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EAST has been upgraded with top and bottom tungsten divertors to facilitate the high power long pulse operation. A visible spectroscopy diagnostic has been developed progressively to monitor W source as well as line emissions from other particles (e.g., D, He, Li, B, C, N, O, Ne, Ar, Mo, D2, WD) in the divertors. Two sets of lens installed in mid-plane port view tangentially the top and bottom divertors, respectively. Two types of optical fibers transfer the plasma light collected by the lens to three different detecting modules. High spectral resolution (λ/Δλ ∼ 14000) can be achieved by the detecting module consisting of spectrometer and Electron-Multiplying Charge Coupled Device (EMCCD), high temporal resolution (50 μs) is obtained by filtered photoelectric multiplier (PMT), and high spatial resolution (1 mm) is acquired by Complementary Metal-Oxide-Semiconductor Transistor (CMOS) camera with collimation optics. The multichannel optical fiber array with 500 μm core diameter is used to connect the lens and the first two detecting modules. The imaging fiber bundle with 30 μm core diameter transmits the two-dimensional divertor plasma light to the third detecting module, i.e. collimation optics and CMOS camera. The diagnostic is operated automatically in normal EAST discharges to support experimental operation and divertor physics study. The paper presents a technical description of the diagnostic and typical measurements during EAST discharges.

In tokamak devices, the performance of plasma facing components (PFCs) under high heat loads is a key challenge for achieving long-pulse and high-performance plasma operations. The lower divertor of EAST adopts an ITER-like cassette modules (CMs) structure. However, regular distribution of damage along the toroidal direction have been observed on the divertor target plates, which not only affects the lifespan of PFCs but also limits the long-pulse and high-performance plasma operations. Therefore, investigating the toroidal distribution of the heat load on the divertor target in EAST is crucial for understanding the mechanism of damage on the target plate surface. The toroidal characteristics of heat loads on CMs of the lower divertor in EAST was analyzed using a three-dimensional magnetic field line tracing program (PFCFlux). For the inner vertical target (IVT) and outer vertical target (OVT) within a single CM, the heat load varies monotonically along the toroidal direction. In contrast, the heat load distribution on the outer horizontal target (OHT) is uniform, except at the chamfers, which is directly correlated with the target’s geometric structure and the magnetic field configuration. The surface heat load distribution exhibits toroidal non-uniformity, with the heat deposition pattern being generally consistent and displaying periodic variations. The toroidally uneven heat load distribution is attributed to the change in the magnetic field’s tilt angle. Additionally, the peak heat load is observed at the chamfer positions between cassettes, and an analysis is conducted on the relationship between peak heat load and the size of the chamfer. Smaller tilt angles result in lower surface heat load. Furthermore, a comparative analysis of peak heat loads at the chamfer under different plasma current conditions reveals that higher plasma currents generally result in increased peak heat loads. This investigation into the toroidal heat load distribution on the divertor surface contributes to understanding of the damage mechanism of W PFC as the target plate and provides valuable data for divertor design of future fusion devices.

Heat flux is a key issue in tokamak devices. The non-uniform high heat flux on Plasma-Facing Components (PFCs) has led to local severe damage, including cracks and melting, in current tokamaks such as EAST and WEST. To characterize the non-uniform heat flux loading on the divertor surfaces, the parallel incident heat flux q‖, the decay length λq along the radial direction and the Gaussian spreading width S are used. The q‖ can lead to a very high peak heat flux loading on the divertor surfaces, which may cause critical heat flux problems. Additionally, the decay length is a key consideration for future tokamak designs like ITER. Every effort on the present tokamak devices contributes to updating the scaling of the heat flux. In EAST, a calculation method based on a high spatial resolution IR camera is employed to obtain the heat flux and decay length. The main process involves comparing the surface temperature distribution calculated by Fluent simulation with that measured by an infrared camera. Taking a high heating source discharge (#123059 ∼ 10 MW heating source) as an example, the heat flux is as follows: q‖=216-14+19 MW/m2, with λq=6.2-1.1+1 mm, and S=1.2±0.4 mm; it is in line with Langmuir probe data. The infrared-based heat flux calculation method can calculate the peak incident heat flux and the decay length simultaneously, its result can help to update the scaling model of heat flux, thus not only helping to improve the present device but also offering important reference for future tokamaks.

Pellet injection is the most promising fueling method to achieve a deep particle deposition for future fusion reactors. On metal-wall EAST tokamak, a full density compensation has been achieved by the pellet fueling during ELM suppression with n = 4 RMP which will be used in ITER’s operation. During the pellet injection, ELM suppression is terminated and small ELMs reappear. The plasma stored energy does not have a reduction with the pellet injection. Meanwhile, large sawtooth accompanied with RMP application is modified by the pellet injection. The amplitude of sawtooth crash decreases and high frequency small sawtooth crashes appear during the pellet fueling. The concentration of heavy impurities in the plasma core, such as tungsten and molybdenum, has an obvious reduction with the ELM reappearance and sawtooth modification. Although the same ELM behaviors also appears after divertor gas fueling, pellet fueling can modify the core plasma more significantly. All these experimental results will provide a good support to the ITER operation in the future.

