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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 (>10−1Pa) 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 °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 (∼180 °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 ∼μm 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.

The long-term operations of the EAST experiments result in a substantial redeposition of impurities on the first wall, which is considered as the potential impurity source during the discharge. The transport properties of lithium impurity in the scrape-off layer (SOL) of EAST have been studied by the three-dimensional (3D) edge transport code, the EMC3-EIRENE. The neutral lithium impurities from four impurity source positions on the first wall have been used in the modelling. The detailed analysis of the relationship between the impurity source positions and the density profiles of the lithium impurity for different charge states has been carried out. It is found that the density profiles of the Li1+ ions are determined by the impurity source positions, whereas the density profiles of the Li2+ and Li3+ ions are not related to the first wall source positions.

A new in-between shots evaporative lithium coating system has been successfully developed and implemented on the Experimental Advanced Superconducting Tokamak (EAST) for the 2017 campaign. This paper describes a novel technique for the introduction of lithium into the EAST vessel allowing evaporative coatings to be applied between discharges without requiring a significant amount of time to insert and retract the lithium source. This newly designed system utilizes a rotating axis driven by a vacuum dynamic sealing feedthough under remote control to open and close a shutter. A compression bellows, which drives the delivery tube, guarantees accurate position of the oven thus avoiding possible direct contact with the plasma. This allows the new evaporation system to be operated in two distinct modes. One mode is operated in-between shots to refresh pre-coated lithium film, the other is a real-time lithium vapor injection during plasma operations to provide active wall conditioning during long pulse plasma discharges. In this paper, operation of this new lithium deposition system in the in-between shots replenishment mode is described. Initial experiments show that this system worked well during the interval of discharges in EAST, and the in-between shots lithium coating effectively reduced the wall recycling and impurity concentrations. Moreover, the application of fresh lithium coating lead to sustain the plasma stored energy and confinement, as well as a reduction of impurity radiation.

Reduction of hydrogen content in deuterium-fueled fusion plasmas is important not only to avoid diluting reacting core deuterons but also so as to allow the optimization of hydrogen-minority-heating efficiency in those fusion devices which employ ion-cyclotron-radio-frequency heating systems. In EAST, the amount of hydrogen released from plasma-facing components has been shown to depend strongly on both their composition and their temperature. As measured by thermal desorption spectroscopy, the hydrogen inventory in graphite – used in EAST as lower divertor material – has been determined to be >25 times larger than that of tungsten which comprises the upper divertor. This difference in hydrogen inventory is attributed mostly to the intrinsically porous nature of bulk graphite. Thus the main source of hydrogen release into EAST discharges was identified as the graphite tiles used in the lower divertor. The hydrogen content in EAST plasmas were clearly reduced by first employing a high-temperature vacuum baking of all graphite tiles and then renewing a 100–200 μm thick SiC coating before an EAST experimental run campaign. Subsequent active surface conditioning of all wall components with elemental silicon and then with elemental lithium were seem to again reduce the plasma hydrogen content significantly – with lithium proving to be more effective than silicon. Combining these several techniques, H/(H + D) levels as low as ∼3% have been achieved in EAST discharges. Additionally, the effects of lithium thickness on H surface implantation and retention has been re-examined semi-quantitatively using data from a previous run campaign. These data suggest that relatively thick Li films coated on the first wall can effectively isolate the rich source of hydrogen stored in the porous bulk of the underlying graphite from the deuterium-fueled plasma so as to minimize hydrogen release. Finally an operational maneuver whereby a diverted plasma is repetitively switched from an upper single null configuration to a lower single null configuration is presented. This switching maneuver has been shown to suppress hydrogen influx into a 35s-long EAST discharge by alternately mitigating the rise in divertor temperature.

The study of the corrosion behaviors of structural materials exposed to liquid lithium (Li) is very important for the application of liquid Li as a potential plasma-facing or blanket material in future fusion device. It was demonstrated that the temperature plays an important role in the corrosion process. The corrosion behaviors of 304 stainless steel (304SS) in static liquid Li at 600 K and 640 K for 1320 h under high vacuum have been compared in the present work. The results show that the weight loss and corrosion depth of the 304SS specimens in static liquid Li at 640 K are about 3 times more than those at 600 K. After exposure to liquid Li, all the surfaces produce non-uniform damage which includes pitting and grain boundary corrosion. The surface damage of 304SS specimens at 640 K is more serious than that at 600 K. The corrosion mechanisms include physical dissolution and chemical corrosion; the effect of the latter on the corrosion behavior at 640 K is much more than that at 600 K.

