As a low-Z material, lithium is widely tested and used in tokamak devices for particle recycling and impurity control to improve plasma performance. Lithium application as plasma-facing materials via different methods, such as coating, injection and flowing liquid limiter, has been investigated more than 10 years on EAST. This paper presents the development of the lithium techniques applied on EAST, and reviews the progresses on the investigation of its interaction with plasma and its effects on the improvement on the plasma performances. In addition, this paper further discusses the potential applications of lithium technique in future fusion devices.

In this study, the corrosion behaviours of molybdenum (Mo) and a Mo-based alloy (TZM) were investigated using a static immersion corrosion technique. Weight loss, surface microstructure, and corrosion depth of Mo and the TZM alloy were correlated with elements. The compatibility of Mo and the TZM alloy in static liquid Li was suitable. Mo demonstrated uniform corrosion with a homogenous dissolution of free C and Mo, and TZM alloy exhibited a nonuniform corrosion behaviour with a preferential grain boundary attack caused by the selective dissolution of free C, Ti, and Zr. When the surface oxidation layer of the samples was consumed, free C from Mo and TZM diffused into molten Li to form Li2C2 and then was captured by Zr, Ti, and Mo to form thermodynamically stable carbides, which resulted in the enrichment of the C layer near the sample surfaces. In addition, Ti and Zr acted as N-trappers in liquid Li; the formation of Zr and Ti nitrides resulted in the enrichment of N, Ti, and Zr elements on the surface and led to TZM corrosion increase. Thus, Zr and Ti depletion, pitting, and grain boundary corrosion were problematic for TZM under long-term exposure to liquid Li. Reducing the content of free C and nonmetallic N and increasing the amount of Ti and Zr carbides on the surface of and inside the raw TZM alloy helped improve the corrosion resistance of TZM in liquid Li.

Particle and heat fluxes were successfully controlled by using a continuously flowing liquid Li (FLiLi) limiter in the H-mode discharges with high plasma heating power in the Experimental Advanced Superconducting Tokamak device. There were strong interactions between the FLiLi limiter and high-power plasma with a ∼ 8.3 MW source heating power, and successively, a bright Li radiation ring was produced, which effectively decreased fuel particle recycling by approximately 50%. Due to Li efflux from FLiLi during a series of high-power discharges, an obvious real-time wall conditioning effect was produced, and fuel particle recycling further decreased. Moreover, the value of Zeff decreased from 2.3 to 1.6 due to a decrease in impurity sources; this was attributed to the accumulation of Li deposited on the first wall, which effectively protected the wall materials. The decreased recycling and impurity radiation achieved high-energy confinement plasma, and the average stored energy increased up to ∼ 290 kJ. Moreover, due to the effect of Li vapor shielding, nearly 30% plasma heat flux was dissipated before it arrived at the Li limiter. These results promote further exploration of liquid Li solutions for the critical challenge of heat flux handling and particle control in fusion power plants.

Lithium (Li) is a promising low-Z material for particle recycling and impurity control to improve plasma performance in fusion devices. In the experimental advanced superconducting tokamak (EAST), Li coating has become a routine method for wall conditioning, and a flowing liquid Li (FLiLi) limiter has been successfully tested several times. Deuterium retention characteristics in the Li film coated on the international thermonuclear experimental reactor-like tungsten divertor and FLiLi during plasma discharges, which is important for the utilization of Li in future fusion devices, were investigated in EAST. It is found that the absorption of the fuel particles by Li coatings decreases gradually, and recycling gradually increases over a series of discharges. The maximum net amount of deuterium retained reached similar to 0.8 g, corresponding to 12% deuterium in the Li. This corresponds to a whole day's worth of shots with a total of 87 plasma discharges and a total of similar to 640s plasma time after 11.75 g Li was deposited. Compared to the Li coating, it is shown that FLiLi continuously traps fuel particles and achieves a higher deuterium retention ratio over both the short- and long-term, leading to lower recycling. Meanwhile, it is also observed that the fuel particle retention ratio increases when FLiLi is closer to the plasma. This result is likely due to higher plasma heating power and limiter temperature, which cause an increased Li efflux from FLiLi due to a higher limiter temperature, which then redeposits on the other plasma-facing surfaces and increases fuel particle absorption. It is estimated that 80% of the retained D particles are captured by the continual renewal of the Li redeposition film during the FLiLi operation. This investigation would also be useful for D/T retention in future fusion devices if Li is used as a plasma-facing component.