Operation of EAST tokamak with full metal wall without any wall conditioning are attempted in 2023 and 2024 experimental campaign to address the issues related to ITER tungsten wall operation. It is found that H-mode plasma could be sustained even with a substantial increase in metallic impurity content caused by strong plasma-wall interaction under uncoated metal wall. Boronization wall conditioning is therefore performed to improve the plasma performance with higher injected power. This study aims to quantitatively assess the impact of boronization wall conditioning on metallic impurity concentration and behavior in the EAST tokamak. It is then proved to be an effective wall conditioning approach for significantly controlling high-Z impurity content. In this work, the impurity spectra at extreme ultraviolet (EUV) wavelength range measured by sets of fast-time-response and space-resolved EUV spectrometers are widely used in the data analysis. The variation in the boron content in plasma after boronization are investigated by monitoring the 2nd order of B V line at 48.59 Å. It is found that the persistence time of boron in EAST device after a boronization with 10 g of carborane (C2B10H12) is about 2000 s (∼150 shots) of discharge duration. The impact of different wall conditions (uncoated metal wall and boron coated wall) on metallic impurity content are then quantitatively studied. After boronization, the concentration of tungsten (CW) and molybdenum (CMo) dropped by 85 %, e. g., from 2.0 × 10-4 to 4.1 × 10-5 and from 4.6 × 10-5 to 6.3 × 10-6, respectively. While the concentration of copper (CCu) and iron (CFe) decreased by approximately 50 % and 65 %, respectively, e. g., from 4.3 × 10-5 to 2.1 × 10-5 and from 2.0 × 10-4 to 6.9 × 10-5. A comparison of the line emission profiles from tungsten ions of W26+ − W32+ and W43+ before and after boronization reveals an overall reduction in the intensity while without obvious change in the profile shape, which suggests a reduction in metallic impurities source after boronization instead of altering impurity transport in core plasma significantly.

W/Cu monoblocks will be employed in ITER as divertor targets for high heat loads. Experimental results of W/Cu monoblocks in high heat flux (HHF) tests have demonstrated their endurance on 5000 cycles at 10 MW/m2 and occasionally 300 cycles at 20 MW/m2. However, plasma-facing units (PFUs) serve in a harsh and complex environment in tokamaks and whether W/Cu monoblocks can show the same excellent properties in tokamaks needs to be further tested. Experimentally evaluating the evolution of ITER-like PFUs is one of the main missions for the EAST tokamak. The ITER-like actively cooled W/Cu monoblocks installed in the EAST lower divertor in 2021 exhibited abnormal behavior after 13,346 plasma discharges. The abnormal temperature evolution of the component was observed via infrared cameras (IR). Subsequent non-destructive testing (NDT) indicated that it might primarily arise from interface damage and scanning electron microscopy (SEM) confirmed that there were obvious cracks at the interface. The cracking center was also found above the interface. Simulation of the structural characteristics of W/Cu monoblocks under heat load in EAST can obtain temperature distribution and stress–strain distribution, which can be used to analyze the causes of material interface cracks. These findings unveil the phenomenon and mechanism of interface damage in W/Cu monoblocks after enduring over 10,000 plasma discharges in EAST, providing valuable insights for the future long-term application of W/Cu monoblocks in ITER and fusion reactors.

Ion cyclotron range of frequency (ICRF) heating system is an important auxiliary heating method in the experimental Advanced Superconducting Tokamak (EAST). In EAST, several megawatts of power are transmitted with coaxial transmission lines and coupled to the plasma. For the long pulse and high power operation of the ICRF waves heating system, it is very important to effectively control the power and initial phase of the ICRF signals. In this paper, a power and phase feedback control system is described based on field programmable gate array (FPGA) devices, which can realize complicated algorithms with the advantages of fast running and high reliability. The transmitted power and antenna phase are measured by a power and phase detector and digitized. The power and phase feedback control algorithms is designed to achieve the target power and antenna phase. The power feedback control system was tested on a dummy load and during plasma experiments. Test results confirm that the feedback control system can precisely control ICRF power and antenna phase and is robust during plasma variations.

To study material erosion and deposition during different wall conditioning processes in the Experimental Advanced Superconducting Tokamak (EAST), two newly developed quartz crystal microbalances (QMBs) at the mid-plane of port C (C-QMB) and port J (J-QMB) with the same radial radius were used for in-situ and real-time measurements during the 2021 autumn experimental campaign. It was found that the material deposition was mainly influenced by the location of the antenna and the power of the ion cyclotron range of frequency (ICRF), rather than the location of the gas inlet, during the wall cleaning with helium plasma and siliconization assisted with ICRF plasma. The higher energy and density of plasma could result in higher deposition rates near the antenna. While, during lithium (Li) coating, the material deposition was greatly influenced by the location of the Li oven due to the deposition of unionized Li atoms. The ICRF power influenced the material deposition behavior far away from the Li ovens. Besides, a stronger effect on impurity suppression was observed with a more uniform deposition trend of Li coating. This necessitates further study on the influence of factors such as the location of the Li oven and ICRF antenna, the temperature of ovens, and ICRF power on the homogeneity of wall conditioning.

