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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

Results from KSTAR powder injection experiments, in which tens of milligrams of boron nitride (BN) were dropped into low-power H-mode plasmas, show an improvement in wall conditions in subsequent discharges and, in some cases, a reduction or elimination of edge-localized modes (ELMs). Injected powder is distributed by the plasma flow and is deposited on the wall and, over the course of several discharges, was observed to gradually reduce recycling by 33%, and decrease both the ELM amplitude and frequency. This is the first demonstration of the use of BN for ELM mitigation. In all of these experiments, an Impurity Powder Dropper (IPD) was used to introduce precise, controllable amounts of the materials into ELMy H-mode KSTAR discharges. The plasma duration was between 10 s and 15 s, I p = 500 kA, B T = 1 . 8 T, P NBI = 1 . 6 MW, and P ECH = 0 . 6 MW. Plasma densities were between 2 and 3 × 1 0 19 m−3. In all cases, the pre-fill and startup gas-fueling was kept constant, suggesting that the decrease in baseline D α emission is in fact due to a reduction in recycling. The results presented herein highlight the viability of powder injection for intra-shot and between-shot wall conditioning.

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.

Injection of low-Z granules into high performance discharges on DIII-D has been shown to promptly trigger Edge Localized Modes (ELMs) providing high-Z impurity control without significant plasma degradation. The ability to provide ELM triggering over a range of injection and discharge parameters suggests that the mechanical introduction of granules can be considered as an additional method of impurity control in ITER. Utilizing a spherically symmetric vapor shielding model for granule ablation, benchmarked with impurity granule injections on DIII-D, we simulate the injection of beryllium granules into ITER baseline discharges. By comparing the granule induced ELM triggering size required for deuterium and non-fuel pellets on DIII-D and cross-correlating with a previously simulated JOREK calcuation of D pellet size required for ELM triggering in ITER, we estimate that a beryllium pellet of 1.5 mm diameter should provide reliable ELM triggering on ITER. This size pellet, delivered at 200 m/s should penetrate 3.5 cm past the separatrix, solidly within the H-mode steep gradient region, a location found to be advantageous for ELM triggering with minimal pellet size. Keywords: ELM pacing, granule injection, beryllium

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

Injection of solid nonfuel pellets has been actively used as a tool for pacing and mitigation of edge localized modes (ELMs). In DIII-D, effective ELM pacing has been demonstrated by the high-frequency injection of lithium and carbon submillimeter spheres, using the impurity granule injector (IGI). This device injects granules into the plasma at speeds up to 150 m/s, through impact with a rotating impeller. In the IGI, high-frequency granule delivery was accomplished through a vibrational granule dropper, in which high time-average rates are obtained at the cost of lack of period control. We present a new in-line granule feeder, capable of delivering granules of size 0.2–2 mm with no restriction of material properties, at quasi-periodic rates of up to 150 Hz, for 0.7-mm diameter lithium granules (600 Hz using 0.3-mm granules). The new dropper mechanism combines two piezo-in-line units; one which feeds the impeller and one which circulates granules that are filtered out of the feeder path. A remotely adjustable filter eliminates granules that are stacked, oversized, or side by side allowing the formation of a single moving granule injection line. The granules fall off the in-line feeder exit one at a time, achieving a quasi-periodic delivery rate proportional to the exit speed. At higher rates, the periodicity deteriorates. This behavior was studied using high-speed cameras and electrostatic measurements, and it was found that at drop rates

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.

A critical challenge facing the basic long-pulse H-mode for ITER is to control edge-localized modes (ELMs). A new method using a multichamber lithium (Li) granule injector (LGI) for ELM triggering experiments has been developed in Experimental Advanced Superconducting Tokamak (EAST). First experimental results of the control of ELMs are obtained in EAST with a tungsten divertor. It is found that the injector has good capacities, i.e., allowing good flexibilities in granule size selection, injection rate, and injection velocity. LGI has successfully triggered ELMs during the H-mode. These results indicate the LGI would be a promising method to control ELMs in long-pulse steady-state tokamaks.

