Flowing liquid lithium is a promising fusion technology because it can provide a renewable Plasma-Facing Component (PFC) surface, modify recycling, support power exhaust, and potentially connect plasma-facing components with fuel recovery. Its deployment, however, is limited by the need to manage chemical reactivity, fire and aerosol hazards, inert gas operation, maintainability, and rapid experimental iteration. This paper develops a semi-quantitative hazard complexity framework for selecting secondary containment architectures for flowing liquid lithium systems. The framework is applied to six representative containment scenarios and to the Lithium Experimental Application Platform (LEAP) at Princeton Plasma Physics Laboratory. LEAP is under construction with a modular, room-scale argon gloveroom as an inert secondary containment boundary for a staged flowing lithium program with heating, diagnostics, magnetic field exposure, and future device interface capability. The analysis shows that an inert, airtight secondary enclosure without scrubbers around a liquid lithium loop provides a practical balance between hazard reduction and facility complexity, as defined by the design requirements. The resulting architecture offers a deployable path for lithium PFC development and a transferable design logic for other reactive or conductive liquid metal systems.

LTX-beta, the upgraded Lithium Tokamak Experiment, recently began its first campaign with a goal to study the transport properties of gradient-free electron temperature profile equilibria with increased toroidal field and neutral beam injection. The temperature-gradient-free equilibria in LTX were enabled by the lithium plasma-facing surface. To make the lithium surface conditions more consistent for LTX-beta, a new, faster lithium evaporator system has been developed. The toroidal magnetic field has been nearly doubled to >= 0.3 T. Auxiliary heating and fueling with a high flux (35 A, 20 kV) neutral beam has begun. The beam provides core fueling without cold edge neutrals, to test whether flat temperature profiles can be sustained, and makes possible core ion temperature measurements via charge exchange recombination spectroscopy. Neutral beam fueling has been demonstrated with increases in line-integrated density. The coupling of the beam matches well with NUBEAM predictions for shine through. Improved performance has been observed following lithium evaporative coatings and an increase in field strength.

Lithium coatings in the Lithium Tokamak eXperiment (LTX) led to flat temperature profiles. The flat temperature profiles were observed along with a hot, low density edge, implying a broad, collisionless scrape-off layer (SOL). Additionally, in vacuo X-ray photoelectron spectroscopy (XPS) measurements established that lithium coatings evaporatively deposited onto high-Z plasma facing components (PFCs) became oxidized while retaining the ability to achieve good plasma performance long after lithium was applied to the PFCs. Longstanding theory predicted flat temperature profiles with low recycling walls, which was presumed to be due to hydrogen binding with elemental lithium to form lithium hydride. The presence of oxidized lithium, however, raised questions regarding the exact mechanism of hydrogen retention in LTX. To investigate these questions, the upgraded facility LTX- includes a new sample exposure probe (SEP) for more detailed in vacuo analysis of PFC samples. The SEP is equipped with a vacuum suitcase capable of transporting samples representative of the LTX-outer midplane PFCs to a stand-alone XPS system while maintaining pressures lower than the LTX-base vacuum to limit the contamination between sample exposure and analysis. The low-energy resolution XPS system used in past experiments could only enable the determination of elemental percentages on the PFC sample surfaces. Because the new XPS system has higher energy resolution, it is more direct to assign chemical compounds to the measured binding energies. This capability has been confirmed by comparing XPS data from PFC test samples with measurements using a commercial high-resolution XPS system. Quartz crystal microbalances (QCMs) were used to quantify the thickness of the deposited lithium on the LTX- PFCs. This article describes the application of the SEP to characterize the PFC surfaces using XPS and their relationship to plasma conditions.

The application of lithium to plasma facing components (PFCs) has been widely used to improve plasma performance in magnetic confinement devices. The ability of lithium to retain hydrogen isotopes has been exploited to lower wall recycling. Low recycling leads to high edge plasma temperatures, and decreased temperature gradients reduce the free energy source for microinstabilities that cause transport. A variety of techniques have been developed to introduce lithium into fusion research devices, and its beneficial effects have been observed in their plasmas. The mechanism for the retention of hydrogen isotopes has been elucidated by detailed surface science studies, and 'first principles' molecular dynamics simulations. The remaining challenges primarily pertain to long pulse operation. Our knowledge of erosion, transport, and redeposition needs to improve for the mixed material PFCs lithium creates. Flowing liquid lithium systems also must be demonstrated for the replenishable lithium walls that fusion reactors require.

We present a study of the role of boron and oxygen in the chemistry of deuterium retention in boronized ATJ graphite irradiated by the extreme environment of a tokamak deuterium plasma. The experimental results were obtained by the first XPS measurements inside the plasma chamber of the National Spherical Torus Experiment Upgrade, between the plasma exposures. The subtle interplay of boron, carbon, oxygen and deuterium chemistry is explained by reactive molecular dynamics simulations, verified by quantum-classical molecular dynamics and successfully compared to the measured data. The calculations deciphered the roles of oxygen and boron for the deuterium retention and predict deuterium uptake into a boronized carbon surface close in value to that previously predicted for a lithiated and oxidized carbon surface.