In Experimental Advanced Superconducting Tokamak (EAST) experiments, to achieve long pulse and high-power ICRH system operation, a new kind of ICRH antenna has been designed. One of the most critical factors in limiting the operation of long pulse and high power is the intense heat load in the front face of the ICRH antenna, especially the Faraday Screen (FS). Therefore, the cooling channels of FS need to be designed. According to thermal-hydraulic analysis, the FS tubes are divided into several groups to achieve more excellent water cooling capability. The number of series and parallel tubes in one group is chosen as six. This antenna went into service in the spring of 2021, and it is delightful that the temperature distribution of the FS tube is below 400 °C in 14.5 s and 1.8 MW ICRH system operation. However, the active water-cooling design was not carried out on the upper and lower plates of FS, which led to severe ablations on that region under long pulse and high power operation, and the temperature is up to 800. Therefore, the upper and lower side plates of the FS were designed with water cooling based on thermal-hydraulic analysis. During the 2022 winter experiments, the temperature of ICRH antenna FS was lower than 400 in the pulse of 200s and the power of 1 MW operation.

In the twelfth and thirteenth centuries, after the First Crusade in 1095–1098 the eastern Mediterranean coastline was inhabited by Western European settlers, who formed a lordship known as the County of Tripoli in what is currently Lebanon and Syria. A number of architectural achievements were built during the creation of this county, and many of them remain from that period, including castles, towers, churches, etc. Besides the most known and important monuments, and the scatter amount of both historical and archaeological sources which indicate the existence of a variety of smaller structures in the area covering the northern governorate of Lebanon, it offers an ideal area for the analysis of the settlement pattern and typologies of the twelfth and thirteenth centuries. In this paper, we will only attempt to briefly describe some of the main points concerning the Crusader settlements in Northern Lebanon.

During the long-pulse operation of nuclear fusion devices, plasma-material interaction processes such as material erosion, transportation of eroded materials, and impurity redeposition can deteriorate the performance of plasma-facing components (PFCs). Accurately identifying impurities and their spatial distributions on the first wall is crucial to comprehensively and deep understanding these processes. Laser-induced breakdown spectroscopy (LIBS) is a powerful spectral analytical technique that offers rapid, in situ, multi-elemental, and portable instrumentation advantages over the other techniques. Due to its unique and distinctive features, LIBS is a practical technique for composition and profilometry analysis in nuclear fusion devices. In this study, we report the development of a novel, compact, and portable LIBS device for in situ, real-time measurement of impurity distributions in nuclear fusion devices. This device is equipped with a mini-sized laser that can deliver 10 mJ per pulse, a main unit enclosing a power supply module, spectrometer modules, and customized software. The system was employed to characterize impurities on the first wall of the EAST tokamak under atmospheric conditions after the 2021 EAST spring-summer campaign. The successful in situ measurement demonstrates the feasibility of using the portable LIBS system integrated into a robotic arm in nuclear fusion device. The implement of this portable device can facilitate the operation of the nuclear fusion device by enabling real-time and continuous monitoring of impurities as well as the fuel retention.

EAST is the first tokamak to feature fully actively water-cooled ITER-like W/Cu monoblocks on the divertor target. In 2014 and 2021, the graphite tiles in the upper and lower divertor were upgraded to W/Cu monoblocks, respectively. With the increase in plasma parameters, severe melting phenomena were inspected at the leading edges of upper divertor from 2017 to 2020. Through theoretical analysis and numerical simulation, two main factors, radial misalignment and chamfer structure have been identified to mainly impact the leading-edge induced melting of MBs. It was found that assembly misalignments increased gradually during the plasma discharges from 2015 to 2017 and the maximum misalignment can reach up to 3 mm, which play an important role that is responsible for the melting during 2018 to 2020. In 2020, engineering assembly improvements were implemented, resulting in a significant improvement compared to the previous situation, with all misalignments brought below 1 mm, and thus effectively prevent melting at the leading edge on upper divertor. In addition, the modification of chamfer structure to 1.5 × 17 mm at inter-CMs of the lower divertor further mitigated the issue of leading-edge-induced melting compared to upper divertor. Such experiences and lessons from the mitigation of leading edge induced melting provide important references for future fusion devices.

Plasma disruption is one of the most dangerous events which will directly influence operation safety of future large-scale fusion devices. During the thermal quench (TQ) stage, extremely high transient heat flux up to thousands of MWm−2 (in several ms) is deposited on the surface of plasma-facing components (PFCs), which will undoubtedly cause damage, namely, roughness, cracking, and even melting of the metal wall material, seriously shortening the lifetime of the PFCs. Based on temperature evolution measurement by a high temporal and spatial resolution IR diagnostic system in combination with the thermal calculation using ANSYS, the transient heat flux on the divertor during plasma disruption was investigated in the EAST. The maximum surface temperature rise during plasma disruption can reach up to 1000 °C in the case of stored energy ∼250 kJ, and the corresponding local peak heat flux during plasma disruption calculated to be more than several hundred MWm−2 (∼ms), which is possible to induce the damage to metal PFCs. The transient heat-flux-induced cracking and melting phenomena on the dome and baffle plates of the divertor, also illustrate that the local transient heat flux during plasma disruption in EAST may be significant and cannot be ignored. Moreover, the statistical analysis of plasma discharge parameters reveals that the maximum temperature rises and local transient heat flux on wall surface during the disruption phase generally increases with the increasing of energy storage in the core plasma. Thus, the EAST should pay attention to mitigating of plasma disruption in the future high parameter operations.