Experiments have been conducted on DIII-D investigating high repetition rate injection of non-fuel pellets as a tool for pacing Edge Localized Modes (ELMs) and mitigating their transient divertor heat loads. Effective ELM pacing was obtained with injection of Li granules in different H-mode scenarios, at frequencies 3–5 times larger than the natural ELM frequency, with subsequent reduction of strike-point heat flux (Bortolon et al., Nucl. Fus., 56, 056008, 2016). However, in scenarios with high pedestal density (∼6 × 1019 m−3), the magnitude of granule triggered ELMs shows a broad distribution, in terms of stored energy loss and peak heat flux, challenging the effectiveness of ELM mitigation. Furthermore, transient heat-flux deposition correlated with granule injections was observed far from the strike-points. Field line tracing suggest this phenomenon to be consistent with particle loss into the mid-plane far scrape-off layer, at toroidal location of the granule injection.

An impurity granule injector on the DIII-D tokamak (IGI) injects granules into the plasma to trigger Edge Localized Modes (ELMs). Impurities, such as lithium, carbon, and boron, are used. The IGI d...

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.

At DIII-D, lithium granules were radially injected into the plasma at the outer midplane to trigger and pace edge localized modes (ELMs). Granules ranging in size from 300 to 1000 microns were horizontally launched into H-mode discharges with velocities near 100 m/s, and granule to granule injection frequencies less than 500 Hz. While the smaller granules were only successful in triggering ELMs approximately 20% of the time, the larger granules regularly demonstrated ELM triggering efficiencies of greater than 80%. A fast visible camera looking along the axis of injection observed the ablation of the lithium granules. The duration of ablation was used as a benchmark for a neutral gas shielding calculation, and approximated the ablation rate and mass deposition location for the various size granules, using measured edge plasma profiles as inputs. This calculation suggests that the low triggering efficiency of the smaller granules is due to the inability of these granules to traverse the steep edge pressure gradient region and reach the top of the pedestal prior to full ablation.

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.

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.

The impact of lithium (Li) on plasma performance was investigated at the ASDEX Upgrade tokamak, which features a full tungsten wall. Li pellets containing 1.6 × 1020 Li atoms were launched with a speed of 600 m s−1 to achieve deep penetration into the plasma and minimize the impact on the first wall. Homogeneous transient Li concentrations in the plasma of up to 15% were established. The Li sustainment time in the plasma decreased with an increasing heating power from 150 to 40 ms. Due to the pellet rate being restricted to 2 Hz, no Li pile-up could take place. No significant positive impact on plasma properties, as reported from other tokamak devices, could be found; the Li pellets rather caused a small reduction in plasma energy, mainly due to enhanced radiation. Due to pellet injection, a short-lived Li layer was formed on the plasma-facing components, which lasted a few discharges and led to moderately beneficial effects during plasma start-up. Most pellets were found to trigger type-I ELMs, either by their direct local perturbation or indirectly by the altered edge conditions; however, reliability was less than 100%.

The injection of boron (B) and boron nitride (BN) powders into ASDEX Upgrade H-mode discharges have demonstrated effective control of tungsten influx in low density/collisionality operational regimes, similar to conventional boronization methods. Sub-mm powder particles are gravitationally accelerated into the upper edge of a lower single null H-mode plasma with a boundary shape roughly conforming to the shape of the poloidal midplane limiters. Visible spectroscopy measurements at one of the outer limiter showed increases in both B and N signal levels, as well as elevated B levels in the divertor, and an increase in total radiated power by greater than a factor of two during BN injection. Globally the BN injection improved energy confinement by 10%–20%, associated with improvements in pedestal performance similar to gaseous N injection. Following conditioning discharges with B powder injection, three low gas-fueling discharges with magnetic perturbations for ELM suppression were successfully conducted. These first results suggest that the application of B containing powders can be used to both improve plasma performance in real-time, and to improve overall wall conditions for subsequent discharges.

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.

