W/Cu flat-type components are promising candidates for plasma-facing components in future fusion devices due to their flexible and efficient heat sink design. They are expected to be used in the inner/outer Reflector plates and Dome of ITER. To investigate their damage behaviors, a high-resolution infrared camera was employed to monitor the damaged C4 W/Cu flat-type component with a hypervapotron heat sink, located in the outer horizontal target (OHT) of the EAST lower divertor. The results show that the formation of W protrusions (‘Hills’) connects the early crack initiation stage with the final W exfoliation. To prevent further damage progression, it is recommended that for tokamaks, the damaged area be located two decay lengths away from the strike point—this distance is sufficient to guarantee safety. In such a scenario, if the damage lies within the private flux region, the parallel heat flux will be reduced to nearly zero. Although thermal performance degraded over time, the damaged OHT component was still able to maintain an approximate thermal equilibrium during each discharge, and no abnormal W impurity levels were observed. Linear fitting analysis estimates a total lifetime of approximately 3257 ± 277 shots, with a plasma stored energy of 298 ± 20 kJ corresponding to the onset of W melting in the surface protrusions. These findings shed light on the damage behavior of W/Cu flat-type components, provide key support for the development and validation of finite element analysis models, and offer practical guidance for the design and evaluation of ITER-like W/Cu flat-type components.

For ITER and BEST, it is essential to identify potential thermal fatigue damage of plasma-facing components (PFCs) and to evaluate their impact on plasma discharges before D–T nuclear fusion reaction. In 2021, ITER-like W/Cu monoblocks featuring a large chamfer (17 mm × 1.5 mm) were installed on the lower divertor target of EAST. The inclined structural design at the inter-gap of W/Cu monoblocks can reduce the probability of leading edge melting but may induce other types of damage due to a more steady-state heat flux loading on the surface, similar to the fish-scale surface of the divertor in ITER. As plasma discharges progressed, a strange phenomenon was found that bright hot spots indicative of localized thermal anomalies were frequently observed by infrared (IR) cameras, affecting approximately 41% of the modules along the toroidal direction. Subsequent investigations confirmed these anomalies primarily originated from interface damage, which significantly degraded heat transfer performance in the PFCs. IR thermography measurements combined with thermal simulations showed that interfacial heat transfer coefficient decreased to 2 × 10 ^⁴ W (m ^2 ·K) ^−1 after approximately 4800 plasma discharges, resulting in an estimated 18% increase in the steady-state surface temperature. With continued operation through autumn 2022, the interfacial heat transfer conductance further decreased to 5 × 10 ^3 W (m ^2 ·K) ^−1 , leading to a 31% increase in surface temperature. The progression of interface cracking led to an abnormal temperature rise and thus facilitated incipient melting at the leading edges of the cassette modules. As the heating power continues to rise above 10 MW, melting is expected to occur more extensively along the 17 mm × 1.5 mm chamfer. Fortunately, before the onset of melting, the existing interface cracks within W/Cu monoblocks had negligible influence on plasma discharges. These findings provide critical reference data for the long-term application of W/Cu monoblocks in ITER and future fusion reactors.

The tungsten-copper (W/Cu) flat-type component is a promising candidate for plasma-facing components, praised for its flexible heat sink design and cost-effective fabrication. It is considered for the ITER divertor dome section and holds potential for divertor targets for future fusion devices. Since 2021, three types of ITER-like W/ Cu flat-type components with chamfer angles of 2.40 degrees, 3.76 degrees, and 5.19 degrees have been installed and tested on the outer horizontal target of the lower divertor in EAST. During plasma operations, some W/Cu flat-type components experienced cracking, melting and exfoliation of tungsten, causing plasma disruptions and even the termination of experiments. To address this, this study uses Fluent to evaluate the melting thresholds of these components based on simulations that consider actual conditions. The key findings are summarized: Firstly, only certain W plates can reach high temperatures because the in-situ heat flux is localized, explaining why damage often occurs only on some specific plates. During leading-edge events, the highest temperatures for both W and oxygen-free copper (OFC) occur near the sides, but W's maximum temperature is at the edge, while OFC's is near the edge. The high-temperature area for OFC is larger than that for W, resulting in a more uniform temperature distribution. Secondly, for components with chamfer angles of 3.76 degrees and 5.19 degrees, the heat flux q0 required for OFC to melt is always lower than that for W, meaning OFC melts first. For the 2.40 degrees chamfer, OFC melts first when the incident heat flux angle exceeds 2.5 degrees, while W melts first below this angle. At a 10 MW total heating power, all three components face a melting risk at the maximum incident angle of 5 degrees during EAST operations. Thirdly, a linear relationship was identified between the maximum temperatures of W and OFC under such complex loading conditions, offering a method for monitoring OFC maximum temperature based on the IR measurement of surface W temperature. Moreover, this linear relationship can be extended to other heat sink materials. These results provide valuable guidance for EAST's plasma operations and offer reference data for the use of W/Cu flattype components in next-generation devices like ITER and CFETR.