The impacts of neon injection positions and directions on the toroidally asymmetric heat load distributions on the divertor targets have been investigated for EAST tokamak by means of three-dimensional (3D) edge transport code EMC3-EIRENE. The heat loads on the outboard divertor targets are studied with five neon injection positions near the outboard strike point, which have different distances to separatrix. In comparison with neon injections away from the separatrix, neon injection close to the separatrix (within ∼ 1.2 cm) can lead to the asymmetric heat load profile on the outboard divertor targets. Furthermore, a parameter study of poloidal angle of neon injection has been conducted to estimate the effect of impurity injection direction on the suppression of heat loads. It is found that neon injection direction parallel to the magnetic surface results in the more effective mitigation of heat loads compared with other injection directions. Moreover, the neon accumulations in the core region are discussed for different neon injection positions and directions.

Plasma facing materials (PFMs) are subjected to long-duration high-energy particle streams and radiation in tokamak devices. The PFMs of EAST have been upgraded several times and Titanium-Zirconium-Molybdenum (TZM) tiles were installed into EAST as its first wall since 2011. However, with the gradually increasing of plasma parameters, several unexpected TZM melting phenomena were found at the high field side by post mortem inspection after each EAST plasma experimental campaign since 2017. The resolidified melted surface is general in wave shape with unobvious motion of melting layer. Three different grain shapes, i.e., columnar grain, isometric crystal and original rolled crystal from surface to deep region are found by means of metallurgical analysis, in which the superficial layer columnar grain is very thin with a thickness of 100 ∼ 200 μm and the thickness of intermediate isometric crystal is also small only about 300 ∼ 400 μm, strongly indicating there was a large temperature gradient near surface when melting occurred. Combined with plasma operation parameters and temperature evolution, the melting of TZM tiles were concluded to be induced by the transient heat flux during plasma disruption. These results imply the transient heat flux during plasma disruption in EAST can severely destruct the metal PFMs and should not be ignored, suggesting the active mitigation of plasma disruption is necessary for future long pulse and high parameters operation.

Laser-induced breakdown spectroscopy (LIBS) was employed for the depth-resolved identification of impurities deposited on the small test tiles used in the experimental advanced superconducting tokamak (EAST). LIBS spectra show the impurity elements of molybdenum (Mo), tungsten (W), carbon (C), copper (Cu), lithium (Li), titanium (Ti), silicon (Si), iron (Fe), and chromium (Cr) in the impurity deposition on the test tiles used in the EAST. The analysis of impurities was performed at various spot positions on the surfaces of the tiles. The impurity deposition is the final destination of the migrating eroded materials in the EAST. The interaction of high-heat plasma with the plasma-facing components (PFCs) causes the erosion of W upper and lower divertors, Mo main wall, the doped graphite (C, Si & Ti) limiters, Li wall conditioning, and CuCrZr heat conductors. Surface morphology was performed by using the scanning electron microscope (SEM) and energy-dispersive X-ray spectroscopy (EDX). Depth profile analysis showed the thickness of impurity deposition is in fractions of micrometers because the impurity signals appeared just in a few laser shots. The spectral intensity of impurity was observed differently during the consecutive laser shots on various spot positions that showed the impurity deposition is uneven on the tiles’ surfaces.

Siliconization using 10 % SiD4 + 90 % He assisted by ion cyclotron range of frequency discharge (ICRF) or glow discharge (GD) was performed in EAST with full tungsten (W) diverters. It was found that the coated Si film on W sample at a higher baking temperature (160 °C) was smoother than that at a lower baking temperature (60 °C). This was mainly because physically adsorbed gaseous impurities were released at a higher baker temperature. This process improved the adhesion between the film and the surface of W. An increase in the ICRF working power from 20 kW to 40 kW further increased the Si content and the film thickness by 1.5 times. The cleaning efficiency of ICRF on the surface of the W sample before siliconization was higher at 40 kW than that at 20 kW, which facilitated the removal of oxides and other compounds from the surface of W. The ratio of silicon/oxygen (Si/O) and the thickness improved considerably due to greater ionization and deposition of SiD4. Additionally, siliconization via GD was more uniform and 1.5 nm thicker than that via ICRF, which was because of the greater working gas pressure, coating duration, and homogeneity in GD. These results might be used as a reference for evaluating the effect of siliconization on plasma performance and its application in fusion devices.