The intense heat and energetic particle fluxes expected in ITER and future magnetic fusion reactors pose prohibitive problems to the design, selection and maintenance of the first wall and divertor. Micropellet injection (MPI) technologies can offer some innovative solutions to the material and extreme heat challenges. Basic physics of micropellet motion, ablation and interactions with high-temperature plasmas and energetic particles are presented first. We then discuss MPI technology options and applications. In addition to plasma diagnostic applications, controlled injection of micropellets of different sizes, velocities and injection frequencies will offer several possibilities: (1) better assessment of the core plasma cooling due to dust produced in situ; (2) better understanding of the plasma–material interaction physics near the wall; (3) new methods for plasma fuelling and impurity control; and (4) techniques for edge cooling with minimal impact on the plasma core. Dedicated small-scale laboratory experiments will complement major fusion experiments in development and applications of MPI.

This paper reports the main results of the RFX-mod fusion science activity since the last 2014 IAEA Fusion Energy Conference. The RFX-mod device is characterized by a unique flexibility in terms of accessible magnetic configurations. Axisymmetric and helically shaped Reversed-field pinch equilibria have been studied, along with tokamak plasmas in a wide range of q(a) regimes (spanning from 4 down to 1.2 values). The full range of magnetic configurations in between the two, the so-called ultra low-q ones, has been explored, with the aim of studying specific physical issues common to all equilibria, such as, for example, the density limit phenomenon. The powerful RFX-mod feedback control system has been exploited for MHD control, which allowed to extend the range of experimental parameters, as well as to induce specific magnetic perturbations for the study of 3D effects. In particular, transport, edge and isotope effect in 3D equilibria have been investigated, along with runaway mitigations through induced magnetic perturbations. The first transitions to H-mode in circular and D-shaped tokamak plasmas have been obtained thanks to an active modification of the edge electric field through a polarized electrode. The experiments are supported by intense modelling with 3D MHD, gyrokinetic, guiding center and transport codes. Proposed modifications to the RFX-mod device, which will enable further contributions to the solution of key issues in the roadmap to ITER and DEMO, are also briefly presented.

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.

Impurity seeding of tokamak edge plasmas in the form of dust is modeled using the recently coupled DUSTT/UEDGE code. The effects on edge plasma performance due to lithium and carbon dust continuously injected at different poloidal locations in NSTX and ITER are modeled. Modeling demonstrates that dust injection into the edge plasma with rates of the order of several 10 mg · s–1 for NSTX and 1 g · s–1 for ITER can significantly increase radiation power losses leading to substantial reduction of the divertor heat load. The impact of different rates of dust injection on plasma operation and stability is discussed. The simulated edge plasma performance with injected dust is compared with the case of injection of an equivalent amount of dust material in the form of atomic vapor. It is shown that injection of dust in ITER can lead to better assimilation of impurities in the divertor plasma as compared to neutral gas injection (© 2012 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Introduction of micron-sized dust into the scrape-off layer (SOL) of a plasma has recently found many applications aimed primarily at determining dust behavior in future fusion reactors. The dust particles are typically composed of materials intrinsic to a fusion reactor. On DIII-D and TEXTOR [1] carbon dust has been introduced into the SOL using a probe inserted from below into the divertor region. On NSTX, both Li and tungsten dust have been dropped from the top of the machine into the SOL throughout the duration of a discharge, by utilizing a vibrating piezoelectric based particle dropper [2] . The original particle dropper was developed to inject passivated Li powder ∼40 μm in diameter into the SOL to enhance plasma performance. A simplified version of the dropper was developed to introduce trace amounts of tungsten powder for only a few discharges, thus not requiring a large powder reservoir. The particles emit visible light from plasma interactions and can be tracked by either spectroscopic means [3] or by fast frame rate visible cameras [4] . This data can then be compared with dust transport codes such as DUSTT [5] to make predictions of dust behavior in next-step devices such as ITER. For complete modeling results, it is desired to be able to inject pre-characterized dust particles in the SOL at various known poloidal locations, including near the vessel midplane. Purely mechanical methods of injecting particles are presently being studied using a modified piezoelectric-based powder dropper as a particle source and one of several piezo-based transducers to deflect the particles into the SOL. Vibrating piezo fans operating at 60 Hz with a deflection of ±2.5 cm can impart a significant horizontal boost in velocity. The highest injection velocities are expected from rotating paddle wheels capable of injecting particles at 10s of meters per second depending primarily on the rotation velocity and diameter of the wheel. Several injection concepts have been tested and will be discussed below.

