V. Soukhanovskii

The effect of lithium surface coatings on plasma performance in the National Spherical Torus Experiment

National Spherical Torus Experiment [which M. Ono et al., Nucl. Fusion 40, 557 (2000)] high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to plasma facing components found previously in limited plasmas can occur also in high-power diverted configurations. Lithium coatings were applied with pellets injected into helium discharges, and also with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium oven depositions from a few milligrams to 1g have been applied between discharges. Benefits from the lithium coatings were sometimes, but not always, seen. These benefits sometimes included decreases in plasma density, inductive flux consumption, and edge-localized mode occurrence, and increases in electron temperature, ion temperature, energy confinement, and periods of edge and magnetohydrodynamic quiescence. In addition, reductions in lower divertor D, C, and O luminosi...

Recent liquid lithium limiter experiments in CDX-U

Recent experiments in the Current Drive Experiment-Upgrade (CDX-U) provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall to gain engineering experience with a liquid metal first wall and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R = 34 cm, a = 22 cm, Btoroidal = 2 kG, IP = 100 kA, Te(0) ∼ 100 eV, ne(0) ∼ 5 × 10 19 m −3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium pool limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium pool limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5–8 increase in gas fuelling to achieve a comparable density, indicating that recycling is strongly reduced. Modelling of the discharges demonstrated that the lithium limited discharges are consistent with Zeffective < 1.2 (compared with 2.4 for the pre-lithium discharges), a broadened current channel and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced.

H-mode pedestal, ELM and power threshold studies in NSTX

H-mode operation plays a crucial role in National Spherical Torus Experiment (NSTX) research, allowing higher beta limits due to reduced plasma pressure peaking, and long pulse operation due to high bootstrap current fraction. Here, new results are presented in the areas of edge localized modes (ELMs), H-mode pedestal physics and power threshold studies. ELMs of several types as reported by higher aspect ratio tokamaks have been observed: (1) large, Type I ELMs, (2) intermediate-sized Type III ELMs and (3) tiny ELMs. Many high performance discharges in NSTX have the tiny ELMs (recently termed Type V), which have some differences as compared with small-magnitude ELM types in the published literature. A divertor multifaceted axisymmetric radiation from the edge (MARFE) on the inboard leg provides an effective light source to examine the effect of the ELMs on the divertor plasma; it is clear that only the large ELMs burn through the MARFE. The time difference between observation of the ELM flux at the outer and inner targets is substantially longer for the smallest ELMs as compared with the large ELMs. In addition, the visible light patterns show finger-like striations during the tiny ELMs. H-mode pedestal studies have commenced, with the observation that the pedestal contains between 25% and 33% of the total stored energy, and the NSTX pedestal energy agrees reasonably well with a recent international multi-machine scaling. A power threshold identity experiment between NSTX and the Mega-Amp Spherical Tokamak shows comparable loss power at the L?H transition in balanced double-null discharges. Both machines require more power for the L?H transition as the balance is shifted toward lower-single null. High-field side gas fuelling enables more reliable H-mode access in NSTX, but does not always lead to a lower power threshold, e.g. with a reduction of the duration of early heating.

Solenoid-free plasma startup in NSTX using transient CHI

Experiments in NSTX have now demonstrated the coupling of toroidal plasmas produced by the technique of coaxial helicity injection (CHI) to inductive sustainment and ramp-up of the toroidal plasma current. In these discharges, the central Ohmic transformer was used to apply an inductive loop voltage to discharges with a toroidal current of about 100 kA created by CHI. The coupled discharges have ramped up to > 700 kA and transitioned into an H-mode demonstrating compatibility of this startup method with conventional operation. The electron temperature in the coupled discharges reached over 800 eV and the resulting plasma had low inductance, which is preferred for long-pulse high-performance discharges. These results from NSTX in combination with the previously obtained record 160 kA non-inductively generated startup currents in an ST or tokamak in NSTX demonstrate that CHI is a viable solenoid-free plasma startup method for future STs and tokamaks.

Demonstration of 300 kA CHI-startup current, coupling to transformer drive and flux savings on NSTX

Discharges formed by transient coaxial helicity injection (CHI) in the National Spherical Torus Experiment (NSTX) have attained peak currents of 300 kA for the first time. CHI-started discharges are coupled to induction, and ramped up to over 1 MA. Up to an additional 400 kA of toroidal current is produced, compared with discharges with the same inductive drive without CHI. These CHI-inductively coupled discharges demonstrate flux savings over standard NSTX inductive-only discharges, requiring significantly less transformer flux to reach 1 MA of toroidal current, as well as exhibiting higher elongation and lower internal inductance. These results indicate the potential for substantial current generation capability by CHI in NSTX and in future toroidal devices.

The snowflake divertor

The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. Among potential beneficial effects of this geometry are: increased volume of a low poloidal field around the null, increased connection length, and the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation to the existing theoretical models is described.

Spectroscopic imaging diagnostics for burning plasma experiments

Spectroscopic imaging of plasma emission profiles from a few electron volts to tens of kilo-electron volts enables basic diagnostics in present day tokamaks. For the more difficult burning plasma conditions, light extraction and detection techniques, as well as instrument designs need to be investigated. As an alternative to light extraction with reflective optics, we discuss normal incidence, transmissive-diffractive optics (e.g., transmission gratings), which might withstand plasma exposure with less degradation of optical properties. Metallic multilayer reflectors are also of interest for light extraction. Although a shift of the diffraction peak might occur, instrument designs that accommodate such shifts are possible. As imaging detectors we consider “optical” arrays based on conversion of the short-wavelength light into visible light followed by transport of the visible signal with hollow lightguides. The proposed approaches to light extraction and detection could enable radiation resistant diagnostics.

