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Using liquid metals confined in capillary porous structures (CPSs) as a plasma-facing component (PFC) could prolong the lifetime of the divertor in the high heat flux area. However, the high atomic number of tin (Sn) limits its acceptable fraction in the main plasma. Therefore, a crucial step in developing this concept is to test it in a tokamak environment, particularly in the diverted plasma region, e.g. ASDEX Upgrade (AUG). In this paper, the design of liquid tin module (LTM) is explained, and the testing in the high heat flux device GLADIS before its use in AUG is presented. The LTM was additively manufactured using selective laser melting, consisting of a 1.5mm porous layer tungsten (W) directly attached to a solid W bulk. The LTM has a plasma-facing area of 16×40mm2 and was filled with 1.54g of Sn. In GLADIS, the module was exposed to power loads between 2 and 8MWm−2 for 1 up to 10s, first unfilled and later filled with Sn. The surface temperature was monitored with infrared imaging and pyrometry. The thermal response was used to compare with simulations in Ansys Mechanical, enabling a determination of the module’s effective thermal properties. Sn droplets could be observed on the infrared camera, until a surface temperature of about a 1000°C was reached. The enhanced wetting of tin on the plasma-facing surface, which was observed by a visible camera, suggests that there is a conditioning of the surface, possibly due to the removal of impurities and oxides. Subsequent examinations of the adjacent tile revealed minor Sn leakages emanating from the module’s edge. Furthermore, the module showed no indication of mechanical failure. Therefore, these results indicated that the LTM qualifies for the heat fluxes expected in ASDEX Upgrade.

Surface modification and deuterium (D) retention in pure tungsten (W) and tungsten-rhenium (W-Re) alloys with Re concentration of 1, 3, 5 and 10 wt.% were investigated after exposure to D plasma with an incident energy of 38 eV/D at about 400 K and various fluences ranging from 7.2 × 1023 to 2.6 × 1025 D/m2. It is found that blistering depends strongly on the exposure fluence for both W and W-Re alloy samples. When the D fluence is higher than 2.8 × 1024 D/m2, the surface of W is covered by two types of blisters. One type is small blisters (8 μm) and extents over several grains. In the cases of W-Re alloys, only small blisters can be observed on their surface. The blisters on the surface of W-Re alloys preferentially appeared on grains with surface orientation close to (111). Cross-section views of blisters show that the blisters on W-Re alloys always originate from intra-granular cavities. The small and large blisters on W originate from intra- and inter-granular cavities, respectively. The surface blistering fluence threshold for W-Re alloys is lower than for W. For both types of materials, rupture of blisters occurs if the underlying crack reaches the surface at the edge of a blister. The number of ruptured blisters increases with increasing D fluence. TDS results show that there is an additional high-temperature peak at ∼ 800 K in W-Re alloys. The amount of D released from W and W-Re alloys increases with increasing D fluence. In addition, the accumulation of D in W-Re alloys is not noticeably influenced by the amount of Re doping at the exposure temperature of ∼ 400 K.

The first wall (FW) of a blanket or divertor faces high temperature plasma. Thus, the FW needs to withstand extremely high heat fluxes. Periodic plasma pulses bring some reliability challenges to FW structures. The FW is composed of tiles, heat sink and a bolt with different connection between the tile and the heat sink. In this study, the performances of two connection forms for the tile and the heat sink were investigated. The FLUENT software was used to evaluate the heat transfer performances for the two types structure. Elastic-plastic analysis was subsequently adopted to estimate the structural properties based on the thermal hydraulic results. At the same time, the thermal and structural results under a transient high thermal load (2 MW/m2) were also analyzed. Finally, structural fatigue analysis under a cyclic thermal load with 0.5 MW/m2 was also performed. The results indicated that there was a large discrepancy in the tile results for the two connections, while the differences in the heat sink result were not significant. The advantages and disadvantages of two connections also be discussed at last.

