Rapid growth of nanostructure on tungsten thin film by exposure to helium plasma

A fiberform nanostructure was synthesized by exposing high-density helium plasma to a 100 nm thick tungsten thin film in the linear plasma device NAGDIS-II. After helium plasma exposure,the cross-section of samples was observed by a scanning electron microscope, transmission electron microscope, and focused ion beam scanning electron microscope. It is shown that the thickness of the nanostructured layer increases significantly for only a short irradiation time. The optical absorptivity remains h...     

Bonding tungsten,W–Cu-alloy and copper with amorphous Fe–W alloy transition

W/Cu graded materials are the leading candidate materials used as the plasma facing components in a fusion reactor. However, tungsten and copper can hardly be jointed together due to their great differences in physical properties such as coefficient of thermal expansion and melting point, and the lack of solid solubility between them. To overcome those difficulties, a new amorphous Fe–W alloy transitional coating and vacuum hot pressing (VHP) method were proposed and introduced in this paper. The morphology, composition and structure of the amorphous Fe–W alloy coating and the sintering interface of the specimens were analyzed by scanning electron microscopy (SEM), energy dispersive spectrometer (EDS) and X-ray diffraction (XRD). The thermal shock resistance of the bonded composite was also tested. The results demonstrated that amorphous structure underwent change from amorphous to nano grains during joining process, and the joined W/Cu composite can endued plasma thermal shock resistance with energy density more than 5.33 MW/m². It provides a new feasible technical to join refractory tungsten to immiscible copper with amorphous Fe–W alloy coating.     

The effects of tungsten addition on the microstructural stability of 9Cr–Mo Steels

The effects of tungsten addition on the microstructure and high-temperature tensile strength of 9Cr–Mo steels have been investigated by using three different steels: M10 (9Cr–1Mo), W18 (9Cr–0.5Mo–1.8W), and W27 (9Cr–0.1Mo–2.7W) steels. The tungsten-added 9Cr steels have revealed better high-temperature tensile strength. Microchemical analysis for (Cr,Fe)₂ (C,N) revealed that the tungsten addition increased the Cr/Fe ratio, which resulted in the lattice expansion of (Cr,Fe)₂ (C,N), and then the enhanced pinning effect on the glide of dislocation. In addition, in M10 steel, the M₂₃C₆ carbides quickly grew and agglomerated, while the tungsten-added 9Cr steels revealed a fine and uniform distribution of M₂₃C₆ carbides. Dislocation recovery during tempering treatments was delayed in tungsten-added 9Cr steels, which was correlated with the stabilized precipitates and the decreased self-diffusivity of iron. It is, thus, believed that the excellent high-temperature tensile strength of tungsten-added 9Cr steels is attributed to the stabilized M₂X carbo-nitrides and M₂₃C₆ carbides and the decreased self-diffusivity of iron.     

Overview of the US–Japan collaborative investigation on hydrogen isotope retention in neutron-irradiated and ion-damaged tungsten

The effect of neutron-irradiation damage has been mainly simulated using high-energy ion bombardment. A recent MIT report (PSFC/RR-10-4, An assessment of the current data affecting tritium retention and its use to project towards T retention in ITER, Lipschultz et al., 2010) summarizes the observations from high-energy ion bombardment studies and illustrates the saturation trend in deuterium concentration due to damage from ion irradiation in tungsten and molybdenum above 1 displacement per atom (dpa). While this prior database of results is quite valuable for understanding the behavior of hydrogen isotopes in plasma facing components (PFCs), it does not encompass the full range of effects that must be considered in a practical fusion environment due to short penetration depth, damage gradient, high damage rate, and high primary knock-on atom (PKA) energy spectrum of the ion bombardment. In addition, neutrons change the elemental composition via transmutations, and create a high radiation environment inside PFCs, which influences the behavior of hydrogen isotope in PFCs, suggesting the utilization of fission reactors is necessary for neutron-irradiation. Under the framework of the US–Japan TITAN program, tungsten samples (99.99 at.% purity from A.L.M.T. Co.) were irradiated by fission neutrons in the High Flux Isotope Reactor (HFIR), Oak Ridge National Laboratory (ORNL), at 50 and 300 °C to 0.025, 0.3, and 2.4 dpa, and the investigation of deuterium retention in neutron-irradiated tungsten was performed in the Tritium Plasma Experiment (TPE), the unique high-flux linear plasma facility that can handle tritium, beryllium and activated materials. This paper reports the recent results from the comparison of ion-damaged tungsten via various ion species (2.8 MeV Fe²⁺, 20 MeV W²⁺, and 700 keV H^?) with that from neutron-irradiated tungsten to identify the similarities and differences among them.     

Preparation of tungsten coatings on graphite by electro-deposition via Na2WO4–WO3 molten salt system

Tungsten coating on graphite substrate is one of the most promising candidate materials as the ITER plasma facing components. In this paper, tungsten coatings on graphite substrates were fabricated by electro-deposition from Na₂WO₄–WO₃ molten salt system at 1173 K in atmosphere. Tungsten coatings with no impurities were successfully deposited on graphite substrates under various pulsed current densities in an hour. By increasing the current density from 60 mA cm⁻² to 120 mA cm⁻² an increase of the average size of tungsten grains, the thickness and the hardness of tungsten coatings occurs. The average size of tungsten grains can reach 7.13 μm, the thickness of tungsten coating was in the range of 28.8–51 μm, and the hardness of coating was higher than 400 HV. No cracks or voids were observed between tungsten coating and graphite substrate. The oxygen content of tungsten coating is about 0.022 wt%.     