Overview of the physics and engineering design of NSTX upgrade

The spherical tokamak (ST) is a leading candidate for a fusion nuclear science facility (FNSF) due to its compact size and modular configuration. The National Spherical Torus eXperiment (NSTX) is a MA-class ST facility in the U.S. actively developing the physics basis for an ST-based FNSF. In plasma transport research, ST experiments exhibit a strong (nearly inverse) scaling of normalized confinement with collisionality, and if this trend holds at low collisionality, high fusion neutron fluences could be achievable in very compact ST devices. A major motivation for the NSTX Upgrade (NSTX-U) is to span the next factor of 3–6 reduction in collisionality. To achieve this collisionality reduction with equilibrated profiles, NSTX-U will double the toroidal field, plasma current, and NBI heating power and increase the pulse length from 1–1.5s to 5s. In the area of stability and advanced scenarios, plasmas with higher aspect ratio and elongation, high βN, and broad current profiles approaching those of an ST-based FNSF have been produced in NSTX using active control of the plasma β and advanced resistive wall mode control. High non-inductive current fractions of 70% have been sustained for many current diffusion times, and the more tangential injection of the 2nd NBI of the Upgrade is projected to increase the NBI current drive by up to a factor of 2 and support 100% non-inductive operation. More tangential NBI injection is also projected to provide non-solenoidal current ramp-up (from IP = 0.4MA up to 0.8–1MA) as needed for an ST-based FNSF. In boundary physics, NSTX and higher-A tokamaks measure an inverse relationship between the scrape-off layer heat-flux width and plasma current that could unfavorably impact next-step devices. Recently, NSTX has successfully demonstrated very high flux expansion and substantial heat-flux reduction using a snowflake divertor configuration, and this type of divertor is incorporated in the NSTX-U design. The physics and engineering design supporting NSTX Upgrade are described.

Progress towards steady state on NSTX

In order to reduce recirculating power fraction to acceptable levels, the spherical torus concept relies on the simultaneous achievement of high toroidal β and high bootstrap fraction in steady state. In the last year, as a result of plasma control system improvements, the achievable plasma elongation on NSTX has been raised from K ∼ 2.1 to K ∼ 2.6-approximately a 25% increase. This increase in elongation has led to a substantial increase in the toroidal β for long pulse discharges. The increase in β is associated with an increase in plasma current at nearly fixed poloidal β, which enables higher β t with nearly constant bootstrap fraction. As a result, for the first time in a spherical torus, a discharge with a plasma current of 1 MA has been sustained for 1 s (0.8 s current flat-top). Data are presented from NSTX correlating the increase in performance with increased plasma shaping capability. In addition to improved shaping, H-modes induced during the current ramp phase of the plasma discharge have been used to reduce flux consumption and to delay the onset of MHD instabilities. Based on these results, a modelled integrated scenario, which has 100% non-inductive current drive with very high toroidal β, will also be discussed. The NSTX poloidal field coils are currently being modified to produce the plasma shape which is required for this scenario, which requires high triangularity (δ ∼ 0.8) at elevated elongation (K ∼ 2.5). The other main requirement of steady state on NSTX is the ability to drive a fraction of the total plasma current with RF waves. The results of high harmonic fast wave heating and current drive studies as well as electron Bernstein wave emission studies will be presented.

A toroidal liquid lithium limiter for CDX-U

Attention has focused recently on flowing liquid lithium as a first wall for a reactor because of its potentially attractive physics and engineering features. In order to test the suitability of liquid lithium as a plasma facing component, the Current Drive eXperiment - Upgrade (CDX-U) at the Princeton Plasma Physics Laboratory has recently installed a fully toroidal liquid lithium limiter. CDX-U is a compact (R = 34 cm, a = 22 cm, B/sub toroidal/ = 2 kG, I/sub p/ =100 kA, T/sub e/(O) /spl sim/ 100 eV, n/sub e/(0) /spl sim/ 5 /spl times/ 10/sup 19/ m/sup -3/ short-pulse (< 25 msec) spherical torus (ST) with extensive diagnostics. The limiter, which consists of a shallow circular stainless steel tray of radius 34 cm and width 10 cm, is filled with lithium to a depth of a few millimeters, and forms the lower limiting surface for the discharge. Heating elements beneath the tray are used to liquefy the lithium (melting point = 180.5/spl deg/C) prior to the experiment. The total area of liquid lithium exposed to the plasma is approximately 2000 cm/sup 2/. The design of the limiter, modifications to CDX-U to accommodate in-vessel inventories of approximately 1 liter of liquid lithium, techniques for loading lithium onto the limiter, and other preparations will be described. CDX-U has previously been successfully operated with a smaller area cm/sup 2/) liquid lithium rail limiter. Diagnostics specific to lithium operations include multichord spectrometry of the 135 /spl Aring/ LiIII line in the core plasma, monitors for neutral lithium light at the lithium limiter, and a fast (10,000 frame per second) camera which monitors motion of the liquid during the discharge. First results of plasma operations with the toroidal liquid lithium limiter will also be given.