Liquid lithium (Li) is considered as the tritium breeder as well as a potential blanket coolant and/or plasma-facing material (PFM) in fusion devices, while steels are widely used as an important structural material in the fusion reactors. Investigation of corrosion behaviors of steels exposed to liquid Li is one of the keys to simultaneous apply Li and steels as potential PFM or blanket materials in the fusion devices. Hence, in this study, the corrosion behavior of 304 stainless steel (304 SS) and 316L stainless steel (316L SS) in static liquid Li, at 600 K, for 1320 h, under Argon (Ar) atmosphere is analyzed. After exposed to liquid Li, the mass loss of 304 SS is about 3.6 times more than that of 316L SS. And the corrosion depth rate of 316L SS and 304 SS is 0.71 mu m/a and 2.6 mu m/a, respectively. Both the materials have shown a non-uniform corrosion behavior. The Signs of grain boundary attack, pitting corrosion, and corrosion products formation, due to the selective dissolution of elements and chemical reactions are also observed. The percentage of corrosion product formation (i.e., carbides) on the surface of 304 SS, is higher than 316L SS. The hardness of two SS materials has a slightly increased after liquid Li exposure. (C) 2021 Elsevier B.V. All rights reserved.

A centrifugal pellet injection system for high velocity impurity pellet injection to control ELMs and research impurity transport has been developed for the Experimental Advanced Superconducting Tokamak (EAST). The system is composed of the dropper, the centrifugal accelerator, vacuum pumping system and the diagnostic unit. The centrifugal accelerator is designed to accelerate the pellet by the rotating arm driven by a KYKY F100 turbomolecular pump with an adjusted rotating velocity from 0 to 42,300 rpm. In order to evaluate the stability of the rotating arm, the stress distribution of the rotating arm at different rotating velocities is analyzed by the stress field simulation, and the result shows the maximum stress of the rotating arm appears at the transition position between the large circular groove and the acceleration groove. By simulation analysis, it is noted that the arm can work steadily at 2500 rad/s (corresponding to about 398 Hz arm rotation frequency) rotating velocity and may be broken at 4000 rad/s (637 Hz). By the bench testing using the Li pellets of 1 mm diameter as injection material, it shows the centrifugal accelerator can work steadily for a long time under the condition that the rotating velocity doesn't exceed 2200 rad/s (350 Hz) and the maximum velocity of the pellet can reach 286 m/s. Based on the results of simulation and bench test, the centrifugal pellet injection system was installed and then injected Li pellets into the plasma steadily and continuously with the rotating velocity <2200 rad/s on EAST.

We report an observation of robust suppression of edge-localized modes (ELMs) in the Experimental Advanced Superconducting Tokamak (EAST), enabled by continuous boron (B) powder injection. Edge harmonic oscillations appear during B powder injection, providing sufficient particle transport to maintain constant density and avoid impurity accumulation in ELM-stable plasmas. Quasi-steady ELM suppression discharges are demonstrated with modest energy confinement improvement and over a wide range of conditions: heating power and technique variation, electron density range over a factor similar to 3.5, deuterium or helium ion species, and with either direction of the toroidal magnetic field. ELM suppression is observed above a threshold edge B intensity and ceases within 0.5 s of termination of the B injection. In contrast to ELM suppression accompanied by recycling reduction during Li powder injection in NSTX and EAST (Maingi et al 2018 Nucl. Fusion 58 024003), reduced recycling due to hydrogenic species retention is unnecessary for the ELM suppression with B powder injection, paving the way for its consideration as an ELM control tool for future fusion devices.

The lithium and boron powder have been successfully real-time injected into Experimental Advanced Superconducting Tokamak (EAST) high-confinement (H−) mode discharges, showing good wall conditioning effects, especially on the impurities control. Both of lithium and boron powder particles are gravitationally accelerated into the upper edge of a upper single null discharge. The lithium powder real-time injection effectively reduced the tungsten impurity content both in the plasma edge and core. Contrary to the lithium injection, the boron injection mainly reduced the impurities content in the plasma core, while the impurities increased in the edge. This mainly due to that there was an edge harmonic mode stimulated by boron injection, providing sufficient particles transport. Moreover, the impurities control through boron injection has much wider operation windows than lithium powder injection. Furthermore, there was accumulative wall conditioning effects after sequential boron powder injection discharges.

