Low energy sputtering of Mo surfaces Granberg, F.; Litnovsky, A.; Nordlund, K.
Journal of nuclear materials,
October 2020, 2020-10-00, 20201001, Letnik:
539
Journal Article
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Surfaces of materials subject to irradiation will be affected by sputtering, which can be a beneficial effect, like in the coating industry where a material is sputtered and redeposited on to another ...material to coat it. However, in most cases sputtering is an unwanted side-effect, for instance in nuclear fusion reactors, where the wall material will be degraded. This effect needs to be understood in order to be able to predict its consequences. To understand the sputtering, on an atomistic level, we have thoroughly investigated molybdenum surface sputtering by computational means. Molybdenum was chosen as detailed experimental studies have been carried out on it and it is one candidate material for the diagnostic mirrors in ITER, facing the plasma. In this study, we thoroughly investigate the molybdenum samples of different surface orientations, and their response to low energy argon plasma irradiation, by molecular dynamics simulations. We find both a surface orientation and ion energy specific sputtering yield of the samples, and a very good agreement with the experiments available in the literature. A few different setups were investigated to observe differences as well as to understand the key features affecting the sputtering events. The different simulation setups revealed the optimal one to represent the experimental conditions as well as the mechanisms behind the observed discrepancies between different modelling setups.
Plasma-facing materials and components in a fusion reactor are the interface between the plasma and the material part. The operational conditions in this environment are probably the most challenging ...parameters for any material: high power loads and large particle and neutron fluxes are simultaneously impinging at their surfaces. To realize fusion in a tokamak or stellarator reactor, given the proven geometries and technological solutions, requires an improvement of the thermo-mechanical capabilities of currently available materials. In its first part this article describes the requirements and needs for new, advanced materials for the plasma-facing components. Starting points are capabilities and limitations of tungsten-based alloys and structurally stabilized materials. Furthermore, material requirements from the fusion-specific loading scenarios of a divertor in a water-cooled configuration are described, defining directions for the material development. Finally, safety requirements for a fusion reactor with its specific accident scenarios and their potential environmental impact lead to the definition of inherently passive materials, avoiding release of radioactive material through intrinsic material properties. The second part of this article demonstrates current material development lines answering the fusion-specific requirements for high heat flux materials. New composite materials, in particular fiber-reinforced and laminated structures, as well as mechanically alloyed tungsten materials, allow the extension of the thermo-mechanical operation space towards regions of extreme steady-state and transient loads. Self-passivating tungsten alloys, demonstrating favorable tungsten-like plasma-wall interaction behavior under normal operation conditions, are an intrinsic solution to otherwise catastrophic consequences of loss-of-coolant and air ingress events in a fusion reactor. Permeation barrier layers avoid the escape of tritium into structural and cooling materials, thereby minimizing the release of tritium under normal operation conditions. Finally, solutions for the unique bonding requirements of dissimilar material used in a fusion reactor are demonstrated by describing the current status and prospects of functionally graded materials.
Prediction of material performance in fusion reactor environments relies on computational modelling, and will continue to do so until the first generation of fusion power plants come on line and ...allow long-term behaviour to be observed. In the meantime, the modelling is supported by experiments that attempt to replicate some aspects of the eventual operational conditions. In 2019, a group of leading experts met under the umbrella of the IEA to discuss the current position and ongoing challenges in modelling of fusion materials and how advanced experimental characterisation is aiding model improvement. This review draws from the discussions held during that workshop.
Topics covering modelling of irradiation-induced defect production and fundamental properties, gas behaviour, clustering and segregation, defect evolution and interactions are discussed, as well as new and novel multiscale simulation approaches, and the latest efforts to link modelling to experiments through advanced observation and characterisation techniques.
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•W-11.2Cr-1.7 Zr presents the optimal oxidation resistance.•The addition of Zr can improve the high temperature oxidation resistance.•The power law is suitable to exactly describe the ...oxidation curves.
W-Cr-Zr systems with different compositions were oxidized in a mixed gas (Ar + 20 vol.% O2) atmosphere at 1000 °C. The power law which was used to describe the oxidation behaviour, indicates that W-11.2wt.%Cr-1.7wt.%Zr has an excellent oxidation behaviour. As analysed from the ten-hour exposure, W-Cr-Zr thin film oxidation shows a self-passivating stage followed by a linear oxidation stage. Furthermore, a study on the addition of zirconium indicates that zirconia particles act as diffusion barriers for the chromium cation diffusion and another function as the nucleation sites for the formation of the initial oxide scale.
•Yttrium improves the self-passivation of the investigated W-Cr-Y alloys.•Yttrium reduces the oxidation rates by one order of magnitude compared to the W-Cr system.•The effects of yttrium are the ...suppression of mixed oxides and pores.•The optimal composition found in this study is of W- 12 wt.%Cr-0.6 wt.%Y.•An idealized simulation shows that the loss of alloying elements due to diffusion and sputtering is small.
Tungsten is a prime material candidate for the first wall of a future fusion reactor. In the case of a loss-of-coolant accident (LOCA) wall temperatures of about 1450 K could be reached lasting about 30–60 days due to nuclear decay heat. In the worst case scenario combining LOCA with air ingress, the formation and release of highly volatile and radioactive tungsten trioxide (WO3) into the environment can occur. Smart self-passivating tungsten alloys preventing the formation of WO3 can be a way to mitigate this release.