Electro-deposition metallic tungsten coatings in a Na2WO4–WO3 melt on copper based alloy substrate

The tungsten coating was prepared by electro-deposition technique on copper alloy substrate in a Na₂WO₄–WO₃ melt. The coating's surface and cross-section morphologies as well as its impurities were investigated by XPS, SEM and line analysis. Various plating durations were investigated in order to obtain an optimal coating's thickness. The results demonstrated that the electro-deposited coating was compact, voidless, crackless and free from impurities. The tungsten coating's maximum Vickers hardness was measured to be 520 HV. The tungsten coating's minimum oxygen content was determined to be 0.018 wt%. Its maximum thickness was measured to be 1043.67 μm when the duration of electrolysis was set to 100 h. The result of this study has demonstrated the feasibility of having thicker tungsten coatings on copper alloy substrates. These electrodeposited tungsten coatings can be potentially implemented as reliable armour for the medium heat flux plasma facing component (PFC).     

Effect of the displacement damage from argon ion irradiation on the synergistic effect of helium–hydrogen in tungsten

The effect of argon ion pre-irradiation on helium and hydrogen ion irradiation was investigated in tungsten. At the same time, comparative experiments were carried out on the irradiation of helium and hydrogen ions in tungsten. Without the argon ion irradiation, the energy of 35 keV hydrogen ions mainly accelerated the coalescence of defects created by the 60 keV helium ions, the irradiation damage degree increased with hydrogen ion fluence increasing. With the argon ion irradiation, lots of voids were created by argon ion irradiation, which increased the helium and hydrogen retention and the synergistic effect of helium–hydrogen in tungsten. In the same hydrogen fluence, the surface damage degree with argon ion pre-irradiation was higher than that without argon ion pre-irradiation.     

Comparison of hydrogen isotope retention and irradiation damage behaviors in tungsten and SS-316 with simultaneous C+–D2+ implantation

Behaviors of hydrogen isotope retention and damages in tungsten and SS-316 with simultaneous C⁺–D₂⁺ implantation were compared to those with only D₂⁺ implantation using X-ray photoelectron spectroscopy (XPS), Thermal desorption spectroscopy (TDS), glow discharge-optical emission spectroscopy (GD-OES) and Transmission electron microscopy (TEM).The total D retention for SS-316 with only D₂⁺ implantation was about 45% as large as that for tungsten. The D retention for simultaneous C⁺–D₂⁺ implanted tungsten and SS-316 clearly increased as a factor of 1.7, which is almost the same among these samples. The density of dislocation loops was enhanced by the simultaneous C⁺–D₂⁺ implantation, indicating the D trapping site would be produced by C⁺ implantation. As for the D desorption temperature, small shift toward lower temperature side was found for SS-316 compared to tungsten, indicating the D trapping energy by dislocation loops and grain boundary for SS-316 is lower than that for tungsten.     

Non-invasive optical characterization and estimation of Zn porosity in gas tungsten arc welding of Fe–Al joints using CR model and OES measurements

In this study, we employed a non-invasive approach based on the collisional radiative(CR) model and optical emission spectroscopy(OES) measurements for the characterization of gas tungsten arc welding(GTAW) discharge and quantification of Zn-induced porosity during the GTAW process of Fe–Al joints. The OES measurements were recorded as a function of weld current, welding speed, and input waveform. The OES measurements revealed significant line emissions from Zn-I in 460–640 nm and Ar-I in 680–80...     

Investigation of plasma exposed W–1% La2O3 tungsten in a high ion flux,low ion energy,low carbon impurity plasma environment for the International Thermonuclear Experimental Reactor

Previous investigations of tungsten for the International Thermonuclear Experimental Reactor (ITER) were focusing on using energetic ion beams whose energies were over 1 keV. This study presents experimental results of exposed W–1% La₂O₃ in high ion flux (10²² m^–2), low ion energies (about 110 eV) steady-state deuterium plasmas at elevated temperatures (873–1250 K). The tungsten samples are floating during plasma exposure. Using a high-pressure gas analyzer, the residual carbon impurities in the plasma are found to be about 0.25%. No carbon film is detected on the surface by the EDX analysis after plasma exposure. An infrared pyrometer is also used as an in situ detector to monitor the surface emissivities of the substrates during plasma exposure. Using the scanning electron microscopy, microscopic pits of sizes ranging from 0.1 to 5 μm are observed on the plasma exposed tungsten surfaces. These pits are believed to be the results of erupted deuterium gas bubbles, which recombine underneath the surface at defect locations and grain boundaries, leading to substrate damage and erosion loss of the substrate material. Low temperature plasma exposure of a tungsten foil indicates that deuterium gas (D₂) is trapped inside the substrate. Macroscopic blisters are observed on the surface. The erosion yield of the W–1% La₂O₃ increases with temperature and seems to saturate at around 1050 K. Scattered networks of bubble sites are found 5 μm below the substrate surface. High temperature plasma exposure appears to reduce the population as well as the size of the pits. The plasma exposed W–1% La₂O₃ substrates, exposed above 850 K, retain about 10¹⁹ D/m², which is two orders of magnitude less than those retained by the tungsten foils exposed at 400 K.