In future ITER operations, it is crucial to maintain the performance of plasma-facing components, particularly in regions subject to high thermal loads. Studies on component damage in existing tokamaks provide important reference data for predicting potential damage in ITER and future fusion devices, ensuring the most effective response strategies during future operations. In 2021, ITER-like W/Cu monoblocks featuring a large chamfer (1.5 mm × 17 mm) were installed on the lower divertor target of EAST. The structural design of W/Cu monoblocks with inclination angles can reduce the probability of melting but may also induce other types of damage, similar to the fish-scale surface of the divertor in ITER (toroidal bevel with a depth of 0.5 mm). After three plasma campaigns, post-mortem analysis identified significant surface damages, including crust formation, increased surface roughness, macrocracks, and other types of damage. The spatial distribution of damage is strongly correlated with the heat flux distribution and the degree of misalignment. At the leading edge areas, macrocracks are caused by plastic strain under steady-state heat loads, while microcracks are brittle cracks mainly caused by transient heat loads. The net-like cracks displayed four distinct morphological characteristics, attributed to variations in heat load and incident angles. Abnormal grain growth was observed in the vicinity of the melted components, with grain sizes reaching an extraordinary 7.1 mm. Both the formation and propagation of cracks were related to material degradation, such as the reduction of mechanical strength, the decrease in fracture toughness, and increased embrittlement. Molten tungsten (W) and low-Z carbon impurities can react to form tungsten carbide (W _2 C) at extremely high temperatures. Such damages on W/Cu monoblocks for the lower divertor in EAST provide critical insights into the service performance of these ITER-like W/Cu monoblocks, which can provide important reference data for the future use in ITER and other fusion reactor devices.

The heat load distribution on the first wall(FW)and the wall material erosion have been investigated for the CFEDR conventional H-mode scenario.The FW shaping has been optimized based on the 3D assessment of the heat flux distribution for both the start-up and steady-state phases using the PFCFlux code.This optimization ensures that the leading edge is protected even with 10 mm misalignment considered for each wall segment.During the start-up phase,the peak heat load is about 0.85 MW/m2 located at the high-field side wall,whereas for the steady-state phase,the peak heat load is about 0.64 MW/m2 located at the low-field side wall,with the main contribution from the loss of α(W)edge transport have been simulated using the SOLPS-OSM-DIVIMP code package.Simulation results confirm that the W source from the FW can significantly increase the core W density by more than an order of magnitude for the detached divertor conditions,thereby determining the core W concentration.A suitable D2 injection as a trade-off for Ar injection has been demonstrated to effectively reduce W erosion rates while maintaining the core W concentration and material lifetime at an acceptable level for CFEDR.

Tungsten is a promising candidate material for plasma-facing components in future fusion reactors. An important issue is the irradiation-induced degradation of its mechanical properties and its typically superior thermal conductivity. In this study, tungsten was irradiated with 270 keV He+ to the damage levels of 0.7 dpa and 3.0 dpa at 500 degrees C and 800 degrees C. The overall distribution of the microstructure is observed and its evolutionary relationship with the micromechanics property is discussed. The study of the microstructure reveals that the increase in He (+) ion fluence leads to an increase in the density of bubbles, which conversely decreases with elevating temperature. Both 1/2 <111> and <100> loops have been identified in the current study, and 1/2 <111> loops will gradually transform into <100> loops as the temperature rises. The synergistic interaction between He bubbles and dislocation loops results in irradiation hardening, with the contribution of dislocation loops exceeding that of He bubbles. The increased presence of <100> loops at elevated temperatures further contributes to additional hardening increments. These findings help to understand the influence of bubble evolution and irradiation hardening behavior in tungsten, especially the contribution of different types of defects to hardening, and thus help to design new radiation resistant PFMs.