The importance of conditioning plasma-facing components (PFCs) has long been recognized as a critical element in obtaining high-performance plasmas in magnetic confinement devices. Lithium coatings, for example, have been used for decades for conditioning PFCs. Since the initial studies on the Tokamak Fusion Test Reactor (TFTR), experiments on devices with different aspect ratios and magnetic geometries like the National Spherical Torus Experiment (NSTX) continue to show the relationship between the lithium PFCs and good confinement and stability. While such results are promising, their empirical nature do not reflect the detailed relationship between the PFCs and the dynamic conditions that occur in the tokamak environment. A first step developing an understanding such complexity will be taken in the upgrade to NSTX, or the National Spherical Torus Experiment-Upgrade (NSTX-U) that is nearing completion. New measurement capabilities include the materials analysis and particle probe for in situ surface analysis of samples exposed to tokamak plasmas. The onion-skin modeling for edge analysis (OEDGE) suite of codes, for example, will be used to model the underlying mechanisms for such material migration in NSTX-U. This will lead to a better understanding of how plasma-facing surfaces evolve during a shot, and how the composition of the plasma-facing surface influences the discharge performance we observe. This paper will provide an overview of these capabilities, and highlight their importance for NSTX-U plans to transition from carbon to high-Z PFCs.

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ∼ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ∼ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment in the United States and Mega Ampere Spherical Tokamak in UK, active ST research has been conducted worldwide. More than 16 ST research facilities operating during this period have achieved remarkable advances in all fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.

The VUV/XUV spectrum has been measured on the Lithium Tokamak eXperiment (LTX) using a transmission grating imaging spectrometer (TGIS) coupled to a direct-detection x-ray charge-coupled device camera. TGIS data show significant changes in the ratios between the lithium and oxygen impurity line emission during discharges with varying lithium wall conditions. Lithium coatings that have been passivated by lengthy exposure to significant levels of impurities contribute to a large O/Li ratio measured during LTX plasma discharges. Furthermore, previous results have indicated that a passivated lithium film on the plasma facing components will function as a stronger impurity source when in the form of a hot liquid layer compared to a solid lithium layer. However, recent TGIS measurements of plasma discharges in LTX with hot stainless steel boundary shells and a fresh liquid lithium coating show lower O/Li impurity line ratios when compared to discharges with a solid lithium film on cool shells. These new measurements help elucidate the somewhat contradictory results of the effects of solid and liquid lithium on plasma confinement observed in previous experiments.

Dual-band infrared (IR) measurements of the National Spherical Torus eXperiment (NSTX) Liquid Lithium Divertor (LLD) are reported that demonstrate liquid Li is more effective at removing plasma heat flux than Li-conditioned graphite. Extended dwell of the outer strike point (OSP) on the LLD caused an incrementally larger area to be heated above the Li melting point through the discharge leading to enhanced D retention and plasma confinement. Measurement of Tsurface near the OSP demonstrates a significant reduction of the LLD surface temperature compared to that of Li-coated graphite at the same major radius. Modeling of these data with a 2-D simulation of the LLD structure in the DFLUX code suggests that the structure of the LLD was successful at handling up to q⊥,peak = 5 MW/m2 inter-ELM and up to 10 MW/m2 during ELMs from its plasma-facing surface as intended, and provide an innovative method for inferring the Li layer thickness.

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up to similar to 160g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement. H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium. 1300g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1], we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R&D required for use of lithium in future magnetic fusion facilities including ITER. (C) 2011 Elsevier B.V. All rights reserved.

Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The lithium tokamak experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the current drive experiment-upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in ohmically heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy (CHERS). Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions.

The liquid lithium divertor (LLD) to be installed in NSTX has four toroidal panels, each a conical section inclined at 22° like the previous graphite divertor tiles. Each LLD panel is a copper plate clad with ∼0.25mm of stainless steel (SS) and a surface layer of flame sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. LITER (evaporators) already used in NSTX will be upgraded for the LLD. Each has twelve 500W cartridge heaters with thermocouples, 16 other thermocouples, and a channel for helium cooling. During LLD experiments, the LLD will be heated so that the lithium is just above its melting temperature. The length of each shot will be preset to prevent excessive evaporation of lithium from the LLD. This duration depends on the heat load and is likely to be in the range of less than a second to several seconds. Careful thermal control of the LLD is important to maximize the shot times and to guide operation of the LLD. This paper describes the layout of the LLD, its expected thermal performance, the control system, and supporting experiments and analysis. A companion paper in this conference, “Physics design requirements for the national spherical torus experiment liquid lithium divertor,” provides other information.