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Type I edge-localized modes (ELMs) in the Experimental Advanced Superconducting Tokamak (EAST) were completely suppressed via boron powder injection into the X-point region of an upper-single null configuration over a wide range of operating conditions (2.8

The efficiency of two lithium (Li) injection methods used on the National Spherical Torus Experiment (NSTX) are compared in terms of the amount of Li used to produce equivalent plasma performance improvements, namely, Li evaporation over the divertor plates, prior to the initiation of the discharge, and real-time Li injection directly into the plasma scrape-off layer during the discharge. The measurements show that the real-time method can affect the energy confinement and the edge stability of NSTX plasmas in a more efficient way than the Li evaporation method, as it requires only a fraction of the amount of Li used by the evaporation method to produce similar improvements.

Real-time lithium powder injection has been applied to long-pulse (>30 s) H-mode plasmas in EAST. This replenishes the active lithium surface that is routinely consumed by plasma-wall interactions. The real-time injection of Li powder into long H-mode discharges effectively suppresses impurity influx and controls recycling on EAST, with an ITER-like tungsten upper divertor. With lithium powder injection, the concentrations W, Mo, and C were reduced by 50% compared to ELMy H-mode discharges. During lithium injection, two effects play a role in the suppression of impurities influx: a reduced divertor temperature and heat flux and hence reduced erosion, and impurity trapping via deposition of a Li film onto plasma-facing surfaces. The ‘fresh’ injected lithium replenishes the film deposited during daily morning evaporation, restoring the wall's pumping capability. Thus, a measurable reduction in the global recycling coefficient was observed. Keyword: Long pulse plasma, Lithium, Impurity, Recycling

The coating of tokamak walls with thin layers of lithium has been demonstrated to reduce plasma recycling from the plasma-facing surfaces and to improve overall plasma performance. These effects, including reduced divertor $D_{\alpha }$ emission, the elimination of edge-localized modes, and increased energy confinement have been observed in multiple experiments when lithium coatings are applied before plasma discharges. However, this coating technology does not extrapolate to future long-pulse devices, since the lithium coatings will be passivated by the continual plasma flux onto the surface. In order to provide active conditioning capability, a new technology has been developed that is capable of injecting lithium powder into the scrape-off layer plasma during plasma discharges, where it quickly liquefies and turns into an aerosol. The use of this “lithium dropper” is under study at the Experimental Advanced Superconducting Tokamak (EAST), where the potential benefits of real-time wall conditioning via lithium injection are being tested. Here, we present an analysis of the recycling characteristics during EAST experiments testing active lithium injection in order to assess recycling reduction and control. Lithium aerosol was injected from the top of the machine, with one system dropping lithium near the $X$ -point and another into the low-field side divertor leg. The injection of lithium into the SOL reduced divertor recycling, as evidenced by reduced $D_{\alpha }$ emission with ion flux measured by probes relatively unchanged. This effect is strongest in the active divertor, confirming the lithium is transported to strongly plasma-wetted areas. Quantitative analysis of the recycling changes using the SOLPS edge plasma and neutral transport code indicated a ~20% reduction in recycling coefficient with lithium injection.

Large edge-localized modes (ELMs) were mitigated by gravitational injection of lithium granules into the upper X-point region of the experimental advanced superconducting tokamak (EAST) device with tungsten plasma-facing components. The maximum ELM size was reduced by ∼70% in high β N H-mode plasmas. Large ELM stabilization was sustained for up to about 40 energy confinement times, with constant core radiated power and no evidence of high-Z or low-Z impurity accumulation. The lithium granules injection reduced the edge plasma pedestal density and temperature and their gradients, due to increased edge radiation and reduced recycling from the plasma-facing components. Ideal stability calculations using the ELITE code indicate that the stabilization of large ELMs correlates with improved stability of intermediate-n peeling-ballooning modes, due to reduced edge current resulting from the profile changes. The pedestal pressure reduction was partially offset by a core density increase, which resulted in a modest ∼7% drop in core stored energy and normalized energy confinement time. We surmise that the remnant small ELMs are triggered by the penetration of multiple Li granules just past the separatrix, similar to small ELMs triggered by deuterium pellet Futatani et al (2014 Nucl. Fusion 54 073008). This study extends previous ELM elimination with Li powder injection Maingi et al (2018 Nucl. Fusion 58 024003) in EAST because (1) use of small, dust-like powder and the related potential health hazards were eliminated, and (2) use of macroscopic granules should be more applicable to future devices, due to deeper penetration than dust particles, e.g. inside the separatrix with velocities ∼10 m s−1 in EAST.