The synergistic effect of neutron and plasma on tungsten was simulated by heavy ion and plasma irradiation. Here we experimentally reveal the evolution of dislocation rings irradiated by different types of ions at room temperature (RT) and their effects on hardening in tungsten. This was achieved through irradiation using a sequential threebeam sequence of He+Fe+D, with single-beam He and Fe, as well as dual-beam He+Fe setups serving as controls. As the irradiation dose increased, the number density and average size of dislocation loops notably increased for single Fe ions. Significant irradiation hardening was observed in the samples with dual-ion irradiation due to the higher number density and larger size of the dislocation loop. Dislocation loops with b = 1/2<111> and b = <100> were observed in both the samples pre-implanted with He, and the Fe and D ion pro-implantation. D ion and the pre-implantation of He changes a <100> dominated defect population to a 1/2 <111> dominated one. Following D plasma exposure, calculations using the DBH model indicated an increase in nanohardness. However, the irradiation-induced hardening, assessed using the Nix-Gao model, revealed that D ion implantation impeded this increase in nanohardness. This discrepancy can be attributed to variations in the number density and average size of dislocation loops, as well as the synergistic effects involving Burgers vectors. Keywords: Synergistic effect; Dislocation loop; Burgers vector; Irradiation hardening; Corresponding author.

A hyperspectral camera (HSC-type Specim IQ) has been applied at the linear plasma device PSI-2 under steady-state conditions. The camera has the capacity of hyperspectral imaging (HSI) with the dimension of a data array 512 × 512 × 204 (x, y, λ) covering the spectral span from 400 to 1000 nm with moderate average spectral resolution (FWHM ∼7 nm). After radiometric calibration and background/continuum emission subtraction, two main applications of the camera, (i) plasma diagnostics in helium (He) plasmas and (ii) plasma-material interaction studies with tungsten (W) targets in neon (Ne) plasmas, have been carried out. The measurements were complemented by a movable Langmuir double probe system (LP) measuring electron temperature (Te) and electron density (ne) in radial direction r and a fiber-coupled cross-dispersion spectrometer with high spectral resolution (Spectrelle) recording neutral He, W, and Ne emission lines over the full plasma column. (i) Two-dimensional (2D) imaging of Te and ne radial profiles in axial direction z of the He plasma column were for the first time obtained by the regression analysis of Te and ne (from LP) and six He I line ratios (from HSC). The spatially resolved plasma parameters covered in these studies range between Te ∼ 0.8-13.4 eV and ne ∼ 0.2 × 1018-3.9 × 1018 m-3 and permit a reconstruction of the plasma conditions in PSI-2 in 2D without LP perturbation. (ii) W sputtering was studied in situ in Ne plasmas exposing W target samples (negatively biased at 100 V) under perpendicular Ne plasma impact. Simultaneously, the 2D distributions of W (W I line at 429.5 nm) in front of the target and the 2D Ne plasma distribution (Ne I line at 703.2 nm) were recorded with complete spectral separation as confirmed by the Spectrelle spectrometer. This permits the simultaneous measurement of the neutral W penetration and its angular distribution induced in the sputtering process and of the impinging plasma distribution. The HSI technique offers, despite a few technical drawbacks, such as the moderate spectral resolution and poor time resolution, a new possibility to distinguish multiple emission lines from plasma and impurities and complements the portfolio of existing Optical Emission Spectroscopy techniques, providing a good compromise regarding spectral, spatial, and temporal resolution.

