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In ITER and future fusion devices, the plasma facing surfaces are expected to suffer from the superposition of steady-state and transient heat loads. The extreme high transient heat fluxes cause the deterioration of material surface that increases from roughing to cracking even melting. And, the high heat flux tests on pre-heated tungsten implied that the steady-state heat load induced initial surface base temperature would have great influence on the transient heat flux induced damages. The present efforts tried to explain the mechanism of this phenomenon in view of the thermal-mechanical analysis by means of finite element simulation. The surface temperature and stress-strain distribution and evolution of a cylindrical W/Cu block under superposition of the steady-state heat loads corresponding to base temperature (20–600 °C) and transient heat fluxes (5 ms, 0–900 MW m −2 ) were successfully simulated. The corresponding relations between stress-strain and temperature are compared with the yield strength-temperature relation to analyze the surface event evolution during the heating and cooling phases, which showed how the base temperature influenced the transient heat flux induced damages. The analysis results theoretically identified that the pre-heating by steady-state heat load indeed influenced the transient heat flux induced damages. Typically, no any crack would be generated even under extreme high transient heat fluxes close to the melting threshold if the base temperature exceeded the ductile to brittle transition temperature, and the surface also showed deterioration from roughing to cracking while then roughing with increasing of transient heat flux under a certain base temperature, which were in good agreement with the experiments results. [ABSTRACT FROM AUTHOR]

• Leading edge-induced melting of W/Cu monoblocks was found only at inter-CMs. • Thermal analysis identified the current dual chamfer shaping structure (1 mm × 1 mm) did not work as-intend at inter-CMs. • The symmetric CMs with different chamfer solutions at inter-CMs and intra-CM have been proposed to effectively improve the thermal performance of the upper W divertor in EAST. EAST has installed a full ITER-like W divertor with W/Cu monoblocks for targets in the upper part of vacuum chamber since 2014. Although the dual chamfer shaping structure with 1 mm cutting in both toroidal and radial directions (1 mm × 1 mm) has been employed to mitigate the leading edge-induced thermal effect, severe leading edge-induced melting was widely observed on both inner and outer targets. All melting only occurred at the leading edges between neighboring cassette modules (inter-CMs), while it was not found at the leading edges between neighboring monoblocks within one cassette module (intra-CM). To optimize the chamfering structure for avoiding melting, thermal analysis of W/Cu monoblocks was performed by finite element method. It is identified that the current dual chamfer shaping structure (1 mm × 1 mm) does not work for inter-CMs due to the large radial misalignment which was hardly controlled and could even reached up to 3 mm. With such large misalignment, the maximum surface steady state heat load is only 1.2 MW m−2 for current structure. Thus, a symmetric structure of CMs which survive EAST operations requirement with two different toroidal magnetic fields has been proposed to improve the thermal performance and avoid melting. At inter-CMs, the optimized chamfering structure suitable for larger misalignments and heat fluxes has been suggested. And, the current dual chamfer shaping structure is remained at intra-CM after thermal evaluation. Such symmetric CMs should ensure EAST operate with heat flux up to 3 MW m−2 with 3 mm misalignment level. And, further improvement to withstand a higher heat flux is also proposed and discussed. [ABSTRACT FROM AUTHOR]

To investigate the behaviors of tungsten based plasma facing materials under fusion relevant transient heat flux, two kinds of commercial tungsten materials processed with and without rolling respectively, have been tested on the EMS-60 facility using 0–800 MW/m 2 with single pulse duration of 5 ms. The macro and micro cracks were observed as expected, whose distribution patterns were depicted and explained by the residual stress distribution analysis with the finite element method. Moreover, when high heat flux exceeded 600 MW/m 2 (∼5 ms), the refining of grain crystal occurred not only at the central area but also around the micro cracks. [ABSTRACT FROM AUTHOR]

• The partial melting area in the in-situ melting TZM tile is like an impact pit, which causes the melted liquid droplets in the melting area to splash around, while ex-situ melting TZM does not have this feature. • The number and width of cracks in the in-situ TZM brick are more than those in the ex-situ test TZM tiles. Meanwhile the crack extension directions in the in-situ melting TZM tile are various, while those features on the ex-situ TZM tiles is single. • The capillary waves around the castle structure in the test samples are the result of the wetting effect of liquid metal and capillary force, and the holes in the re solidified layer are caused by the overheating boiling of TZM liquid. 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. [ABSTRACT FROM AUTHOR]

• Statistics analysis on the disruptivity due to W monoblock damage have been performed for each EAST campaign from 2017 to 2019. • After the 2017 campaign, the main contribution to impurity-induced disruption is W due to divertor damage. • The disruptivity due to W monoblock damage increases with the increase of the auxiliary heating power. • Normal discharges can be promptly recovered after disruptions due to W monoblock damage. • The W divertor damage is currently tolerable for the overall EAST operations. The upper divertor of EAST has been upgraded to ITER-like tungsten plasma-facing components since 2014. The impact of tungsten monoblock damage on plasma operations in recent EAST campaigns was studied. Various damage on the divertor tiles such as melting, cracking and arcing was found. Statistics analysis on the disruptivity due to W monoblock damage have been performed. W impurity produced from W divertor damage becomes the main impurity that leads to disruption during 2018, 2019a and 2019b campaigns compared to other impurities. With increasing auxiliary heating power, the W divertor damage becomes more and more serious. The current divertor damage is still tolerable for the overall EAST operations. Such studies can provide good references for the future fusion devices with W monoblocks, such as ITER. [ABSTRACT FROM AUTHOR]

• The effects of hot spot and melting on plasma performance are studied for EAST. • Two methods can be used for fast recovery of W melting-induced disruptions in EAST. • A statistic of disruptions due to severe impurity burst for two campaigns is shown. During the tokamak operations, plasma-wall interactions (PWI) can severely damage the plasma-facing components (PFCs), which will lead to degradation of plasma performance due to radiation losses by the impurities released from the damaged PFCs surfaces. The PFCs damage and its effects on the plasma performance are studied experimentally for the EAST tokamak. Substantial dusts were generated from the hot spots at the guard limiter of LHCD antenna and the tungsten (W) melting processes from the W upper divertor, which results in disruption due to high radiation losses. Two methods can be effectively used to recover the plasma discharges after tungsten melting induced disruptions, reducing the auxiliary heating power or moving away the position of strike point from the melting site. Statistics analysis indicate that about 12.0 % and 38.3 % of disruptions were due to severe impurity burst from the strong PWI processes in 2017 and 2018 campaign, respectively. In 2018 campaign, the plasma disruption due to tungsten impurity burst is about 26.3 %, which is much higher than 4.8 % in 2017 campaign. This is due to the higher auxiliary heating power, larger misalignment on W PFCs and the new W guard limiter. The main features on the effects of PFC damage on plasma operation are shown. [ABSTRACT FROM AUTHOR]