Arc spots formed on a plasma-facing component (PFC) have a stunning resemblance with the arc spots formed on a cathode of a vacuum arc. Hence, in this paper, a laboratory-scale vacuum arc process is used to get a better understanding of the arc erosion process of the tungsten (W) material. The performance of lithium (Li) and silicon (Si) coatings in erosion mitigation of W material is analyzed. The virgin W substrate was eroded significantly by arc spots and numerous W droplets were ejected into the interelectrode plasma from cathode with velocity of tens m/s. However, the W substrates can be effectively protected by Li and Si coatings from arc erosion. Preferential formation of arc spots on the coatings is responsible for the arc erosion mitigation. Compared with the virgin W substrate, no ejection of W droplets from the cathode was observed in the cases of coatings. Li and Si droplets were accidently ejected from the cathode and the ejected droplets were rapidly ablated in the arc plasma. For the case of Si coating, the W substrate was slightly melted and eroded due to the arc spots formed on the coating. By contrast, no obvious melting and erosion were observed on the W substrate for the case of Li coating, which is possibly due to the vapor shielding effect of Li between the W substrate and arc spots. These results confirm coatings served as a sacrificial protective layer and can reduce tungsten surface erosion caused by arcs, which provides a new reference for protecting first wall materials in future fusion devices.

A reproducible, quasi-stationary edge localized mode (ELM)-suppressed scenario was obtained over a wide range of plasma parameters by continuous injection of boron (B) powder into an upper-single null discharge in the experimental advanced superconducting tokamak [Sun et al., Nucl. Fusion 61, 014002 (2021)]. This powder-induced ELM-absent regime is associated with an edge harmonic mode (EHM) that provides continuous particle exhaust to maintain constant density without confinement degradation and impurity accumulation, the latter of which is often observed in ELM-free regimes. A flow rate threshold of B powder injection, leading to a threshold intensity of the EHM, is necessary for full ELM suppression. The fundamental harmonic of the EHM exhibits a toroidal mode number n = 1. The mode is observable in the entire poloidal cross section with a peak near the upper X-point in an upper-single null configuration. The EHM spans radially across the pedestal and scrape-off layer, peaking inside the separatrix. The EHM appears to be insensitive to q(95), heating power, plasma toroidal rotation, and pedestal collisionality.

Transport dynamic 2D simulations for supersonic molecular beam injection (SMBI) fueling into H-mode deuterium plasmas on the Experimental Advanced Superconducting Tokamak (EAST) is first simulated using a seven-field two-fluid model in the BOUT++ framework. The SMB is assumed to be injected into plasma from the midplane at the low field side with a fixed width and constant molecular flux. The different densities and injection velocities of SMBI are investigated within the upper single-null geometry of EAST. The simulations indicate that the SMBI has a self-shielding effect on molecules' inward transport into the plasma, and the deposition of SMBI leads to a large increase in plasma density and decrease in plasma temperature. There is a velocity threshold for SMB penetrating the pedestal and depositing at the top of density pedestal. However, the deposition point would be back toward the plasma boundary after SMB arriving at the deepest penetration position. Comparing the different molecular injection velocities and densities, the outcomes demonstrate that the depth of deposition is closely related to the injection velocity rather than the molecular density. The simulated results show good agreement with the EAST experiments by comparing the electron density profiles obtained by simulation and experiment, respectively. These results will be helpful for guiding future experiments, as well as the design of the SMBI system.

A novel continuously flowing liquid lithium limiter (FLiLi) which employs an in-vessel electro-magnetic pump to drive liquid Li flowing on the surface of limiter has been successfully designed and tested in the experimental advanced superconducting tokamak (EAST) device in 2014. In order to better control the surface temperature, an upgraded design and the real-time gas cooling were performed in 2016. Two kinds of cooling gas, helium (He) and argon (Ar) were tested on the FLiLi system prior to the experiment. It was found that the cooling rates of He and Ar at 2.5 MPa are 34.2 degrees C/min, 10.2 degrees C/min or 31.7 degrees C/kL and 28.1 degrees C/kL, respectively. The cooling performance of He is more effective than Ar. Experimental results show that the real-time He cooling could effectively contain the FLiLi surface temperature increase and prevent the strong passive evaporation of lithium during plasma discharges. Limiter heat flux up to 0.2 MW/m(2) was removed during ohmic discharge with 2.5 MPa He gas. Finally, Because of the upgraded design and real-time gas cooling, the surface of FLiLi was not damaged by heat flux after entire plasma discharge.

To study the potential applications of liquid lithium (Li) as plasma facing material, the third generation of flowing liquid Li (FLiLi) limiter has been designed and successfully tested in EAST tokamak H-mode discharges in the 2018 run campaign. The stability of liquid Li on FLiLi surface, and Li splashing and its effect on the plasma performance are investigated in detail. Li splashing was observed in L-mode and H-mode discharges due to Rayleigh-Taylor (R-T) instability, resulting from electromagnetic forces J→×B→. Li splashing in low-parameter plasma decreased both core and edge electron temperatures, resulting in the degradation of plasma energy confinement and even H-L back transition, however, it almost did not influence plasma confinement performance in high-parameter plasma. These findings indicate that high-parameter plasma has higher tolerance to Li impurities than low-parameter plasma. By contrast, tungsten splashing from the melted layer on the upper divertor was also induced by the R-T instability but generally resulted in plasma disruption, which suggests lower tolerance to tungsten impurities in EAST tokamak.