In this paper, we will present experimental results from EAST on the mitigation of edge localized modes (ELMs) using recently developed deuterium/lithium pellet injections as well as supersonic molecular beam injections (SMBI). Using a Laval nozzle, ELM mitigation with SMBI has been demonstrated in EAST in quasi-steady state. Using a D2 pellet injector, a giant ELM appears followed by a burst of high frequency ELMs at ∼300 Hz with duration of a few tens of milliseconds. Furthermore, for the first time, a novel technology using a simple rotating impeller to inject sub-millimeter size lithium (Li) granules at speeds of a few tens of meters per second was successfully used to pace ELMs. These experiments indicate that, on EAST, several technologies can contribute to the database supporting ELMs control in future fusion devices, such as ITER.

Lithium has been introduced into the DIII-D tokamak, and migration and retention in graphite have been characterized since no lithium was present in DIII-D initially. A new regime with an enhanced edge electron pedestal and H 98 y 2 ⩽ 2 has been obtained with lithium. Lithium deposition was not uniform, but rather preferentially deposited near the strike points, consistent with previous 13 C experiments. Edge visible lithium light (LiI) remained well above the previous background during the entire DIII-D campaign, decaying with a 2600 plasma-second e-fold, but plasma performance was only affected on the discharge with lithium injection. Lithium injection demonstrated the capability of reducing hydrogenic recycling, density, and ELM frequency. Graphite and silicon samples were exposed to a lithium-injected discharge, using the DiMES system and then removed for ex - situ analysis. The deposited lithium layer remained detectable to a depth up to 1 μm.

Over the last two EAST campaigns, lithium coatings by oven evaporation were carried out as a routine wall conditioning method and significant progresses has been achieved. By upgrading the EAST lithium coating systems, lithium area coverage increased from ∼35% in 2010 to ∼85% in 2012. Accompanying the increased lithium coverage, carbon, oxygen and molybdenum impurities were decreased to extremely low levels. In addition, hydrogen concentration was further decreased with the H/(H + D) ratio falling as low as 2.5%. The effective recycling coefficient decreased step-by-step to ∼0.89 and remained below unity for ∼100 discharges. This allowed for effective feedback control of the plasma density. The wall retention rate increased from 55% to 75%, which also indicated stronger pumping of deuterium particles with increased Li coverage. With the help of increased lithium coverage, H-mode plasmas were generally easier to obtain and the EAST parameter space was enlarged.

Deuterium fueling experiments were conducted with the NSTX Liquid Lithium Divertor (LLD). Lithium evaporation recoated the LLD surface to approximate flowing liquid Li to sustain D retention. In the first experiment with the diverted outer strike point on the LLD, the difference between the applied D gas input and the plasma D content reached very high values without disrupting the plasma, as would normally occur in the absence of Li pumping, and there was also little change in plasma D content. In the second experiment, constant fueling was applied, as the LLD temperature was varied to change the surface from solid to liquid. The D retention was relatively constant, and about the same as that for solid Li coatings on graphite, or twice that achieved without Li PFC coatings. Contamination of the LLD surface was also possible due to compound formation and erosion and redeposition from carbon PFCs.