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

A 10 Hz deuterium/hydrogen pellet injection system (PI20) was deployed for plasma fueling of the Experimental Advanced Superconducting Tokamak (EAST). The pellet injection system requires high vacuum, system state monitoring, remote control, logical sequence of control stages, precise temperature control, safety interlocks, and data publishing. Owing to the requirements of system safety and operational reliability, a Supervisory Control and Data Acquisition (SCADA) system was developed and integrated to provide the functions of pellet injection process control, pellet video capture, vacuum devices control, and data acquisition. In this control system having two-layer architecture, an Omron Programmable Logic Controller (PLC) is used as the lower controller to control the process of pellet fabrication, cutting, acceleration, and relevant vacuum devices. An IPC(Industrial Personal Computer)is used as an upper computer with a LabView runtime environment that enables video capture and data archiving in conjunction with an operator interface (OPI) to monitor and control the status of the lower controller. The control system was successfully applied to the EAST campaign. In this paper, we describe the framework, development details, and experimental results of the control system.

Lithium coating technology and liquid lithium limiter have been applied on HT-7 tokamak and many significant results were obtained. The study of exploring lithium as potential plasma facing material is being carried out on EAST tokamak. A flowing liquid lithium (FLiLi) limiter has been successfully tested in the EAST in 2014, and the in-vessel electromagnetic (EM) pump was validated to make liquid lithium circulate from the bottom collector to the distributor of the limiter, which realized the FLiLi operated under steady-state and also drastically reduced the total amount of lithium used in the experiment. To adapt to the plasma shape during discharge, the limiter can be moved along the guide rail into vacuum vessel to meet various plasma scenarios by the driving system. During the operation, the lithium can be easily controlled by adjusting the current of EM pump and the

A control system of a flowing liquid lithium (FLiLi) limiter for the Experimental Advanced Superconducting Tokamak (EAST) was developed and implemented. The control system is not only able to control the direct current (DC) electromagnetic pump and heating power but can also set scanning parameters, receive the shot number, acquire the temperature, etc. The system consists of multifunctional LAN eXtensions for Instrumentation (LXI) instrument, temperature-acquisition module, programmable DC power supply, and programmable logic controller (PLC). The multi-range DC power supply is programmed to meet the operational requirements of the DC electromagnetic pump. The LXI instrument and temperature-acquisition module are used to obtain temperature data. The PLC is adopted to control the temperature of the FLiLi limiter. A safety interlock and protection function was developed for the FLiLi limiter control system. The software was designed by using LabVIEW to achieve data interaction between multiple protocols. The FLiLi limiter control system can acquire experimental data at a speed of 100 S/s and store it for later analysis. The control system was successfully applied to a FLiLi limiter to study the interaction between plasma and a fixed wall in the EAST campaign. This paper presents the framework, the implementation details, and results of the control system.

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/m2 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℃ 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.

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 ℃/min, 10.2 ℃/min or 31.7 ℃/kL and 28.1 ℃/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/m2 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.

A high frequency deuterium pellet injection system based on extrusion and pneumatic acceleration has been recently developed and built for EAST tokamak, which is one of fully superconducting Tokamaks with ITER-like plasma configuration. It is mainly used for plasma edge localized mode control and can achieve 1–50 Hz injection rate. It has been completely installed on EAST up to now. A new differential pumping system and pellet transmission lines have been designed according to the actual condition of EAST. Recently we have tested the system systematically and injected multiple pellets into type-I ELMy H-mode plasma on EAST for the first time. The test and experimental results show that the system can inject continuous pellets into plasma stably, and ELMs can be triggered by deuterium pellets in H-mode plasma with negligible impact on density and stored energy. And it was also found that ELM trigger efficiency is less than 100 % sometimes. This lays an important foundation for further experimental study of ELM modulation using deuterium pellet injection on EAST.

Two different designs of flowing liquid lithium limiter were first tested for power exhaust and particle removal in HT-7 in 2012 autumn campaign. During the experiments, the reliability and compatibility of the limiters within Tokamak were experimentally demonstrated, and some positive results were achieved. It was found that the flowing liquid lithium limiter was effective for suppressing H concentration and led to a low ratio of H/(H + D). O impurity was slightly decreased by using limiters as well as when using a Li coating. A significant increase of the wall retention ratio was also observed which resulted from the outstanding D particles pumping ability of flowing liquid lithium limiters. The strong interaction between plasma and lithium surface could cause lithium ejection into plasma and lead to disruptions. The stable plasmas produced by uniform Li flow were in favor of lithium control. While the limiters were applied with a uniform Li flow, the normal plasma was easy to be obtained, and the energy confinement time increased from ∼0.025 s to 0.04 s. Furthermore, it was encouraging to note that the application of flowing liquid lithium limiters could further improve the confinement of plasma by ∼10% on the basis of Li coating. These remarkable results will help for the following design of flowing liquid lithium limiter in EAST to improve the plasma operation.

In 2012, lithium coating with an upgraded system on EAST, the first application of lithium granules injection for ELMs pacing on EAST, and the first flowing lithium limiter experiments on HT-7 have successfully been carried out and several new results were obtained. On EAST, it was found that both the Mo first walls and the C divertors were well coated by lithium and the lithium film coverage was increased up to 85%, which greatly contributed to the new achievements of EAST, especially stationary H-mode plasma over 30 s and long pulse plasma over 400 s. And at the same time, ELMs suppression by active lithium conditioning and ELMs pacing using lithium granules injection were demonstrated and reported for the first time on EAST. On HT-7, flowing liquid lithium limiters using the TEMHD concept and using a thin flowing film concept were also initially tested and some references were obtained for the future development. Those experiments show that lithium should be an important material for fusion devices. It could be used for wall conditioning, ELMs mitigation and also provide a self-recovery plasma facing components in future fusion devices.

