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Purpose Herringbone groove thrust bearings are typically used in high-speed, light-load applications, such as spindle motors for hard disk drives. In the past researches, the effect of shaft misalignment was little considered. This study aims to reveal effects of shaft misalignment on the microscopic flow regime in the water-lubricated herringbone groove thrust bearing. Design/methodology/approach The liquid film in a thrust herringbone groove bearing was investigated by computational fluid dynamics. The effects of micro-grooves on the flow field were carefully explored. Two-dimensional liquid films at four different sites were examined for obtaining the rich flow field properties. Findings The distributions of pressure, temperature and water vapor volume fraction were obtained, the micro hydrodynamic effects were formed by the herringbone grooves and the effects of the shaft misalignment on lubrication and sealing performance could be found. Originality/value The influence of misalignment on the herringbone groove thrust bearing performance was investigated in detail. The obtained results could give the reference guideline for the bearing design.

MARFE movement which can cause density limit disruption is often encountered during high density operation on many tokamaks. Therefore, identifying and predicting MARFE movement is meaningful to mitigate or avoid density limit disruption for the steady-state high-density plasma operation. A machine learning method named Random Forest (RF) has been used to predict the MARFE movement based on the density ramp-up experiment in the 2022’s first campaign of EAST. The RF model shows that besides Greenwald fraction which is the ratio of plasma density and Greenwald density limit, dβ p /dt, H 98 and dW mhd /dt relatively important parameters for MARFE-movement prediction. Applying the RF model on test discharges, the test results show that the successful alarm rate for MARFE movement causing density limit disruption reaches ~85% with a minimum alarm time of ~40 ms and mean alarm time of ~700 ms. At the same time, the false alarm rate for non-disruptive and non-density-limit disruptive discharges can be kept below 5%. These results provide a reference to the prediction of MARFE movement in high density plasmas, which can help the avoidance or mitigation of density limit disruption in future fusion reactors.

There is an urgently need for the high-performance rechargeable electrical storage devices as supplement or substitutions of lithium ion batteries due to the shortage of lithium in nature. Herein we propose a stable 2D electrene T'-Ca2P as anode material for Na/K ion batteries by first-principle calculations. Our calculated results show that T'-Ca2P monolayer is an antiferromagnetic semiconducting electrene with spin-polarized electron gas. It exhibits suitable adsorption for both Na and K atoms, and its anisotropic migration energy barriers are 0.050/0.101 eV and 0.037/0.091 eV in b/a direction, respectively. The theoretical capacities for Na and K are both 482 mAh/g, while the average working voltage platforms are 0.171-0.226 V and 0.013-0.267 V, respectively. All the results reveal that the T'-Ca2P monolayer has promised application prospects as anode materials for Na/K ion batteries.

Massive gas injection (MGI) is a traditional plasma disruption mitigation method. This method directly injects massive gas into the pre-disruption plasma and has been developed on EAST. Different noble gas injection experiments, including He, Ne, and Ar, were performed to compare the mitigation effect of plasma disruption by evaluating the key parameters such as flight time, pre-thermal quench (pre-TQ), and current quench (CQ). The flight time was shorter for low atomic number (Z) gas, and the decrease in flight time by increasing the amount of gas was insignificant. However, both pre-TQ and CQ durations decreased considerably with the increase in gas injection amount. The effect of atomic mass on pre-TQ and CQ durations showed the opposite trend. The observed trend could help in controlling CQ duration in a reasonable area. Moreover, the analysis of radiation distribution with different impurity injections indicates that low Z impurity could reduce the asymmetry of radiation, which is valuable in mitigating plasma disruption. These results provide essential data support for plasma disruption mitigation on EAST and future fusion devices.

Mastering nuclear fusion, which is an abundant, safe, and environmentally competitive energy, is a great challenge for humanity. Tokamak represents one of the most promising paths toward controlled fusion. Obtaining a high-performance, steady-state, and long-pulse plasma regime remains a critical issue. Recently, a big breakthrough in steady-state operation was made on the Experimental Advanced Superconducting Tokamak (EAST). A steady-state plasma with a world-record pulse length of 1056 s was obtained, where the density and the divertor peak heat flux were well controlled, with no core impurity accumulation, and a new high-confinement and self-organizing regime (Super I-mode = I-mode + e-ITB) was discovered and demonstrated. These achievements contribute to the integration of fusion plasma technology and physics, which is essential to operate next-step devices.