Movable liquid lithium limiter (LLL) experiments with both free-surface and capillary-pore system (CPS) configurations were successively utilized on HT-7 in 2009. In the campaign of 2011, experiments with a new lithium (Li) limiter, which used a CPS configuration with a pore size of about 100 μm and active liquid Li injection from outside of HT-7, were performed. It was found that liquid Li could flow freely driven by only gravity. Confinement of the liquid Li was improved by using the CPS configuration. It was also found that plasma performance was improved due to low recycling and significantly reduced impurity radiation. However, when the CPS LLL is employed as the primary limiter the plasma disruptivity rate increases from ∼15% to ∼90% possibly due to Li emission.

Recently, routine coatings of plasma facing materials with lithium were carried out on EAST using both upgraded evaporative ovens and real-time injection of lithium powder. Employing daily lithium coatings of 10–30 g, the H/(H + D) ratio has been decreased below 10% and both impurity levels and MHD activity have been suppressed. Using these coating technologies, plasma performance has been improved significantly. For example, a 10 s H-mode plasma was achieved at the beginning of the 2012 EAST campaign. Techniques for removing Li coatings from the vacuum vessel have been developed in EAST and rapid recovery of plasma performance following air vents has been documented.

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 ∼160 g 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, 1300 g 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.

NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma-facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core

The first self-consistent modeling of impact of dust on edge plasmas in tokamaks with the coupled dust–plasma transport code DUSTT/UEDGE is presented. The code validation for the modeling of lithium dust with radius ∼20 μm in the plasmas is performed using 3D reconstructed dust trajectories measured during lithium dust injection experiments on NSTX. The modeling demonstrates that the dust injection with rates of several of tens mg/s can have profound effect on the edge plasma profiles, transport, and stability. The differences between the dust injection and the injection of equivalent amounts of gaseous impurities in the plasmas are discussed.

Calibrated amounts of 40 μm lithium dust and 10 μm tungsten powder have been dropped from above into the SOL of the National Spherical Torus Experiment (NSTX) to benchmark modeling of dust dynamics and transport. By combining the output from two visible-range fast cameras, 3-D trajectories are reliably obtained and have resulted in the generation of several hundred individual particle tracks. Particles are observed to undergo a variety of accelerations both parallel and perpendicular to the magnetic field, as well as abrupt large-angle changes in direction. All tracks obtained to date display particle motion that is constrained to within a few centimeters of the last closed flux surface. The 3-D trajectories are presented and compared to the location of the last closed flux surface as determined by EFIT.

Lithium, owing to its many advantages, is of immense interest to the fusion community for its use as plasma facing component (PFC) material. Various experiments are under progress in the Center for Plasma Material Interactions (CPMI) at the University of Illinois at Urbana Champaign (UIUC) aimed at understanding the plasma–lithium interactions. In one such experiment called Solid/Liquid Lithium Divertor Experiment (SLiDE), it was recently observed that the flow of liquid lithium in the presence of magnetic fields is dominated by thermoelectric Magnetohydrodynamic (TEMHD) effects. To describe these results accurately, a knowledge of the thermoelectric properties of lithium is essential. For this purpose, an apparatus to measure the Seebeck coefficient of lithium was developed. Using this apparatus, the Seebeck coefficient of lithium as a function of temperature has been obtained. The Seebeck coefficient of lithium-7 is found to gradually increase from 11 μV/K to 25 μV/K, as the temperature is raised from 25 °C to 240 °C. These measurements are in good agreement with Kendall’s thermoelectric measurements on natural Li. Furthermore, using the same apparatus, the thermoelectric curve of lithium-6 is obtained and for the first time are reported in this paper.

Recent experiments on the National Spherical Torus eXperiment (NSTX) have shown the benefits of solid lithium coatings on carbon PFC's to diverted plasma performance, in both L- and H-mode confinement regimes. Better particle control, with decreased inductive flux consumption, and increased electron temperature, ion temperature, energy confinement time, and DD neutron rate were observed. Successive increases in lithium coverage resulted in the complete suppression of ELM activity in H-mode discharges. A liquid lithium divertor (LLD), which will employ the porous molybdenum surface developed for the LTX shell, is being installed on NSTX for the 2010 run period, and will provide comparisons between liquid walls in the Lithium Tokamak eXperiment (LTX) and liquid divertor targets in NSTX. LTX, which recently began operations at the Princeton Plasma Physics Laboratory, is the world's first confinement experiment with full liquid metal plasma-facing components (PFCs). All materials and construction techniques in LTX are compatible with liquid lithium. LTX employs an inner, heated, stainless steel-faced liner or shell, which will be lithium-coated. In order to ensure that lithium adheres to the shell, it is designed to operate at up to 500–600 °C to promote wetting of the stainless by the lithium, providing the first hot wall in a tokamak to operate at reactor-relevant temperatures. The engineering of LTX will be discussed.