Mo has been applied to nuclear material, in which H and carbon impurities are unavoidable. In view of this, we have studied the effect of carbon on the H retention in Mo using first-principles simulations. In perfect Mo, there is always repulsion between H and carbon, and impurity carbon cannot capture H atoms. With the appearance of vacancy in Mo, vacancy and carbon-vacancy cluster trap seven and six H atoms, respectively. This means that impurity carbon has little effect on the H vacancy capturing. Finally, we explore the H diffusion by considering the presence of one Mo-carbon layer in Mo. Away from the Mo-carbon layer, H jumps along the optimal route with a diffusion barrier of 0.12 eV. As H moves close and passes through the Mo-carbon layer, the H diffusion barrier is increased to 0.73 eV. Therefore, the repulsive interaction between H and carbon can increase the H energy barrier in the vicinity of the Mo-carbon layer, which prevent the H diffusion and permeation in Mo. The current results can explain the promoting mechanism of bubble formation due to impurity carbon implantation and help us design future Mo-based nuclear material.

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.

Deuterium plasma-driven permeation (PDP) through reduced activation ferritic steel alloy F82H with and without sputter-deposited tungsten (SP-W) coatings has been conducted using a laboratory-scale steady-state plasma facility. SP-W coatings have been found to enhance PDP flux with a much slow breakthrough to reach steady-state. Static deuterium retention has been evaluated with thermal desorption spectroscopy. Enhanced deuterium retention in SP-W coated F82H has been observed, compared with that in bare F82H. Secondary ion mass spectrometry (SIMS) has been carried out to obtain the depth profile of deuterium concentration crossing the W/F82H interface. SIMS data indicate the deuterium concentration is much higher in W coatings than that in F82H substrate, exhibiting an “uphill diffusion” profile. Keywords: Plasma-driven permeation, SIMS, Tungsten coatings, F82H, Deuterium retention

A space-resolved spectroscopy using a 3m normal incidence vacuum ultraviolet (VUV) spectrometer was developed to measure the VUV emission profiles in wavelength range of 300–3200Å from impurities in the edge plasmas in Large Helical Device (LHD). The edge plasma of the LHD is characterized by stochastic magnetic fields with three-dimensional structure intrinsically formed by helical coils called the “ergodic layer”. The VUV spectroscopy has revealed that intensity profiles of the impurity emission from the ergodic layer have strong up-down asymmetries, namely, the emission intensities are quite different between top and bottom plasma edges. Observations of the up-down asymmetries are summarized on the following impurity line emissions: (1) CII 1335.71Å (2s-2p), CIII 977.02Å (2s-2p), and CIV 1548.20Å (2s-2p) from intrinsic carbon impurity ions sputtered from the carbon divertor plates, (2) NeVII 465.22Å (2s-2p), NeVIII 770.41Å (2s-2p), ArVII 585.75Å (3s-3p), and ArVIII 700.24Å (3s-3p) from Ne or Ar ions introduced in plasmas by gas puffing, and (3) WVI 639.68Å (5d-6p) from W ions introduced in plasmas by pellet injection.

A micro-structuring of the tungsten plasma-facing surface can strongly reduce near surface thermal stresses induced by ELM heat fluxes. This approach has been confirmed by numerical simulations with the help of ANSYS software. For experimental tests, two 10 × 10 mm2 samples of micro-structured tungsten were manufactured. These consisted of 2000 and 5000 vertically packed tungsten fibres with dimensions of Ø240 µm × 2.4 mm and Ø150 µm × 2.4 mm, respectively. The 1.2 mm bottom parts of the fibres are embedded in a copper matrix. The top parts of the fibres have gaps about of 10 µm so they are not touching each others. The top of all tungsten fibres was electro-polished. A Nd:YAG laser with a pulse duration 1 ms and a pulse repetition frequency of 25 Hz was used to simulate up to 105 ELM-like heat pulses. No damage on either of the micro-structured tungsten samples were observed. Neon plasma erosion rate and fuel retention of the micro-structured tungsten samples were almost identical to bulk tungsten samples. Keywords: Micro-structured tungsten, PFC, PFM, High heat load, Thermal cycling, Retention, Erosion, Emissivity