Erosion Of Plasma-facing Materials Research Articles

Characterization of dust produced during the 2021 first campaign in EAST tokamak

• The dust collected from 2021 first experimental campaign in EAST was characterized and analyzed. • By using the diagnostic observation window, dust was collected during the campaign for several times. The local dust production rate decreases as the plasma operation goes. • Ferromagnetic dust with the shape of strips, flakes, irregular blocks and spheroids has been found in EAST. 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.

The impact of surface morphology on the erosion of metallic surfaces – Modelling with the 3D Monte-Carlo code ERO2.0

• The micro-scale model for simulation of the surface roughness effect on the erosion of PFCs and transport of sputtered material implemented previously into the 3D Monte-Carlo code ERO2.0 was applied for modelling of sputtering and deposition in the JET-ILW divertor conditions (smooth Tile 5 and rough Tile 6). • Micro-scale simulations have been conducted using two different assumptions: with and without tracing of incident plasma/impurity ions near the rough surface. Tracing of incident ions near the surface forms more realistic shadowing patterns than the case when this pattern is calculated based on magnetic field lines. • Simulations without tracing of incident ions have shown that surface roughness leads to a reduction of the effective sputtering yield by up to 50% depending on surface parameters. • Simulations including tracing of incident ions near the surface have confirmed decrease of the net erosion from the rough surface of the Tile 6 in comparison to the smoother surface of the Tile 5 by up to 50%. The roughness of metallic surfaces has a vital impact on the erosion of plasma-facing materials. Roughness determines the effective sputtering yield Y eff of the facing material. The angular/energy distribution of sputtered particles, and the spatial erosion and deposition distribution. The model for simulation the effect of the surface roughness was earlier implemented into the 3D Monte-Carlo code ERO2.0 and validated using results of ion beam experiments and experiments in the linear plasma device PSI-2. In the present study the developed ERO2.0 surface morphology model was applied to the JET-ILW tungsten (W) divertor consisting of smooth bulk W and W-coated CFC components. Influence of the surface roughness on the W erosion as well as on the transport of sputtered material in conditions of inclined magnetic field was investigated. Simulation results are in a good agreement with existing experimental findings.

Erosion and deposition in the JET divertor during the second ITER-like wall campaign

Erosion of plasma-facing materials and successive transport and redeposition of eroded material are crucial processes determining the lifetime of plasma-facing components and the trapped tritium inventory in redeposited material layers. Erosion and deposition in the JET divertor were studied during the second JET ITER-like wall campaign ILW-2 in 2013–2014 by using a poloidal row of specially prepared divertor marker tiles including the tungsten bulk tile 5. The marker tiles were analyzed using elastic backscattering with 3–4.5 MeV incident protons and nuclear reaction analysis using 0.8–4.5 MeV 3He ions before and after the campaign. The erosion/deposition pattern observed during ILW-2 is qualitatively comparable to the first campaign ILW-1 in 2011–2012: deposits consist mainly of beryllium with 5–20 at.% of carbon and oxygen and small amounts of Ni and W. The highest deposition with deposited layer thicknesses up to 30 μm per campaign is still observed on the upper and horizontal parts of the inner divertor. Outer divertor tiles 5, 6, 7 and 8 are net W erosion areas. The observed D inventory is roughly comparable to the inventory observed during ILW-1. The results obtained during ILW-2 therefore confirm the positive results observed in ILW-1 with respect to reduced material deposition and hydrogen isotopes retention in the divertor.

Effect of surface segregation and mobility on erosion of plasma-facing materials in magnetic fusion systems

The present work studies the combined effects of collisional sputtering and mixing processes of carbon impurities in tungsten plasma-facing material integrated with thermal processes including surface segregation and diffusion. We used our ITMC-DYN package, which includes description of all collisional and thermal processes, for the analysis of recent experimental results of tungsten erosion and carbon implantation at various target temperatures. Self-consistent integrated modeling predicted thermal processes effects on erosion/deposition dynamics and defined decisive parameters range and their importance. Critical parameters were estimated based on available experimental data. The integrated simulation reproduced the experimental results and predicted the transition from enhanced tungsten erosion to significant carbon coverage on the tungsten surface. These effects for wider range of system conditions with C/H ions irradiation and for reactor conditions can be predicted by including detailed modeling of chemical erosion processes in a self-consistent manner.

Investigation of RF-enhanced plasma potentials on Alcator C-Mod

Abstract Radio frequency (RF) sheath rectification is a leading mechanism suspected of causing anomalously high erosion of plasma facing materials in RF-heated plasmas on Alcator C-Mod. An extensive experimental survey of the plasma potential ( Φ P ) in RF-heated discharges on C-Mod reveals that significant Φ P enhancement (>100 V) is found on outboard limiter surfaces, both mapped and not mapped to active RF antennas. Surfaces that magnetically map to active RF antennas show Φ P enhancement that is, in part, consistent with the recently proposed slow wave rectification mechanism. Surfaces that do not map to active RF antennas also experience significant Φ P enhancement, which strongly correlates with the local fast wave intensity. In this case, fast wave rectification is a leading candidate mechanism responsible for the observed enhancement.

Effects of transient heating events on tungsten plasma-facing materials in a steady-state divertor-plasma environment**Present address: University of California-San Diego, KLA-Tencor, 1 Technology Drive, Milpitas, CA 95035, USA.

Nearly all heat-pulse tests of plasma-facing materials (PFMs) have been completed in vacuum environments without the presence of a background plasma. Thus, combined effects of thermal transients on materials undergoing plasma exposure need to be explored. Heat-pulse experiments have been conducted in the PISCES-A device using a pulsed laser in a divertor-like plasma background. The results indicate that the erosion of PFMs is enhanced as compared with transient-only or plasma-only experiments, and the threshold energy for material removal by a transient heat pulse in a steady-state plasma background is reduced. There appears to be a minimum ion fluence to the surface needed to cause these effects. Initial experiments with pre-loaded material samples exposed to plasmas and heat pulsed in the lower divertor of the DIII-D tokamak using the Divertor Material Evaluation System indicate that similar effects occur in confinement devices.

Present Research Status on Divertor and Plasma Facing Components for Fusion Power Plants

ABSTRACTThis paper is a review of the present status of research activities on the divertor and the first wall for fusion power plants in Japan. There are several key issues as follows; 1) plasma facing materials, 2) fabrication techniques, 3) thermal hydraulics, and 4) component design. It can be said that solid surface plasma facing components are the present reference design of fusion power plants in Japan. Based on plasma conditions, it turns out that sputtering erosion of plasma facing materials could not be negligible, and some armor would be necessary. From this view point, tungsten is one of the most promising materials. This paper reviews the present engineering status on tungsten as a plasma facing material for fusion power plants.

Experimental Investigation into Melt-Layer Erosion of Plasma Facing Materials

This paper describes melt-layer erosion experiments conducted at North Carolina State University. Metallic samples are exposed to a pulsed plasma heat flux produced by an electrothermal plasma gun. Results for aluminum (2042 Al), copper (OFHC Cu), and stainless steel (316 SS) samples are discussed. The electrothermal plasma gun operated at discharge energies between 0.07 and 8.62 kJ. The sample absorbed fluence ranged between 0.09 and 1.93 MJ/m2. The net erosion depth of aluminum approached 1 mm at high energies. Following exposure the samples are bisected to reveal the thickness of material resolidified to the surface. A thickness of resolidified material on an aluminum sample in excess of 50 μm is observed.