Experiments with liquid lithium limiters (LLLs) have been successfully performed in HT-7 since 2009 and the effects of different limiter surface structures on the ejection of Li droplets have been studied and compared. The experiments have demonstrated that strong interaction between the plasma and the liquid surface can cause intense Li efflux in the form of ejected Li droplets – which can, in turn, lead to plasma disruptions. The details of the LLL plasma-facing surface were observed to be extremely important in determining performance. Five different LLLs were evaluated in this work: two types of static free-surface limiters and three types of flowing liquid Li (FLLL) structures. It has been demonstrated that a FLLL with a slowly flowing thin liquid Li film on vertical flow plate which was pre-treated with evaporated Li was much less susceptible to Li droplet ejection than any of the other structures tested in this work. It was further observed that the plasmas run against this type of limiter were reproducibly well-behaved. These results provide technical references for the design of FLLLs in future tokamaks so as to avoid strong Li ejection and to decrease disruptive plasmas.

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.

Lithium has the ability of H recycling suppression and impurities absorption and it can be used as plasma facing material (PFM) in tokamaks. Lithium conditioning experiments were launched on EAST, HT-7 and some other tokamaks for many years by using the methods of GDC, IRCF and evaporation. Liquid lithium has better performances in effective lifetime and heat removal aspects compared to non-liquid lithium. While, applying liquid lithium in the tokamak would cause the safety problem as the lithium can react with many substances violently and the magnetohydrodynamic behavior is difficult to be handled. EAST liquid lithium limiter (LLL) system is under developing and will be applied in EAST to study the main technologies of the liquid lithium application. The normal operation temperature of the limiter is expected as 230–550 °C under the active cooling of water. Capillary porous system (CPS) is used to prevent the lithium from splashing under large electromagnetic force by increasing the surface tension of the lithium. In order to investigate the cooling performance of the cooling design, the thermal-hydraulic analysis was done which shows that with 3 m/s flowing velocity, the water can keep the limiter under 550 °C all the time if the heat flux is lower than 0.7 MW/m2. Under heat flux of 1 MW/m2, the limiter should be retreated within 7 s to avoid erosion. The pressure drop of the coolant under 3 m/s is less than 40 kPa with temperature difference nearly 34 °C which meet the design requirements very well. The key manufacture process and technologies like vacuum bonding between the CuCrZr heat sink and 316L guide plate were well studied in the R&D process. The heating test on the test bench showed that the CPS can be heated efficiently by the heaters attached into the heat sink.

This paper is a report on using a liquid jet to generate spherical lithium granules which can be used in fusion devices. The liquid jet uses liquid lithium as the continuous phase and liquid paraffin as the dispersed phase which can effectively prevent the lithium granule from oxidation and carbonization considering the active chemical property of lithium. The liquid lithium flows through a nozzle and breaks up into globules that have diverse diameters. A deeper investigation indicates that the temperature and pressure difference, which directly influence the flow velocity, have impacts on the distribution of granule size. The increases of temperature and pressure result in the granule size approaches the orifice diameter. Thus, suggesting a method to generate small spherical lithium granules that are controllable in size by adjusting the orifice size, temperature and pressure difference.

Lithium is a very attractive element due to its very low radiation power, strong H retention as well as strong O getter activity. Flowing liquid lithium (FLiLi) device, to be used as a plasma-facing limiters, has been designed and will be tested in HT-7 tokamak. It is mainly composed of distributor, guide plate, collector, and heater as well as cooling loop. The heater uses heater strip and cooling loop design, to control the temperature of lithium on the guide plate ranging from 200 °C to 400 °C. The distributor attached to feeding pipe, distributes liquid lithium (LiLi) flowing on the guide plate. The collector was designed to reclaim the superfluous LiLi and transport it out of device. The paper focuses on the design of flowing liquid lithium device. In addition to the process of design, thermal analysis has been carried out using finite element method (FEM) for optimizing the structure of heater and cooling loop and results of analysis are presented.

First experiment of liquid lithium limiter was successfully carried out on HT-7 tokamak and a few positive results were obtained. The results showed that by using lithium limiter, specially liquid lithium limiter, Hα intensity reduced 20–30%, the emission of CIII and OV decreased about 10–20%, loop voltage had a slight decline, the core electron temperature slightly increased, the particle confinement time increased by a factor of 2, and the energy confinement time increased 20%. After lithium coating, the hydrogen recycling decreased, and core electron temperature increased significantly by a factor of 2. At the same time, after lithium coating, electron density of edge plasmas obviously decreased while electron temperature slightly increased. These encouraging results are very useful for further research of long tray lithium limiter on HT-7 and liquid divertor on EAST.