Lithium wall coating techniques have been experimentally explored on NSTX for the last five 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 lithium evaporation system which can evaporate up to ~ 100 g of lithium onto the lower divertor plates between lithium reloadings. The unique feature of the lithium research program on NSTX is that it can investigate the effects of lithium in H-mode divertor plasmas. This lithium evaporation system thus far has produced many intriguing and potentially important results; the latest of these are summarized in a companion paper by H. Kugel. In this paper, we suggest possible implications and applications of the NSTX lithium results on the magnetic fusion research which include electron and global energy confinement improvements, MHD stability enhancement at high beta, ELM control, H-mode power threshold reduction, improvements in radio frequency heating and non-inductive plasma start-up performance, innovative divertor solutions and improved operational efficiency.

NSTX high power divertor plasma experiments have used in succession lithium pellet injection (LPI), evaporated lithium, and injected lithium powder to apply lithium coatings to graphite plasma facing components. In 2005, following the wall conditioning and LPI, discharges exhibited edge density reduction and performance improvements. Since 2006, first one, and now two lithium evaporators have been used routinely to evaporate lithium onto the lower divertor region at total rates of 10-70 mg/min for periods 5-10 min between discharges. Prior to each discharge, the evaporators are withdrawn behind shutters. Significant improvements in the performance of NBI heated divertor discharges resulting from these lithium depositions were observed. These evaporators are now used for more than 80% of NSTX discharges. Initial work with injecting fine lithium powder into the edge of NBI heated deuterium discharges yielded comparable changes in performance. Several operational issues encountered with lithium wall conditions, and the special procedures needed for vessel entry are discussed. The next step in this work is installation of a liquid lithium divertor surface on the outer part of the lower divertor.

We report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3–5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previous ELM elimination results via Li injection into the lower carbon divertor in EAST (Hu et al 2015 Phys. Rev. Lett. 114 055001). These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs (Lang et al 2017 Nucl. Fusion 57 016030), highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.

Two lithium evaporators were used to evaporate more than 100 g of lithium on to the NSTX lower divertor region. Prior to each discharge, the evaporators were withdrawn behind shutters, where they also remained during the subsequent HeGDC applied for periods up to 9.5 min. After the HeGDC, the shutters were opened and the LITERs were reinserted to deposit lithium on the lower divertor target for 10 min, at rates of 10–70 mg/min, prior to the next discharge. The major improvements in plasma performance from these lithium depositions include: (1) plasma density reduction as a result of lithium deposition; (2) suppression of ELMs; (3) improvement of energy confinement in a low-triangularity shape; (4) improvement in plasma performance for standard, high-triangularity discharges; (5) reduction of the required HeGDC time between discharges; (6) increased pedestal electron and ion temperature; (7) reduced SOL plasma density; and (8) reduced edge neutral density.

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15–25% n e decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, n e / n GW ∼ 1), to enable n e scan capability (×2) in the H-mode, to test the ability to operate at significantly lower density (e.g., n e / n GW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m 2 ) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Lithium evaporated onto plasma facing components in the NSTX lower divertor has made dramatic improvements in discharge performance. As lithium accumulated, plasmas previously exhibiting robust Type 1 ELMs gradually transformed into discharges with intermittent ELMs and finally into continuously evolving ELM-free discharges. During this sequence, other discharge parameters changed in a complicated manner. As the ELMs disappeared, energy confinement improved and remarkable changes in edge and scrape-off layer plasma properties were observed. These results demonstrate that active modification of plasma surface interactions can preempt large ELMs.