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

Based on deep learning, a Dust Ablation Trail Analysis (DATA) code package is developed to detect dust ablation trails in tokamaks, which is intended to analyze a large amount data of tokamak dusts. To validate and benchmark the DATA code package, 2440 plasma jet images are exploited for the training and test of the deep learning DATA code package, since plasma jets resemble the shape and size of dust ablation clouds in tokamaks. After being trained by 1920 plasma jet images, the DATA code package is able to locate 100% plasma jets, classify plasma jets with the accuracy of >99.9%, and output image skeleton information for classified plasma jets. The DATA code package trained by the plasma jet images is also used to analyze the dust ablation trails captured in the Experimental Advanced Superconducting (EAST) tokamak with the satisfactory performance, further verifying its applicability in the fusion dust ablation investigation. Based on its excellent performance presented here, it is demonstrated that our DATA code package is able to automatically identify and analyze dust ablation trails in tokamaks, which can be used for further detailed investigations, such as the three-dimensional reconstruction of dusts and their ablation trails.

In 2021, EAST realized a steady-state long pulse with a duration over 100 s and a core electron temperature over 10 keV. This is an integrated operation that resolves several key issues, including active control of wall conditioning, long-lasting fully noninductive current and divertor heat/particle flux. The fully noninductive current is driven by pure radio frequency (RF) waves with a lower hybrid current drive power of 2.5 MW and electron cyclotron resonance heating of 1.4 MW. This is an excellent experimental platform on the timescale of hundreds of seconds for studying multiscale instabilities, electron-dominant transport and particle recycling (plasma-wall interactions) under weak collisionality.

Experiments on pellet fueling have been carried out in edge localized mode (ELM)-free high-confinement mode (H-mode) plasmas with q 95 ∼ 6 in the EAST tokamak. Cryogenic deuterium pellets were injected into the ELM-free plasmas at a frequency of 10 or 5 Hz from ∼45 cm above the mid-plane on the low-field side. It is found that the ELM-free H-mode plasmas are still sustained even if both the edge and core plasma are impacted by the pellet injections (PIs). Several small ELMs would appear and the edge coherent mode accompanying the ELM-free phase fades or even disappears just after the pellet events, but the plasma would rapidly recover to the ELM-free state. Although the ELMing phase is very short, it may be an issue that still needs to be resolved in the future. Meanwhile, the 3/2 tearing mode often appearing in ELM-free discharges would be stabilized by the PIs, and the high-Z impurity concentration would be reduced during the PIs. All these results will be meaningful to International Thermonuclear Experimental Reactor and future fusion reactors.

Lithium is chemically active and prone to react with residual gases in the atmosphere, which should be taken into account for the application of lithium in fusion facility. In the present study, effect of N2 on the release temperature of D2 in liquid lithium was investigated. Absorption experiments were performed at 400 °C by exposing lithium in N2 or D2 atmosphere. Thermal desorption spectrometry (TDS) was used to measure the desorption temperature of Li samples. New D2 peaks which are below the decomposition temperature of LiD but higher than that of D2 dissolved in liquid lithium are observed in Li samples that have absorbed both D2 and N2 and the desorption temperature of which varies according to the ratio of N2 and D2. Although the mechanism of this reaction is not clear, these results highlight the importance of monitoring the impurity content of liquid lithium and the potential impact on tritium extraction when lithium is applied in fusion facilities.

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

Copper (Cu) materials are extensively used as heat sinks and sealing gaskets in fusion devices because they have the properties of good thermal and electrical conductivity and high plasticity. Meanwhile, liquid lithium (Li) is considered as a potential blanket coolant and tritium breeder and/or plasma facing material in fusion devices. Studying the corrosion characteristics of Cu materials by liquid Li under extreme fusion conditions is important because the corrosion of Cu by liquid Li may affect simultaneous application of these materials in fusion devices. The corrosion behavior of Cu in static liquid Li at 620 K for 15 h under high vacuum was investigated. After exposure to liquid Li, the weight loss rate of Cu in liquid Li is 466.1 g m−2 h−1, which is equivalent to 458.7 mm⋅a−1 for the average corrosion depth rate. The entire surface of each specimen was seriously damaged. Visible grain boundary corrosion was observed on the surface of the specimens. Also, Cu debris entered the liquid Li from the corroded surface, resulting in considerable Cu loss from the specimen. These results demonstrate a corrosion protection grade of Cu in liquid Li of 10, Cu cannot withstand the corrosion of liquid Li under the given conditions. Additionally, the corrosion process of Cu in liquid Li at 620 K under high vacuum was studied using the Large-scale Atomic/Molecular Massively Parallel Simulator (LAMMPS). The results from these simulations indicate that the corrosion of Cu in liquid Li is induced via physical dissolution and intergranular corrosion, where intergranular corrosion is the dominant mechanism.