Titanium-Zirconium-Molybdenum (TZM) is applied as the first wall material of EAST due to its excellent performance. To investigate the service behavior of TZM under transient heat fluxes, the postmortem inspection of in-situ damaged TZM tiles by plasma disruption heat load in EAST was analyzed. Meanwhile, the EMS-60 electron gun has been also used to test the transient heat load behaviors of TZM. There are some similar characteristics between the in-situ melting TZM and the ex-situ melting TZM tiles, but there are also some obvious differences. Specifically, in some melting zone, the surface morphology of partial in-situ TZM melting region is like that of ex-situ TZM melting areas, and their convex surface is relatively smooth, while the concave surface is relatively rough, and the solidification layer undergoes recrystallization. In other melting areas, there are many small pits on the surface of the TZM tile that was melted in-situ, but there are no obvious pits on the surface of the ex-situ TZM sample, and part of the surface of the TZM tile that was melted in-situ is like a big pit, and the melted TZM droplets in this melting area tend to splash towards the periphery. The capillary waves around the castle structure in the ex-situ test samples are the result of wetting effect of liquid metal and capillary force, and the holes in the resolidified layer is caused by the overheating boiling of TZM liquid. On the other hand, the number and propagation direction of cracks in the in-situ TZM tile are more complex than those in the ex-situ TZM sample. Firstly, the number of cracks on the in-situ TZM tile is large and dense, while the number of cracks on the surface of the ex-situ melting sample is small and scattered. The crack width of the in-situ tile is greater than that of the ex-situ tile. Secondly, the crack extension direction on the in-situ melted TZM is diverse, while the extension direction of the surface crack on the ex-situ TZM sample is relatively simple, and there is only one kind of crack on the tile surface. Such results provide reference for understanding the thermal behavior of TZM under transient heat flux in fusion devices.

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 dis-charges 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 x 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.

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.

Tungsten (W) is one of the most promising plasma-facing materials for future fusion devices. Although its melting point is the highest among all metals, it still has great risk of melting under extremely high plasma heat fluxes, which is a big concern for ITER and future reactors. Actively cooled W plasma-facing components (PFCs) with both monoblocks and flat-type structure have been successfully installed in the lower divertor of the EAST tokamak since 2021, and provide a good opportunity for direct comparison of the damage mechanism for the two types of PFCs. Various in situ melting phenomena on the lower divertor have been observed by CCD cameras, which have been further verified by post-mortem inspections. Severe melting and even exfoliation of the edge-beveled W plates were observed on some W/Cu flat-type components at horizontal outer targets. Many droplets were ejected during long-pulse operations, which induced a significant increase of W impurities and total irradiation in the core plasma, and thus greatly deteriorated the plasma performance and even caused disruptions. Two different shaping structures of flat-type PFCs show different positions of melting and the corresponding mechanisms. Slight melting was found on the sharp leading edges of W/Cu monoblocks between cassette modules (inter-CM) for horizontal targets with small droplet ejection, which was much improved compared to that observed on the upper W divertor, illustrating that the application of a large-sized bevel chamfer inter-CM was generally effective. In addition, an unexpected melting phenomenon on the dome plate was attributed to the extreme transient heat flux during disruption with runaway electrons. The application of both types of W/Cu PFCs for the divertor provides important experiences and lessons for the engineering design and optimization of divertor PFCs in 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 degrees C in the case of stored energy similar to 250 kJ, and the corresponding local peak heat flux during plasma disruption calculated to be more than several hundred MWm(-2) (similar to 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.

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.

The W/Cu monoblock for divertor target will be exposed to cyclic (∼300 cycles) and extremely high heat flux up to 20 MW/m2 in ITER, which may lead to significant macro cracking and W recrystallization according to existing high heat flux tests. Currently, the degradation of W mechanical properties due to W recrystallization, is regarded as one of the most important factors on cracking formation as well as the lifetime of W components. Therefore, to effectively predict W recrystallization behaviors, a method to simulate the evolution and distribution of W recrystallized fraction during the non-isothermal process was introduced. Coupling with finite element thermal analysis, W recrystallization gradients are successfully obtained. The numerical analysis results show very good agreement with those from existing experiments [1]. Further simulations of different heat flux density (15 MW/m2 − 30 MW/m2) and cyclic number (1–104) with heating time of 5 s, 8 s and 10 s were also carried out. It is found that the shape of the final recrystallized W region is according with the isothermal diagram at the end of the heating stage. In particular, the specific temperature of the corresponding isothermal line can be estimated by heating time and cyclic number. A fitting empirical equation was also given, which can be used to make fast prediction of recrystallization depth by comparing with isothermal diagram for such high heat flux tests. This method underlines the high potential to predict the recrystallization behavior of the W plasma-facing component under complex plasma heat flux in fusion devices.