The mission of the Current Drive eXperiment-Upgrade (CDX-U) spherical tokamak is to investigate lithium as a plasma-facing component (PFC). The latest CDX-U experiments used a combination of a toroidal liquid lithium limiter and lithium wall coatings applied between plasma shots. Recycling coefficients for these plasmas were deduced to be 30% or below, and are the lowest ever observed in magnetically-confined plasmas. The corresponding energy confinement times showed nearly a factor of six improvement over discharges without lithium PFC’s. An electron beam (e-beam) for evaporating lithium from the toroidal limiter was one of the techniques used to create lithium wall coatings in CDX-U. The evaporation was not localized to the e-beam spot, but occurred only after the entire volume of lithium in toroidal limiter was liquefied. This demonstration of the ability of lithium to handle high heat loads can have significant consequences for PFC’s in future burning plasma devices. � 2007 Elsevier B.V. All rights reserved. PACS: 28.52.Fa; 52.25.Vy; 52.40.Hf; 52.55.Fa

Lithium coatings on the graphite plasma facing components (PFCs) in NSTX are being investigated as a tool for density profile control and reducing the recycling of hydrogen isotopes. Repeated lithium pellet injection into Center Stack Limited and Lower Single Null ohmic helium discharges were used to coat graphite surfaces that had been pre-conditioned with ohmic helium discharges of the same shape to reduce their contribution to hydrogen isotope recycling. The following deuterium NBI reference discharges exhibited a reduction in density by a factor of about 3 for limited and 2 for diverted plasmas, respectively, and peaked density profiles. Recently, a lithium evaporator has been used to apply thin coatings on conditioned and unconditioned PFCs. Effects on the plasma density and the impurities were obtained by pre-conditioning the PFCs with ohmic helium discharges, and performing the first deuterium NBI discharge as soon as possible after applying the lithium coating.

In the RFX-mod reversed field pinch experiment, lithium wall conditioning has been tested with multiple scopes: to improve density control, to reduce impurities and to increase energy and particle confinement time. Large single lithium pellet injection, lithium capillary-pore system and lithium evaporation has been used for lithiumization. The last two methods, which presently provide the best results in tokamak devices, have limited applicability in the RFX-mod device due to the magnetic field characteristics and geometrical constraints. On the other side, the first mentioned technique did not allow injecting large amount of lithium. To improve the deposition, recently in RFX-mod small lithium multi-pellets injection has been tested. In this paper we compare lithium multi-pellets injection to the other techniques. Multi-pellets gave more uniform Li deposition than evaporator, but provided similar effects on plasma parameters, showing that further optimizations are required.

A test of lithium wettability was performed in high vacuum (3 × 10(-4) Pa). High magnification images of Li droplets on stainless steel substrates were produced and processed using the MATLAB(®) program to obtain clear image edge points. In contrast to the more standard "θ/2" or polynomial fitting methods, ellipse fitting of the complete Li droplet shape resulted in reliable contact angle measurements over a wide range of contact angles. Using the ellipse fitting method, it was observed that the contact angle of a liquid Li droplet on a stainless steel substrate gradually decreased with increasing substrate temperature. The critical wetting temperature of liquid Li on stainless steel was observed to be about 290 °C.

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.

Intense bursts of millimetre wave emission are emitted from plasmas in TFTR when the graphite limiter is well conditioned with elemental lithium through the injection of lithium pellets. The bursts have been observed in ohmic and neutral beam heated plasmas fuelled with deuterium only, a mixture of deuterium and tritium or tritium only. The bursting is characterized by 5-10 μs duration emission spikes of randomly varying amplitude. Spectral analysis of the millimetre wave emission reveals two components; broadband emission, which appears at frequencies below the second harmonic electron cyclotron frequency, and narrowband emission (Δf < 3 GHz), with an equivalent radiation temperature of over 1 MeV and a frequency corresponding to the upper hybrid frequency near the low field plasma edge. The equivalent radiation temperature of the bursts increases with decreasing edge density, decreasing limiter recycling and increasing plasma stored energy.