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.

Major developments on the Mega Amp Spherical Tokamak (MAST) have enabled important advances in support of ITER and the physics basis of a spherical tokamak (ST) based component test facility (CTF), as well as providing new insight into underlying tokamak physics. For example, L–H transition studies benefit from high spatial and temporal resolution measurements of pedestal profile evolution (temperature, density and radial electric field) and in support of pedestal stability studies the edge current density profile has been inferred from motional Stark effect measurements. The influence of the q-profile and E × B flow shear on transport has been studied in MAST and equilibrium flow shear has been included in gyro-kinetic codes, improving comparisons with the experimental data. H-modes exhibit a weaker q and stronger collisionality dependence of heat diffusivity than implied by IPB98(y,2) scaling, which may have important implications for the design of an ST-based CTF. ELM mitigation, an important issue for ITER, has been demonstrated by applying resonant magnetic perturbations (RMPs) using both internal and external coils, but full stabilization of type-I ELMs has not been observed. Modelling shows the importance of including the plasma response to the RMP fields. MAST plasmas with q > 1 and weak central magnetic shear regularly exhibit a long-lived saturated ideal internal mode. Measured plasma braking in the presence of this mode compares well with neo-classical toroidal viscosity theory. In support of basic physics understanding, high resolution Thomson scattering measurements are providing new insight into sawtooth crash dynamics and neo-classical tearing mode critical island widths. Retarding field analyser measurements show elevated ion temperatures in the scrape-off layer of L-mode plasmas and, in the presence of type-I ELMs, ions with energy greater than 500 eV are detected 20 cm outside the separatrix. Disruption mitigation by massive gas injection has reduced divertor heat loads by up to 70%.

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.

The ability of an injected lithium granule to promptly trigger an edge localized mode (ELM) has been established in multiple experiments. By horizontally injecting granules ranging in diameter from 200 microns to 1 mm in diameter into the low field side of EAST H-mode discharges we have determined that granules with diameter >600 microns are successful in triggering ELMs more than 95% of the time. It was also demonstrated that below 600 microns the triggering efficiency decreased roughly with granule size. Granules were radially injected from the outer midplane with velocities ~80 m s−1 into EAST upper single null discharges with an ITER like tungsten monoblock divertor. These granules were individually tracked throughout their injection cycle in order to determine their efficacy at triggering an ELM. For those granules of sufficient size, ELM triggering was a prompt response to granule injection. By simulating the granule injection with an experimentally benchmarked neutral gas shielding (NGS) model, the ablatant mass deposition required to promptly trigger an ELM is calculated and the fractional mass deposition is determined.

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.

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.

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.

An injector device to trigger frequent edge localized modes (ELMs) was developed at Princeton Plasma Physics Laboratory (PPPL) and subsequently installed and operated on DIII-D, to inject lithium spherical granules to launch into tokamak plasmas. One size granule per shot, selectable between shots, of four different spherical lithium granule sizes are available to launch horizontally into the plasma mid-plane at speeds up to 120 m/s. Pre-sorted granules are stored in a 4-chamber reservoir containing granules at various sizes, 900, 700, 500, and 300 micron. A manually controlled gating device opens the subject reservoir compartment channeling granules to a piezo crystal with a central aperture. The disk is then electrically vibrated at its resonant frequency causing granules to fall through the aperture into a vertical tube that terminates slightly above a rotating impeller strike zone. The variable speed rotating impeller, hits the granules through an aperture gated drift tube into the plasma. The 44 cm vertical drop imparts sufficient vertical speed to the granules to reach to the impeller center. A high-speed camera records granule/impeller interaction events to provide the injection count. Another on-axis high-speed camera records the plasma edge ablation events. Impeller speed is monitored via a photodiode using an external illumination system. Initial operations injected granules at frequencies up to 120 Hz (900 micron) and >600 Hz (300 micron). ELM triggering efficiency approached 100% for 700 and 900 micron in ELMing H-mode plasmas. The injector can be used to inject other materials e.g. boron, tungsten, carbon, etc. The system and planned upgrades are discussed.

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.