Three generation of flowing liquid lithium (FLiLi) limiters have been developed and successfully tested in EAST since 2014. It was confirmed that liquid Li could be driven by an innovative in-vessel DC electromagnetic J x B pump to form a recirculating loop. The FLiLi design was continuously upgraded, resulting in improvement of the liquid Li surface coverage (>80%), elimination of surface erosion, and the increase of heat exhaust capability. Promising results are obtained during FLiLi operations, including a reduction of impurities and recycling to improve plasma confinement, and a mitigation of ELM activity in H-mode plasmas. The FLiLi limiter program has shown basic compatibility with H-mode discharges with auxiliary heating power up to 8.3 MW. These efforts will expand to flowing liquid Li PFC in high heat flux zones, e.g. near the divertor strike point in reactors.

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 < Paux < 7.5 MW, 3.8 × 1019 < ne < 6 × 1019 m−3, RF-only and RF + NBI heating scenarios, both grad-B drift directions, and even He ion majority plasmas) (Sun et al. in Nucl. Fusion, 2020). A window of edge B concentration for stable long pulse operation was identified: too low and ELMs return, too high and the discharge suffers radiative collapse. The injection of boron powder above the minimum for ELM suppression coincided with the occurrence of an edge harmonic oscillation detected in magnetics (both on the high-field side and low-field side), in AXUV diodes near the upper X-point, divertor D $${\upalpha }$$ emission, and in a range of other diagnostics (Diallo et al. in: Proceedings of 2020 IAEA fusion energy conference, 2021). No harmonic oscillation was observed when ELMs were present, and stored energy was slightly increased at constant density during ELM suppression. Core tungsten emission during ELM suppression either increased or decreased relative to ELMy H-mode, but the W emission was maintained at acceptable levels. The threshold B injection rate was measured for several conditions, and found to increase with heating power. Li powder injection into comparable discharges also resulted in a short phase of ELM suppression, but density and stored energy both decreased due to the strong pumping effect of lithium; no edge harmonic oscillation was observed with Li injection, indicating that the ELM suppression mechanisms differ. The new set of B-seeded, ELM-suppressed discharges exhibited certain characteristics of quiescent H-mode (Burrell et al. in Phys Plasmas 8:2153, 2001), but did not require high shear, counter beams, etc. The wide operating window and compatibility with RF-only discharges paves the way for future experiments targeting long pulse H-mode discharges with complete ELM suppression.

Liquid metal plasma-facing components (PFCs) provide certain promising advantages as compared with solid PFCs. A new flowing liquid lithium (FLiLi) limiter using a Titanium-Zirconium-Molybdenum Alloy (TZM) material substrate was developed and inserted into H-mode plasmas in EAST. As compared to the previous FLiLi limiters that used a copper heat sink covered with a stainless steel (SS) protective coating, the TZM limiter has better erosion resistance and wettability. Also, helium-cooling and water-cooling simulations show acceptable performance under continuous 2 MW/m(2) heat flux deposited on the plate for 10 s. Both simulations show a peak surface temperature located at the ends of the plate where the cooling tubes can't quite reach just below 650 degrees C while ignoring the influence of Li flowing, evaporation, and vapor shielding. Moreover, FLiLi is introduced to plasma lead to lower recycling and slight storage energy of plasma increase.

Control of impurities, fuel recycling and hydrogen content by lithium evaporative coatings and real-time lithium powder injection (LPI) in EAST are studied for high performance H-mode discharges of up to similar to 100 s. The results show that the lithium evaporative coatings significantly reduced both the low-Z impurity carbon and high-Z impurity tungsten as well as molybdenum concentration in the plasmas, and the impurities concentration significantly reduced with the accumulated lithium coatings and maintained well afterwards. Specifically the high-Z tungsten core impurity concentration was maintained between 3 ppm-15 ppm during the 101 s H-mode discharge, which is acceptable for the long pulse operation. In addition, real-time wall conditioning via LPI successfully reduced the core high-Z metal impurities by 50% during similar to 35 s long pulse H-mode discharge, exhibiting strong compatibility between real-time LPI with long pulse discharges. In addition, evaporative lithium coatings demonstrated fuel recycling control, with fuel recycling obviously reducing with lithium coatings and maintaining well afterwards. Also, the lithium evaporative coatings reduced the hydrogen minority species content, represented by the density ratio H/(H + D), from similar to 50% down to similar to 5% with accumulated lithium coatings; low hydrogen fraction improved the ICRF minority heating efficiency. Finally the real-time LPI reduced the recycling coefficientR(global)from 0.95 to 0.82. With these wall conditionings help, the plasma density controlled well during the 101 s long pulse H-mode discharge. These results provide valuable references on impurities, fuel recycling and hydrogen content control for future longer pulse high performance H-mode operation (>= 400 s) in EAST and future fusion devices.