In the reversed field pinch RFX-mod at the highest plasma current of 2 MA, when error fields are not effectively feedback controlled, localized thermal loads up to tens of MW m-2 can be produced. The graphite tiles withstand such high power loads, but the high hydrogen retention makes density control extremely difficult. Several wall conditioning techniques have been optimized in the last campaigns, including helium glow discharge cleaning and wall boronization by diborane glow discharges. More recently, lithium conditioning has been applied for the first time in a reversed field pinch by the evaporation technique. The main results are discussed in this paper. Lithization leads to important operational advantages: a significant improvement of the density control is obtained. Densities up to n/nG ≈ 0.5 can be produced in a controlled way. At the same value of input power, plasmas at higher densities can be sustained. However, due to the short particle confinement time, such densities are reached with high rates of gas puffing and the resulting profiles at high density are edge peaked. A lithium multipellet injector, to be applied in order to obtain a more uniform deposition, has been tested. © 2013 IAEA, Vienna.

High-resolution electron cyclotron emission (ECE) image reconstruction has been used to observe (m,n)=(2,1) and (3, 2) island structures on Tokamak Fusion Test Reactor [Plasma Phys. Controlled. Fusion 33, 1509 (1991)], where m and n are the poloidal and the toroidal mode number, respectively. The observed island structure is compared with other diagnostics, such as soft x-ray tomography and magnetic measurements. A cold elliptic island is observed after lithium pellet injection. Evidence for the enhancement of the heat transfer due to the island is observed. A relaxation phenomenon due to the m=2 mode is newly observed in Ohmic plasmas.

Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium–tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 1021 m−3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high‐performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral‐beam heating is begun.

In the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] a substantial improvement in fusion performance has been realized by combining the enhanced confinement due to tritium fueling with the enhanced confinement due to extensive conditioning of the limiter with lithium. This combination has resulted in not only significantly higher global energy confinement times than have previously been obtained in high current supershots, but also in the highest central ratio of thermonuclear fusion output power to input power observed to date.

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.

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. 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. __________________________________________________

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.

A lithium particle dropper was installed on NSTX during the 2008 campaign. Though the primary purpose of the dropper was to study the effects of Li wall conditioning, this experimental configuration also afforded a unique opportunity to study the interaction of dust of a known size and composition with a reactor grade plasma. Li powder was dropped into NSTX at a rate of ∼ 1–35 mg/sec attaining a velocity of ∼ 5 m/s at the plasma boundary. The individual particles rapidly become incandescent due to their interaction with the plasma. Two fast visible cameras, spatially separated by up to 60 degrees, simultaneously viewed the individual particles. Data from the two cameras was used to reconstruct the first detailed 3-D trajectory information of dust in a plasma with a known size, composition and velocity using a 3-D tracking code [1]. These data will be used to further constrain the comparisons to the dust transport codes such as DUSTT used to predict dust behavior in future fusion reactors. Although Li is not presently considered a reactor material, its mass and thus its behavior should be close to that of Be which is a first wall material on ITER.

Periods of edge localized mode (ELM)-free H-mode with increased pedestal pressure and width were observed in the DIII-D tokamak when density fluctuations localized to the region near the separatrix were present. Injection of a powder of 45 µm diameter lithium particles increased the duration of the enhanced pedestal phases to up to 350 ms, and also increased the likelihood of a transition to the enhanced phase. Lithium injection at a level sufficient for triggering the extended enhanced phases resulted in significant lithium in the plasma core, but carbon and other higher Z impurities as well as radiated power levels were reduced. Recycling of the working deuterium gas appeared unaffected by this level of lithium injection. The ion scale, kθρs ~ 0.1–0.2, density fluctuations propagated in the electron drift direction with f ~ 80 kHz and occurred in bursts every ~1 ms. The fluctuation bursts correlated with plasma loss resulting in a flattening of the pressure profile in a region near the separatrix. This localized flattening allowed higher overall pedestal pressure at the peeling–ballooning stability limit and higher pressure than expected under the EPED model due to reduction of the pressure gradient below the 'ballooning critical profile'. Reduction of the ion pressure by lithium dilution may contribute to the long ELM-free periods.

