre-Li discharges were unstable to low-n peeling or ballooning modes, while broader pressure profiles stabilized the post-Li discharges. Normalized energy confinement increased by 50% post Li, with no sign of ELMs up to the global stability limit.

Enhanced energy confinement and performance in a low-recycling tokamak.

Extensive lithium wall coatings and liquid lithium plasma-limiting surfaces reduce recycling, with dramatic improvements in Ohmic plasma discharges in the Current Drive Experiment-Upgrade. Global energy confinement times increase by up to 6 times. These results exceed confinement scalings such as ITER98P(y,1) by 2-3 times, and represent the largest increase in energy confinement ever observed for an Ohmic tokamak plasma. Measurements of Dalpha emission indicate that global recycling coefficients decrease to approximately 0.3, the lowest documented for a magnetically confined hydrogen plasma.

Controlled optimization of mode conversion from electron Bernstein waves to extraordinary mode in magnetized plasma.

In the CDX-U spherical torus, agreement between radiation temperature and Thomson scattering electron temperature profiles indicates approximately 100% conversion of thermally emitted electron Bernstein waves to the X mode. This has been achieved by controlling the electron density scale length (L(n)) in the conversion region with a local limiter outside the last closed flux surface, shortening L(n) to the theoretically required value for optimal conversion. From symmetry of the conversion process, prospects for efficient coupling in heating and current drive scenarios are strongly supported.

Slow L-H transitions in DIII-D plasmas.

The transition from the low to the high mode of plasma confinement ( L-H transition) is studied in the DIII-D by an experimental technique which allows an arbitrarily slow transition. During an initial transition, periodic turbulent instability bursts are observed near the separatrix which inhibit the full transition. These bursts are damped by self-generated shear flows, and a predator-prey-type relationship is shown to give a good description of the data. As the neutral-beam power is raised, the oscillations change to type III edge localized modes. Another transition then leads to a quiet H mode.