During the tokamak plasma discharge, the strong interaction between the edge plasma and the wall material leads to the overheating of the surface of divertor target plate and the possible melting of the material, which will seriously influence the safe operation of the device. Furthermore, transient heat load induced by edge localized modes (ELMs) is expected to cause surface damage of tungsten (W) based plasma-facing materials (PFMs) in future fusion devices. In order to ensure safe steady-state operation, it is prerequisite to obtain the peak temperature as well as distribution on the divertor target plate in real time. In EAST, a high temporal and spatial resolution infrared camera (IR) located in H port was built to observe the outer target on the L port of upper W divertor. Currently, the maximum temperature measured by IR camera has been characterized and analyzed during deuterium (D) and helium (He) plasma discharges. In addition, the ELM-induced temperature rise during H mode operations for D and He plasma discharges was specially studied. It is found that the maximum temperature on observed targets during both D and He discharges increase with the increase of heating power. In case of the same heating power, the maximum temperature during D discharges is generally lower than that during He discharges no matter L and H modes, which is more obvious when the total heating power exceeds 3 MW. However, the sharp temperature rise by ELM induced transient heat load during D discharges is usually higher than that during He discharges. Such qualitative and statistical analysis on the maximum temperature on upper outer divertor provides important reference for future operation of EAST and other tokamaks.

Real-time wall conditioning and edge-localized mode (ELM) suppression through real-time boron powder injection into high-confinement mode discharges of the Experimental Advanced Superconducting Tokamak (EAST) have been investigated. The boron powder injection effectively conditioned the plasma-facing components, as shown by the reductions in low-Z and high-Z impurities and the global recycling level in the whole inner vessel, and suppressed the strong magnetohydrodynamic (MHD) activity in the plasma core. The improved wall conditions also resulted in an earlier transition from L- to H-mode. Boron powder injection also effectively suppressed ELM in the lower single null configuration with BT in the favorable direction while leaving the plasma density and stored energy largely unaffected.

Fuel recycling control in long pulse discharges is studied in EAST tokamak with tungsten divertors by particle balance. Double null configuration, proper Resonant Magnetic Perturbation (RMP), divertor cryopumps and realtime lithium injection, are all very effective for recycling control. High heating power discharge leads to low recycling due to the enhancement of particle exhausting. High external gas injection is required in a low fuel recycling discharge to keep a constant density, and Dα emission is increased by external injection, therefore a higher Dα emission may indicate low fuel recycling for the same plasma density. Reproducible long pulse discharges of over 100 s with core electron temperature >10 keV, and 1056 s discharge with ∼1.5 MW heating power were obtained in the EAST 2021 campaign, fuel recycling was well controlled by divertor cryopumps and realtime lithium injection. The results provide a valuable reference for the high power and long pulse operation in future fusion reactors, such as ITER and CFETR.

Arcing on the plasma-facing components (PFCs) is one of the mechanisms of material erosion and impurity production in the magnetic-confinement fusion devices. EAST tokamak has been upgraded to almost full metal PFCs. The arcing phenomena in EAST has special features due to the lithium coating and the RF heating. Post-mortem analysis of arc traces is performed on both plasma-wetted components and PFCs in sub-limiter remote areas. The arcs in sub-limiter remote areas nearly have no influence on the normal discharges as most of them were produced during the glow discharge cleaning. But some arcs on the first mirrors of infrared and visible tangential wide-angle viewing systems in EAST can form patterned traces in the videos, which affects the precision of temperature measurements and visible imaging quality. The arcs on plasma-wetted components were mainly observed at two locations: the tungsten upper divertor and the lower hybrid current drive (LHCD) antennas. The random-walk arc traces on the surface of the upper tungsten divertor are limited to a very small area and most probably formed without magnetic field during glow discharge, unable to influence plasma discharges so far, and still need long-term tracking. The arcs on the LHCD antenna surface can be sustained for more than hundreds of milliseconds, and produce large numbers of impurity particles and molten droplets, which can result in plasma current quench and disruption, and the effects of arcing on LHCD heating mode and power should be further studied.

Background and Aims: Previous large studies on the epidemiology and clinical profile of chronic pancreatitis have suggested significant differences in presentation and management in the eastern and western hemispheres. The aim of this systematic review was to compare large multicenter studies across different geographic regions of the world to detect meaningful differences in the presentation and management of this poorly understood disease. Methods: We identified 237 manuscripts through a comprehensive literature review aiming to identify multicenter studies enrolling more than 200 patients to limit reporting biases. After rigorous screening, 12 studies were included for the final analysis. The Asian studies were included in the eastern cohort, and the European and American studies were included in the western cohort. Reported demographics, risk factors, etiologies, clinical presentation, complications, and management strategies were then compared. Results: We found similar demographics across both cohorts including age, prevalence among gender, and predominant etiology. Clinical manifestations including pain, pancreatic calcifications, and diabetes were similar between both cohorts although pseudocysts, pancreatic cancer, and strictures were more common in the west. Notably opioids and surgical/endotherapy management were more common in the west as well. Conclusion: Chronic pancreatitis is a protracted disease affecting predominantly middle-aged people, leading to a decreased quality of life. Chronic pancreatitis now appears to have a fairly similar clinical profile and natural history in the east and west. There is notable variability in management. We hope that international collaboration may identify common targets for research which could lead to significant advances in the understanding and management of chronic pancreatitis.