Liquid lithium (Li) can partly ameliorate lifetime and power-exhaust issues of plasma facing components (PFCs) by enabling a self-healing, self-replenishing surface with a reduced susceptibility to neutron damage in future fusion devices. To assess operational stability and heat-exhaust capability under tokamak exposure, two generations of continuously flowing liquid Li (FLiLi) limiters on the concept of a thin flowing Li film have been successfully designed and tested in high performance discharges in EAST. The design uses a circulating Li layer with a thickness of

A lithium (Li) vapour layer was formed around a flowing liquid Li limiter to shield against the plasma incident power and reduce limiter heat flux in the EAST tokamak. The results revealed that after a plasma operation of a few seconds, the layer became clear, which indicated a strong Li emission with a decrease in the limiter surface temperature. This emission resulted in a dense vapour around the limiter, and Li ions moved along the magnetic field to form a green shielding layer on the limiter. The plasma heat flux loaded on the limiter, measured by the probe installed on the limiter, was approximately 52% lower than that detected by a fast-reciprocating probe at the same radial position without the limiter in EAST. Additionally, approximately 42% of the parallel heat flux was dissipated directly with the enhanced Li radiation in the discharge with the liquid metal infused trenches (LIMIT) limiter. This observation revealed that the Li vapour layer exhibited an excellent shielding effect to liquid Li on plasma heat flux, which is a possible benefit of liquid-plasma-facing components in future fusion devices.

Active control of the excessively high heat and particle fluxes on the divertor target plates is of fundamental importance to the steady state operation of tokamaks, especially for fusion reactors. A series of experiments have been carried out on this critical issue to relieve the plasma-wall interactions in the experimental advanced superconducting tokamak (EAST) in the last ten years, not only contributing to the long pulse operation of EAST itself, but also providing physical understandings and potential techniques to the next-generation devices like ITER. We have characterized the power deposition pattern and broadened the divertor footprint width effectively. The plasma-wetted area is actively handled using either 3-dimentional edge magnetic topology or advanced plasma equilibrium, thereby peak heat flux around the strike point is reduced. Active control of detachment or radiation compatible with core plasma performance has progressed significantly in very recent years, with a series of active feedback control modules developed and utilized successfully, based on the divertor physics advances with both experiments and simulation. The upper divertor of EAST was upgraded from graphite to active water-cooling ITER-like tungsten in 2014, exhibiting much enhanced heat removal capability. As for the particle exhaust including both fueling and impurity particles, in addition to wall conditioning and impurity source control, the efficiency of particle flux exhaust is optimized by making full use of the divertor closure and the plasma drifts in both scrape-off layer and divertor volume. These heat and particle exhaust advances contribute greatly to a series of EAST achievements like H-mode operation over 100 s. A brief near-term plan on the integrated control of divertor plasma-wall interactions in long-time scale will also be introduced, aiming to provide favorable divertor operation solution for ITER and CFETR.

The multifaceted asymmetric radiation from the edge (MARFE), which is generally considered to be the result of a radiation thermal instability in the edge and usually occurs in high density operation, has been first observed to move up and down along the poloidal cross-section due to edge cooling after cryogenic deuterium pellet injections in EAST tokamak with tungsten divertor. A maximum electron density of 0.84 × n GW has been obtained using continuous cryogenic pellet fueling. In the meantime, MARFEs, initially located near the divertors of EAST, moves to the inner wall on high field side after each pellet injection. This movement should be attributed to the asymmetry of the power flow to the two sides of the MARFE after pellet injections. Accompanied with MARFE movement, two kinds of strong density fluctuations have been observed. The ones with continuous and regular frequency spectrum, which does not cause a reduction of main plasma density, are confirmed to be induced by MARFE. The others, appearing with magnetic fluctuations, have been determined to be induced by the m/n = 2/1 magnetohydrodynamic activities after pellet injections. All the investigations in this paper will be meaningful for the steady high density operation of future fusion reactors, such as ITER.