A third generation flowing liquid lithium (FLiLi) limiter with a substrate made of TZM, an alloy with>99% Mo, was fabricated by conventional manufacturing techniques. TZM has a high corrosion resistance, a high sputtering threshold, and a good wettability to Li, as compared to stainless steel (SS), which had been used as an FLiLi substrate surface in 2014 and 2016. The third generation FLiLi was inserted into the edge in EAST H-mode plasmas in an upper single-null configuration with an ion grad-B drift toward the upper divertor with a limiter temperature of 330-380 degrees C and an auxiliary heating power of about 2-8MW. Analysis has shown that by using TZM FLiLi, fuel particle recycling continuously decreased and near-complete edge localized mode elimination was achieved in H-mode plasmas with RF-only heating. The main impurities during the initial FLiLi discharges were Mo, Fe, and W resulting from strong plasma interaction at the Mo plate side, the SS collector, and the upper W divertor. Plasma stored energy increased by about 10kJ in subsequent FLiLi discharges due to decreased impurity radiation. Engineering analysis shows a uniform lithium flow with an similar to 80% Li coverage ratio on the limiter surface, similar to the second FLiLi, even though only one of two JxB pumps was functioning. Despite technical difficulties, the FLiLi gen.3 improved the overall plasma performance, providing support for flowing liquid Li plasma facing component applications in present and future devices. Published under license by AIP Publishing.

Significant engineering and physic progress was made by the upgrade of a flowing liquid lithium limiter (FLiLi) inserted into EAST H-mode discharges in 2016. The progress includes an improvement of liquid Li coverage uniformity, no macroscopic surface erosion by plasma materials interaction, effective removal of 55% of the discharge heat flux using high pressure He cooling during ohmic discharges, and the demonstrated ability to restart Li flow for a second experiment performed a week after the first. With increasing Li flow rate, fuel particle recycling and Fe impurity radiation gradually decreased, and plasma performance slightly improved, but there was no impact on the edge tungsten influx. No obvious Li bursts appeared during operation with up to about 4.5 MW auxiliary heating power with peak heat flux reaching similar to 4 MW m(-2) on FLiLi. In addition, a gradual mitigation of edge localized mode (ELM) activity was observed with FLiLi operation. Finally, short-lived ELM-free phases were observed for the first time in EAST during FLiLi insertion, with increasing tau(E) and transient H-98((y, 2)) < 2.

Improved liquid Li wettability on a solid substrate surface is a necessary component for the research and development of blanket coolants, tritium breeders, and first wall materials for fusion device. A systemic investigation of Li wetting a 316 L SS surface, along with interfacial interactions has been successfully carried out. The results indicate that base vacuum pressure of the vessel, substrate temperature, and surface conditions of the substrate are the main factors that influence Li wetting property. It has been found that poor vacuum (> 10(-1)Pa) along with high impurity content, especially water, can prevent Li wetting due to the formation of a layer of complex Li compounds, such as LiOH, Li3N and Li2CO3. Also, Li wetting can be gradually improved with an increase of substrate temperature, mainly due to a decrease in surface tension of Li droplets. Moreover, Li wetting can be further improved by effective surface conditioning. It has been demonstrated that baking was beneficial in order to release impurity gas by thermal desorption and thereby achieve high vacuum during Li wetting, which resulted in a reduction of > 50 degrees C in Li wetting temperature on the SS substrate. He-GDC and Li coating can further decrease Li wetting temperatures to the Li melting point (similar to 180 degrees C) by avoiding the formation of Li2O at the interfacial surface during Li wetting. This is achieved primarily because these conditioning techniques can effectively remove and isolate the SS surface oxide layer (Cr2O3), respectively, which was verified by the related testing of interfacial interactions between liquid Li and SS. Finally, surface textured structures with trenches similar to mu m in size, are demonstrated to promote Li spreading along the groove direction, driven by capillary forces. These results provide the technical support for liquid Li applications in future fusion reactors.