Dust has been collected and characterized by various methods during the 2021 first experimental campaign in EAST tokamak. It was found that the weight of the in-vessel dust collected at the mid-plane port varied from about 265 mg to about 966 mg. And the dust appeared mainly in the shape of spheroids, flakes, stripes and irregular-shaped particles with the sizes ranging from micrometers to several millimeters. The dust consisted of different elements, which was from the wall-conditioning, erosion of plasma-facing materials (PFMs) as well as the flaking of deposits. The dust collected from the lower port K contained different elements from those collected from the in-vessel port K, with more Fe in the lower port and more Li in the vacuum vessel. The ferromagnetic dust with different shape has been found, and different shape has different O content and size. By using the diagnostic observation window, dust was firstly collected during the campaign and the local dust production rate is calculated, which is found to decrease as the plasma operation goes.

Particle control is one of the key issues for steady-state tokamak operation. The density decay time is used to evaluate the capability of the divertor to exhaust particle with all fueling turn off. Experiments of the EAST divertor's ability to exhaust helium particles were also performed during the EAST helium plasma operation. The experimental results show that the EAST divertor is less capable of exhaust helium particles than deuterium particles. And similar to deuterium, the helium particle exhaust performance is sensitivity on the different strike point location. During 2021 experiment campaign, a new lower tungsten divertor has been developed and installed on EAST, the particle exhaust capacity of the new divertor has also been investigated during this campaign.

Physical sputtering caused by particle bombardment is believed to be the main erosion mechanism of W materials in fusion devices, in which W atoms are the sputtering products. However, the tungsten deuteride molecule (WD) spectra have been observed in both TEXTOR and ASDEX Upgrade, which was believed to be the product of chemically assisted physical sputtering (CAPS), a new sputtering mechanism that has been proposed in recent years. In this paper, we report the spectroscopic observation of WD molecules in the EAST W divertor. The behaviors of WD molecules sputtering are compared with W atoms sputtering via the spectral measurements of the ro-vibrational band emission of WD 6Π → 6Σ+ in the spectral range between 673 nm and 678 nm and the WI line emission at 400.9 nm. The physical sputtering characters and chemical sputtering characters of WD molecule sputtering were confirmed in EAST. The measurements in EAST showed that there is an energy threshold for WD molecules sputtering and that the sputtering energy threshold of WD molecules is smaller than that of W atoms. Furthermore, the dependence of WD molecule sputtering efficiency (the absolute WD photon flux (PhWD) normalized to the particle flux (Γion) reaching the target with a constant Te) on impact energy and heat flux were studied, presenting significantly different behaviors compared with W atom sputtering. In addition, it is found that the decrease of WD sputtering efficiency with the heat flux is accompanied by the rise of the photon flux of Dδ (410.06 nm) normalized to the particle flux (Γion) at the divertor target, which may imply the enhanced deuterium desorption at the W surface. The further increase of heat flux hitting the target surface could elevate the surface temperature and benefit the deuterium release from the surface, which may influence the formation of WD molecules at the surface layer.

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.

Fusion energy is a promising source of clean energy, which could solve energy shortages and environmental pollution. Research into controlled fusion energy has been ongoing for over half a century. China has created a clear roadmap for magnetic confinement fusion development, where superconducting tokamaks will be used in commercial fusion reactors. The Experimental Advanced Superconducting Tokamak (EAST) is the world’s first fully superconducting tokamak with upper and lower divertors, which aims at long-pulse, steady-state, H-mode operation, and 101-s H-mode discharge had been achieved. In 2007, China joined the International Thermonuclear Experimental Reactor (ITER) and became one of its seven members. Thirteen procurement packages are undertaken by China, covering superconducting magnets, power supplies, plasma-facing components (PFCs), diagnostics, etc. To bridge the gap between the ITER and fusion demonstration power plants (DEMOs), China is planning to build the Chinese Fusion Engineering Testing Reactor (CFETR) to demonstrate related technologies and physics models. The engineering design of the CFETR was completed in 2020, and Comprehensive Research Facilities for Fusion Technology (CRAFT) are being constructed to explore the key technologies used in the CFETR.

Comparison of defect formation in tungsten samples exposed only to a helium plasma and those exposed to both helium plasma and subsequent deuterium plasmas is performed via Doppler broadening spectroscopy in the positron annihilation technique (DBS-PA). The S parameters of tungsten samples after helium plasma exposure increased with the increase of the incident fluence, which indicates that low energy helium plasma could induce the formation of defects such as nano-sized helium bubbles and dislocations in tungsten. The S parameters of samples exposed to helium incident fluence of 7.9 × 1021 He/m2 enhanced after deuterium plasmas exposure, indicating that low energy deuterium plasmas could form defects in tungsten too. Unexpectedly, deuterium plasmas exposure did not significantly influence the S parameters of the tungsten sample with helium incident fluence of 3.9 × 1022 He/m2, and reduced the S parameters of the tungsten sample with helium incident fluence of 7.9 × 1022 He/m2. These variations of S parameters could be associated with the occupation of nano-sized helium bubbles by deuterium atoms. This indicates possible competition between the increased density of vacancy-type defects by deuterium plasmas exposure and the filling of nano-sized helium bubbles by deuterium atoms, reducing detectability of these defects.