To study the compatibility especially the corrosion characteristic of structural material stainless steel (SS) with liquid lithium (Li) is an important factor for the simultaneous application of liquid Li and SS as a blanket material or potential plasma facing material (PFM) in future fusion devices. The corrosion behaviors of 304SS exposed to liquid Li at 700 K for 1320 hr under high vacuum were investigated. After exposure to liquid Li, 304SS produces non-uniform corrosion which includes corrosion products and grain boundary corrosion. Compared with the corrosion behaviors at 600 K and 640 K, temperature plays an important role in the corrosion behaviors of 304SS in static liquid Li. The weight loss, corrosion depth and corrosion products of 304SS specimens increase with the corrosion temperature. In general, the compatibility between 304SS and static liquid Li is relatively good under the experimental conditions. And besides, it is unexpectedly found that the compatibility of copper with Li is not good.

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.

Molybdenum (Mo), tungsten (W), and stainless steel (SS) are widely used as important structure materials and first wall materials in fusion devices, while liquid lithium (Li) limiter/divertor can provide an attractive option for withstanding high heat load and solving life-time problem of first wall. Studying the compatibility of these materials exposed to liquid Li is significant for the application in Mo, W, and SS in fusion reactors. The corrosion behaviors of Mo, W, and 304SS exposed to static liquid Li at 600 K up to 1320 h under high vacuum with pressure 10−4 Pa were investigated. After exposure to liquid Li, it was found that the weight loss of Mo, W, and 304SS increases with corrosion time, but the total amount is moderate. 304SS specimens produce a non-uniform corrosion behavior because of Cr, Ni, and carbon (C) elements selectivity depletion and formation of carbides compound near surface. Mo and W surface microstructures are unchanged. 304SS surface hardness increases with corrosion products because these particles include C element, which increases by 49 HV after exposed to liquid Li for 1320 h, while Mo and W surface hardness are unchanged by the reason of their excellent corrosion resistance.

The Hybrid Illinois Device for Research and Applications (HIDRA) is a toroidal fusion device at the University of Illinois at Urbana–Champaign, Urbana, IL, USA. HIDRA is the former WEGA stellarator that was operated at the Max Planck Institut fur Plasmaphysik, Greifswald, Germany. The machine is a five-period, $l=2, m=5$ stellarator, with major radius $R_{0}=0.72$ m, and minor radius $a=0.19$ m. Initial heating is achieved with 2.45-GHz electron cyclotron resonance heating and an on-axis magnetic field of $B_{0}=0.087$ T that can go as high as $B_{0}=0.5$ T. HIDRA has the ability to operate as both a stellarator and a tokamak, initially operating in the stellarator mode. The focus of research on HIDRA will be doing dedicated studies on plasma-material interactions (PMIs) using the wealth of knowledge and experience at the Center for Plasma Material Interactions, Urbana, IL, USA. In early 2016, the first experiments were performed on HIDRA. This paper presents some of the first results obtained from the machine such as initial magnetic fields’ measurements and plasma discharges. It also shows the development of the control system being currently implemented and introduces HIDRA-materials analysis tool, the in situ PMI facility that will be mounted on HIDRA in the near future to further enhance the diagnostics and material testing experiments meant to be conducted on the machine.

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 accelerator-driven subcritical system (ADS) has potential economy and safety and relatively mature technology base, which has been chosen as a priority development advanced nuclear system. Liquid lead bismuth eutectic (LBE) is selected as the spallation target material and core coolant. The reduced activation ferritic/martensitic (RAFM) steel and the silicon carbide composite (SiC ${_{f}}$ ) are selected as its structural material and functional material, respectively. In the present experimental investigation, the special vacuum experimental device has been built, and the “dispensed droplet” modification of the classic sessile droplet technique has been used to investigate the wetting property and interfacial interactions for LBE/RAFM steel, LBE/316L, LBE/SiC, and LBE/SiC f couples. The contact angles were measured between the liquid LBE and the various candidate materials under working temperature from 300 °C to 480 °C. The results could provide meaningful compatibility database of liquid LBE and valuable engineering design reference of candidate structural materials and functional material for future ADS.

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.

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 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.

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