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 $\mathrm{ITER}98\mathrm{P}(y,1)$ by $2--3$ times, and represent the largest increase in energy confinement ever observed for an Ohmic tokamak plasma. Measurements of ${D}_{\ensuremath{\alpha}}$ emission indicate that global recycling coefficients decrease to approximately 0.3, the lowest documented for a magnetically confined hydrogen plasma.

The role of lithium on the modification of recycling regimes in fusion reactors has renewed interest of previous lithium supershot experiments carried out in TFTR. There is a need to understand the interaction between edge plasmas and lithiated plasma‐facing components (PFCs), which have the potential of enabling fusion reactors to operate at low‐recycling regimes. The Interaction of Materials with Particles and Components Testing (IMPACT) facility at Argonne National Laboratory is currently collaborating with Princeton Plasma Physics Laboratory (PPPL) to conduct lithiated surface studies for the National Spherical Tokamak Experiment (NSTX) and the Current Drive eXperiment — Upgrade (CDX‐U). IMPACT has the necessary tools to perform experiments that diagnose the surface dynamics of lithium thin films on metallic and non‐metallic substrates, and can be monitored with multiple in‐situ techniques (LEISS, AES, QMS and XPS) capturing real‐time surface dynamics. Therefore, these techniques are available during ...

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.

Developing a reactor-compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and other plasma performance benefits. During the 2010 NSTX campaign, application of a relatively modest amount of Li (300 mg prior to the discharge) resulted in a ∼50% reduction in heat load on the liquid lithium divertor (LLD) attributable to enhanced divertor bolometric radiation. These promising Li results in NSTX and related modelling calculations motivated the radiative LLD concept proposed here. Li is evaporated from the liquid lithium (LL) coated divertor strike-point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating the divertor heat removal. The LL coating of divertor surfaces can also provide a ‘sacrificial’ protective layer to protect the substrate solid material from transient high heat flux such as the ones caused by the edge localized modes. By operating at lower temperature than the first wall, the LL covered large divertor chamber wall surfaces can serve as an effective particle pump for the entire reactor chamber, as impurities generally migrate towards lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity (e.g., ∼1 l s−1 for ∼1% level ‘impurities’) is envisioned for a steady-state 1 GW-electric class fusion power plant.

A substantial improvement in supershot fusion plasma performance has been realized by combining the enhanced confinement due to tritium fueling with the enhanced confinement due to extensive Li conditioning of the TFTR limiter. This combination has resulted in not only significantly higher global energy confinement times than had previously been obtained in high current supershots, but also the highest ratio of central fusion output power to input power observed to date.

Lithium wall coatings have been shown to reduce recycling, suppress edge-localized modes (ELMs), and improve energy confinement in the National Spherical Torus Experiment (NSTX). Here we document the effect of gradually increasing lithium wall coatings on the discharge characteristics, with the reference ELMy discharges obtained in boronized, i.e. non-lithiated conditions. We observed a continuous but not quite monotonic reduction in recycling and improvement in energy confinement, a gradual alteration of edge plasma profiles, and slowly increasing periods of ELM quiescence. The measured edge plasma profiles during the lithium-coating scan were simulated with the SOLPS code, which quantified the reduction in divertor recycling coefficient from ∼98% to ∼90%. The reduction in recycling and fuelling, coupled with a drop in the edge particle transport rate, reduced the average edge density profile gradient, and shifted it radially inwards from the separatrix location. In contrast, the edge electron temperature (T e) profile was unaffected in the H-mode pedestal steep gradient region within the last 5% of normalized poloidal flux, ψ N ; however, the T e gradient became steeper at the top of the H-mode pedestal for 0.8 ψ N