The application of ITER-like W/Cu plasma facing components in current tokamaks is of special concern by ITER. In 2014, ASIPP successfully upgraded upper divertor of EAST to a full tungsten component structure through independent research and development, and adopted W/Cu monoblocks as the divertor targets for high power exhaust. During recent plasma operations, the leading edge-induced thermal loading caused the melting of W/Cu monoblocks and the melt layer occur motion mainly by electromagnetic force. It was found that the melt layer migrated to neighboring W/Cu monoblocks, solidified into a hill structure or filled the gap between W/Cu monoblocks. Meanwhile, the large sized chamfering structure was residual on the molten W/Cu monoblocks. The thermal analysis shows that the W /Cu monoblcoks with large sized chamfering structure has an increased heat load capacity and it is not prone to secondary melting with the same heat flux in subsequent operations. And the removed molten layer may gradually migrate far away from the strike point and eventually solidify into stable structures. The improvement of the heat load capacity of the molten W/Cu monoblock and the melt-layer motion may be a benign situation, thereby achieving a healing effect to avoid further deterioration of plasma facing components. The evolution behavior of such leading edge-induced melting on ITER-like W/Cu monoblocks in EAST provides significant reference for ITER.

The heat flux deposition on the divertor targets with neon impurity injection has been investigated for EAST tokamak with the help of three-dimensional (3D) edge transport code EMC3-EIRENE. The parameter studies of the upstream electron density, neon impurity injection flux and injected neon atom energy have been conducted to check their impacts on the extent of the toroidal asymmetry of heat load distributions. It is found that high upstream electron density leads to an effective mitigation of heat flux deposition and more significant toroidal asymmetry of heat load distribution. Further, the toroidal asymmetry of heat load distribution increases with neon impurity injection flux. However, a weak dependence of heat load distribution on the neon impurity injection energy is obtained.

Energy dissipation in the plasma edge is key for future tokamaks. The potential of neon as radiating seeding species in disconnected double null (DDN) configuration is assessed in EAST discharges in high confinement mode (H-mode). As the separation between the two separatrices in the studied DDN discharges is minimum 1.5 cm, the configuration is effectively a single null configuration, and the benefits of the double null topology are minimal. Neon seeding, on the other hand, has a favourable effect: both the target heat flux and the divertor temperature decrease more than five-fold with increased seeding rate in high-recycling conditions. Interpretive edge plasma simulations with SOLPS-ITER in support of ongoing transport analysis are presented. For the unseeded case the numerical results agree with the experimental data within a factor two for the target temperature conditions and measured neutral pressures in the active divertor. The key for achieving good agreement is a suitable selection of coefficients for anomalous transport and neutral conductances between the upper cryopump and the main chamber.

Rare isotope tracer injection represents an established way for following particle transport in fusion research, but a tracer experiment is only as good as the tracer can be followed and separated from the common isotope. Physical models for transport analysis such as ERO2.0 require accurate total amounts of deposition with high spatial resolution. This work describes an extended high throughput NRA post analysis approach together with the surface analysis results of 2 technically rough graphite-tiles of 150x150 mm2 surface area exposed to D and He plasmas in EAST.The applied 1.43 MeV deuteron beam NRA offers percent range accuracy for layer thicknesses ranging from few 10 nm to 20 µm. Optimisation of the beam line and beam optics enables operation with ≤20 nA without reaching critical radiation levels, resulting in a measurement time of 5–10 s per spot. This enabled acquiring in total 3070 NRA points on the central 46 × 46 mm2 cut-outs of each sample and an adjacent part of equal size within 12 h of analysis time.The layers contained 50–60% D, ~2% 12C, and ~3% O. The central 20 mm radius around the injection hole contains 98% of the total 13C deposition. NRA finds up to 5.8 ± 0.9 × 1022 13C/m2 close to the hole with an e-folding fall-off length of about 3–10 mm in every radial direction. In the D discharge 24 ± 2% and 16 ± 2% in He of the injected 13C are found on the tiles with triangle shaped deposition profiles following ExB. The comparison to colour fringe analysis and SIMS reveals the particular strength of D-NRA regarding roughness, thick layers, and the separation of 12C and 13C deposits, but in compatible regions all methods agree. Compared to these, the NRA yields additional information indicating depositing species and physical processes.

The behavior of ITER-like W/Cu plasma facing components under complex conditions in current tokamaks is one of the main concerns for ITER. EAST has installed a full upper W divertor with the ITER-like W/Cu monoblocks as targets since 2014. A melting failure of CuCrZr cooling tube of W/Cu monoblocks has occurred during the plasma campaigns in 2019. Due to the loss of cooling water, the leading edge-induced thermal loading can lead to the melting of both W armor and CuCrZr cooling tube, which has been confirmed by thermal simulation and analysis. With movement and migration of the melted Cu through gaps, the structure and function of the CuCrZr cooling tube for W/Cu monoblocks was severely destroyed. As a result, those failed W/Cu monoblocks had to be replaced. Such case in EAST just simulates the extreme condition of accidental loss of coolant in future devices. The melting failure of CuCrZr cooling tube was caused by operation without coolant water at largely misaligned monoblocks, which is a key lesson referenced to other tokamaks which adopt or plan to apply such type of W/Cu monoblocks.