The leading-edge-induced thermal loading effect due to assembly tolerance between neighboring castellated plasma-facing components is a critical issue in fusion devices. Actively cooled ITER-like W/Cu monoblocks were successfully installed for the upper divertor target in EAST which significantly increases the performance of the divertor power exhaust. The misalignment between neighboring monoblocks was formed inevitably during manufacturing and assembly processes, providing a possibility to demonstrate the leading-edge-induced thermal damage. Indeed, the leading-edge-induced melting phenomena of W/Cu monoblocks on upper divertor targets were observed using CCD a camera during plasma discharges with a large number of droplets ejected from the divertor target, which were also identified at the leading edges of W/Cu monoblocks. Not only that, but also many macro cracks with widths of ∼70 μ m and depths of <5 mm along radial and toroidal directions were also found universally at the leading edges of W/Cu monoblocks by post-mortem inspection after plasma campaigns. Thermal–mechanical analysis by means of finite element simulation demonstrated that the maximum temperature could reach W melting point under the current projected heat load of ∼3 MW m −2 on flat top surface with large misalignment up to 3 mm at the leading edges. Meanwhile, the high temperature also induced high thermal stress and strain concentration at the center of leading edges, at which thermal fatigue cracking could be initially generated. Such type of cracks at leading edges on W/Cu monoblocks may be unavoidable due to long-term, pulsed fatigue effects. However, the influence of these cracks seems to be acceptable thanks to the limited propagated distance due to the self-castellation effect, which still needs long-term tracking. The in situ leading-edge-induced melting and cracking damage on W/Cu monoblocks of the EAST upper divertor target provides significant insight on understand the leading-edge-induced thermal effect in ITER and future fusion devices.

Plasma-wall interaction is one of the key issues in tokamaks, as the material erosion and deposition will strongly influence the lifetime of plasma facing materials, fuel retention and plasma performance. Material migration and deposition have been firstly analyzed after the 2017 experimental campaign in EAST with tungsten (W) upper divertor, molybdenum inner wall and graphite lower divertor. It is found that the deposited elements on plasma-facing materials (PFMs) mainly consisted of Li, C, O, W and Mo etc, which was from the deposition of wall conditioning material and the sputtering of PFMs. More than 90 wt% of the deposits were lithium carbonate and lithium hydroxide due to the routine lithium wall conditioning in EAST. Quantitative characterization of element density on graphite tile from lower inner divertor showed that W and Mo deposition increased when closer to the lower divertor, which is consistent with the decreasing normalized poloidal magnetic flux. The element composition of PFMs along poloidal direction has been measured by handheld X-ray fluorescence (XRF). The concentration of deposited Mo and W is less than 0.2 wt% on most of the PFM surface. Double peaks of W concentration near the upper W divertor and at the midplane tile were observed on the TZM inner wall surface at the high-field side, which can be explained by higher redeposition at the location near the upper inner divertor or with a smaller normalized poloidal magnetic flux. Besides, more redeposited Mo and W was found at the inner and outer divertor near strike points compared with the dome region.

The W-1%Y2O3-0.5%Ti composite fabricated by the spark plasma sintering method has been tested on the EMS-60 facility using 0 to 600 MW/m(2) with pulse duration of 5 ms for single and 100 cycles. It is shown that the cracking threshold under single pulse of the W-1%Y2O3-0.5%Ti is close to 300 MW/m(2), which exceeds that of pure tungsten (similar to 200 MW/m(2)). Moreover, the crack morphology is different than that of pure W under the same heat load condition. These experimental data illustrate that the addition of dispersed oxides and alloying elements can, to some extent, ameliorate high heat load behaviors. Meanwhile, it should be noted that the melting and volatility of the second phase if the heat flux exceeds 400 MW/m(2) for a single pulse will narrow the operation range of the composite. And, the severe damages of crack and matrix melting under fatigue shocks illustrate that the composite still needs further efforts to be improved by the optimization of fabrication processing.

Arc erosion on plasma facing materials in tokamak devices is a potential source of impurities and dust in plasma. A two dimensional axially symmetric COMSOL model with heat transfer, fluid dynamics including phase transition and surface tension effect has been used to describe the formation and evolution of arc craters on plasma facing materials. The formation and evolution of arc crater on W cathode is described in detail. The energy flux loading causes melting of central area within a few nanoseconds. Due to the gradients of incident pressure, the melted layer is extruded out, and thus forms the melt jets. The differences of the arc craters on several related materials in tokamaks under the same pressure and energy flux density are also discussed. The crater temperature of W and Mo is much higher than that of Cu and Al. And, the melting volume of refractory metals W and Mo is significantly lower than that of Cu and Al. Refractory metals are more difficult to be damaged by the arcs, and more suitable for plasma facing materials.

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.