In this contribution we present the studies of a new yttrium-containing W-Cr-Y alloys. The extent up to which yttrium acts as an active element improving the adherence and stability of the protective Cr2O3 layer formed during oxidation is assessed. The approach is similar to the one taken for high-temperature steels where active elements stabilize the oxide layers at a substantially reduced thickness by changing the oxygen diffusion and improving the adherence of the protective oxide layer by e.g. avoiding of pores. Further, simulations on mobilized material for the case of a LOCA are developed. In addition, the loss of alloying elements during normal operation of a reactor is estimated. This is done by modelling a thermally activated diffusion, using a diffusion coefficient which is extrapolated from experimental data at higher values.
The oxidation behaviour of magnetron sputtered and therefore alloyed at the atomic level W-Cr-Y alloys is tested in a thermo-gravimetric facility. The isothermal oxidations are performed in a gas mixture, containing 20 kPa oxygen and 80 kPa argon under ambient pressure at temperatures of 1273 K and 1473 K, respectively. Experiments with W-Cr-Y show a parabolic oxidation rate of kp=3·10−6mg2cm−4s−1 which is more than five orders of magnitude lower than that of pure tungsten at 1273 K. Investigations using X-ray diffraction analysis and focused ion beam cross-sections in combination with scanning electron microscopy and energy dispersive X-ray spectroscopy are conducted. A protective Cr2O3 layer is detected on the surface with a thickness between 100 and 300 nm.
•Optical and laser-based diagnostics in ITER will use mirrors to observe plasma.•Single crystal materials are among the best mirror materials.•Single crystal rhodium mirrors just became ...available.•First direct performance test of single crystal molybdenum and rhodium was made.•Condition of test was corresponding to those in in situ mirror cleaning system in ITER.•Sputtering by steady-state He plasma corresponded to 50–100 cleaning cycles in ITER.•Single crystal Rh kept the specular reflectivity despite of 1μm sputtered by plasma.•Single crystal Mo shows degradation of reflectivity especially in the UV range.
All optical and laser diagnostics in ITER will use mirrors to observe the plasma radiation. In the ITER environment, mirrors may become contaminated with plasma impurities hampering the performance of corresponding diagnostics. An in-situ mirror cleaning is proposed, which relies on ion sputtering of the contaminants and the affected mirror material. Previous research demonstrated the advantages of single crystal (SC) molybdenum (Mo) under sputtering conditions over polycrystalline concepts. Recently, the first single crystal rhodium (Rh) mirrors became available and tests have been started at the Forschungszentrum Jülich.
In a direct test, SC Rh and SC Mo mirrors were exposed under identical conditions in steady-state helium plasmas in the linear plasma device PSI 2. The energy of impinging ions was ∼100eV matching conditions expected in the in-situ cleaning system in ITER. During exposure, molybdenum mirrors lost 420–500nm due to sputtering. Rhodium mirrors lost about 1μm. Exposure corresponded to 50–100 cleaning cycles in ITER. Nevertheless, rhodium mirrors have preserved their specular reflectivity, showing the maximum degradation of less than 7% at 250nm. Molybdenum mirrors demonstrated a moderate decrease of specular reflectivity of 12–25%. Results open new perspectives for the use of single crystals in ITER diagnostics.
In this work, W-Cr-Zr as a self-passivating tungsten alloy is studied. Spark plasma sintering (SPS) is used to prepare the sample. The influence of the heating rate on the densification process and ...microstructure evolution during SPS was investigated. The increasing the heating rate enhance the homogeneity of the microstructure, however, did not show great effect on overall the sample density. High heating rate with high thermal inertia led the actual temperature to approach or even exceed WCr alloy solution temperature. Based on the punch displacement curves, it can be found that the densification temperatures are approximately 1380 °C and the high heating rate could promote the densification rate. On the other hand, high heating rate required high current, which can decrease the sintering activation energy and improve the densification process.
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•High dense, fine-grained and homogeneous W-Cr-Zr alloy can be obtained by SPS technology.•High heating rate contributes to the densification rate of W-Cr-Zr samples.•High current can improve the homogenization process by enhancing atomic diffusion.
Field-assisted sintering technology (FAST), as a fast densification method with low process temperature, was used to manufacture self-passivating tungsten alloys (SPTAs) of W-Cr-Zr in this work. To ...clarify the behaviors of grain growth and Cr-rich phase precipitation under the action of electric current during the densification process, interrupted sintering at different temperatures (600–1000 °C) were performed. According to the viscous flow theory, the activation energy of W-Cr-Zr sample for densification is ~23 kJ/mol. The differential form of power law was adopted to evaluate the grain growth behavior. It is found that the W-Cr-Zr alloy consolidated by FAST has a low activation energy for grain growth of 82 kJ/mol. The Cr-rich phase could be confirmed by XRD spectra even when the sintering was interrupted at 600 °C. From the characterization of the cross-sectional microstructure, the Cr-rich phases tend to precipitate at sintering necks and defects (cracks/voids) in particle interiors. The low formation temperature of the Cr-rich phase is attributed to local overheating caused by local high electric current. This work provides significant insight into the mechanisms underlying the densification and the evolution of the microstructure of the SPTAs during the FAST process.
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•The FAST densification of W-Cr-Zr alloy follows viscous flow mechanism.•The activation energy for grain growth of W-Cr-Zr alloy is ~82 kJ/mol.•Local overheating is the reason for the formation of the Cr-rich phase.
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