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.

Objectives: Focusing on the East, Central, and Southern African region, this study examines both regional and country-level initiatives aimed at promoting multisectoral collaboration to improve population health and the methods for their economic evaluation., Methods: We explored the interventions that necessitate cooperation among policymakers from diverse sectors and the mechanisms that facilitate effective collaboration and coordination across these sectors. To gain insights into the demand for multisectoral collaboration in the East, Central, and Southern African region, we presented 3 country briefs, highlighting policy areas and initiatives that have successfully incorporated health-promoting actions from outside the health sector in Zimbabwe, Uganda, and Malawi. Additionally, we showcased initiatives undertaken by the Ministry of Health in each country to foster coordination with national and international stakeholders, along with existing coordination mechanisms established for intersectoral collaboration. Drawing on these examples, we identified the primary challenges in the economic evaluation of multisectoral programs aimed at improving health in the region., Results: We illustrated how decision making in reality differs from the traditional single-sector and single-decision-maker perspective commonly used in cost-effectiveness analyses. To ensure economic evaluations can inform decision making in diverse settings and facilitate regional collaboration, we highlighted 3 fundamental principles: identifying policy objectives, defining the perspective of the analysis, and considering opportunity costs. We emphasized the importance of adopting a flexible and context-specific approach to economic evaluation., Conclusions: Through this work, we contribute to bridging the gap between theory and practice in the context of intersectoral activities aimed at improving health outcomes., Competing Interests: Author Disclosures The authors reported no conflicts of interest., (Copyright © 2023 International Society for Health Economics and Outcomes Research. Published by Elsevier Inc. All rights reserved.)

Background & Aims: Video capsule endoscopy (VCE) is the first-line diagnostic procedure for investigating obscure gastrointestinal bleeding (OGIB). Different re-bleeding rates following index VCE have been reported among Western and Eastern studies., Methods: We conducted a comprehensive literature search to identify studies examining re-bleeding rates after VCE for OGIB. Meta-analysis assessed the pooled proportion of re-bleeding events after VCE for OGIB according to study's origin (Western vs. Eastern) and according to the length of follow-up (≥24 months vs. <24 months). We also calculated the re-bleeding odds ratios (OR; 95% CI) after positive vs. negative index VCE, overt vs. occult initial presentation of bleeding and after interventional treatment for positive index cases, according to the study's origin., Results: We included 46 (30 Western and 16 Eastern) studies with 5796 patients. Significant heterogeneity was detected among meta-analyzed studies. Overall, the pooled re-bleeding rate was similar between Western (29%; 95% CI: 23-34) and Eastern (21%; 95% CI: 15-27) populations, irrespective of the length of follow-up. The odds of re-bleeding was significantly higher after positive as compared to negative index VCE in Eastern studies (OR: 1.77; 95% CI: 1.07-2.94). Application of specific treatment after positive index VCE was associated with lower re-bleeding odds in both Western (OR: 0.37; 95% CI: 0.16-0.87) and Eastern (OR: 0.39; 95% CI: 0.21-0.72) populations., Conclusions: Patients undergoing VCE for OGIB have similar re-bleeding rates in the East and the West, regardless of the length of follow-up. However, increased re-bleeding odds after positive index VCE is observed in Eastern studies., (Copyright © 2018 Editrice Gastroenterologica Italiana S.r.l. Published by Elsevier Ltd. All rights reserved.)

Objectives: The aim of this study was to assess the impact on practice of vena cava filter insertion guidelines (Eastern Association for the Surgery of Trauma: practice management guidelines)., Design: The study was performed at a level 1 trauma centre with data from the 'Trauma Audit and Research Network' cross-referenced to hospital data., Results: A total of 1138 specific 'high-risk' major trauma patients were identified over a 6-year period. The mean age was 46 years (18-102) and the male to female ratio was 3.3:1. The average Injury Severity Score was 23.6 (4-75). The overall DVT rate was 2.6% and the PE rate was 1.8%. A retrievable IVC filter was inserted in 42 cases (3.8%). The filter retrieval rate was 23.8% at a mean of 68.5days (4-107). Only one complication was reported of a breakthrough PE despite filter. Applying the EAST guidelines to this cohort would have suggested filter insertion in 279 (24.6%) cases. The kappa concordance value between observed practice and the 'EAST filter group' was 0.103 (poor). The PE rate in the 'EAST filter group' was 2.2% vs 1.6% in the 'no filter group' (p=0.601, no statistical difference) and the observed odds ratio was 0.814 (95% CI 0.413, 1.602)., Conclusion: The EAST guidelines are useful but may be overestimating the need for filter insertion., (Copyright © 2017 Elsevier Ltd. All rights reserved.)