Reductions in H content and particle recycling are important for the improvement of ion cyclotron range of frequency (ICRF) minority heating efficiency and the enhancement of plasma performance of the EAST superconducting tokamak. During recent years several techniques of surface conditioning such as baking, glow discharge cleaning/ICRF discharge cleaning, surface coatings, such as boronization, siliconization and lithium coating, have all been attempted in order to reduce the H/(H+D) ratio and particle recycling in EAST. Even though boronization and siliconization were both reasonably effective methods to improve plasma performance, lithium coatings were observed to reduce the H content and particle recycling to levels low enough to allow the attainment of enhanced plasma parameters and operating modes on EAST. For example, by accomplishing lithium coating using either vacuum evaporation or the real-time injection of fine lithium powder, the H/(H+D) ratio could be routinely decreased to about 5%, which significantly improved ICRF minority heating efficiency during the autumn campaign of 2010. Due to the reduced H/(H+D) ratio and lower particle recycling, and a reduced H-mode power threshold, improved plasma confinement and the first EAST H-mode plasma were obtained. Furthermore, with increasing accumulation of deposited lithium, several new milestones of EAST performance, such as a 6.4 s-long H-mode, a 100 s-long plasma duration and a 1 MA plasma current, were achieved in the 2010 autumn campaign.

The first high-confinement mode (H-mode) with type-III edge localized modes at an H factor of H IPB98(y,2) ∼ 1 has been obtained with about 1 MW lower hybrid wave power on the EAST superconducting tokamak. The first H-mode plasma appeared after wall conditioning by lithium (Li) evaporation before plasma breakdown and the real-time injection of fine Li powder into the plasma edge. The threshold power for H-mode access follows the international tokamak scaling even in the low density range and a threshold in density has been identified. With increasing accumulation of deposited Li the H-mode duration was gradually extended up to 3.6 s corresponding to ∼30 confinement times, limited only by currently attainable durations of the plasma current flat top. Finally, it was observed that neutral density near the lower X-point was progressively reduced by a factor of 4 with increasing Li accumulation, which is considered the main mechanism for the H-mode power threshold reduction by the Li wall coatings.

Increasing the amount of lithium pellet injection during the supershot conditioning procedures has enabled reliable enhancement of supershot confinement at higher plasma currents. Some shots have exceptionally good performance, with peak global parameters up to {sup {tau}}E=205 ms, S{sub n}=5.6 {times} 10{sup 16}n/s and Q{sub DD}=2. 1 {times} 10{sup {minus}3}.

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...

The Current Drive Experiment-Upgrade [T. Munsat, P. C. Efthimion, B. Jones, R. Kaita, R. Majeski, D. Stutman, and G. Taylor, Phys. Plasmas 9, 480 (2002)] spherical tokamak research program has focused on lithium as a large area plasma-facing component (PFC). The energy confinement times showed a sixfold or more improvement over discharges without lithium PFCs. This was an increase of up to a factor of 3 over ITER98P(y,1) scaling [ITER Physics Basis Editors, Nucl. Fusion 39, 2137 (1999)], and reflects the largest enhancement in confinement ever seen in Ohmic plasmas. Recycling coefficients of 0.3 or below were achieved, and they are the lowest to date in magnetically confined plasmas. The effectiveness of liquid lithium in redistributing heat loads at extremely high power densities was demonstrated with an electron beam, which was used to generate lithium coatings. When directed to a lithium reservoir, evaporation occurred only after the entire volume of lithium was raised to the evaporation temperature. T...

High resolution, rapidly sampled measurements of the light emission from injected impurity pellets have recently been carried out on TFTR. Both wide and narrow views of the pellet light have been accomplished with a tightly collimated fanned array of fiber‐optically coupled photomultipliers. Deep (≊100%), high frequency (40–90 kHz) modulation of the pellet emission has been observed in both neutral beam and rf heated discharges with as little as ≊1.5 MW of heating. This finding is consistent with recent work carried out on ASDEXb) with the exception that this modulation phenomenon is seen not only with hydrogenic pellets but also with lithium and boron pellets. This observation seems to point to an instability associated with the plasma surrounding the pellet rather than an instability which depends upon the atomic physics of the pellet.