Comparison of Damages on Tungsten Surface Exposed to Noble Gas Plasmas

Tungsten was exposed to pure Ar or Ne plasmas over 1550 K at several incident ion energies. Even under the irradiation condition that the tungsten nanostructure is formed by He plasma irradiation, holes/bubbles and fiberform nanostructures were not formed on the surface by exposure to Ar or Ne plasmas. In addition, the results from energy dispersive X-ray spectroscopy supported the facts that Ar and Ne did not remain in the sample. We will discuss the reason for the diferences in the damage to the tungsten surface exposed to noble gas plasmas.     

Mechanism of He release from He implanted samples deduced from post-implantation He release measurements

A combination of post-implantation, room temperature, He release measurements and surface erosion investigation by scanning electron microscopy was used for the study of the possible mechanisms of He release from implanted samples. He release was greatly accelerated at fluences exceeding a critical value. The critical fluence for fast He release was found to be smaller than that needed for the onset of surface erosion and was independent of surface erosion, (i.e. different samples with markedly different amounts of surface erosion exhibited the same He release). Post-implantation He release could be explained in terms of atomic diffusion processes. It was suggested that at high He concentrations this diffusion takes place via micro-channels created by micro-erosion processes that are independent of the known macro-erosion processes (such as flaking, cracking and blistering).     

Comparison of High-Energy He+ and D+ Irradiation Impact on Tungsten Surface in the IR-IECF Device

Polycrystalline tungsten specimens were irradiated in the Iranian Inertial Electrostatic Confinement Fusion device (IR-IECF) by high energy (~100 keV) and high fluency (~10¹⁹ ions/cm²) helium and deuterium plasma to investigate the implantation impact of high energetic ions on tungsten as a candidate for fusion first wall material. Comparison of the exposure by He and D₂ plasma and influence of high temperature (~1,100 °C) implantation of each ion has been examined. Scanning electron microscopy was used to investigate surface morphology changes for various ion fluencies. Results showed the onset of visible surface pores formation especially for helium implanted samples which increased with higher implant fluencies, eventually resulting in a rough and flaky surface structure, unlike deuterium implanted samples on which smoothening of the surface occurred. Microhardness measurements were used to evaluate mechanical properties of implanted tungsten. Each specimen sustained surface hardening after implantation which was observed to increase with greater ion dose. The phase formation and structural evolution were studied by X-ray diffractometry method.     

Fabrication and characterization of He-charged ODS-FeCrNi films deposited by a radio-frequency plasma magnetron sputtering technique

He-charged oxide dispersion strengthened(ODS)FeCrNi films were prepared by a radiofrequency(RF)plasma magnetron sputtering method in a He and Ar mixed atmosphere at150℃.As a comparison,He-charged FeCrNi films were also fabricated at the same conditions through direct current(DC)plasma magnetron sputtering.The doping of He atoms and Y_2O_3 in the FeCrNi films was realized by the high backscattered rate of He ions and Y_2O_3/FeCrNi composite target sputtering method,respectively.Inductive coupled plasma(ICP)and x-ray photoelectron spectroscopy(XPS)analysis confirmed the existence of Y_2O_3 in FeCrNi films,and Y_2O_3 content hardly changed with sputtering He/Ar ratio.Cross-sectional scanning electron microscopy(SEM)shows that the FeCrNi films were composed of dense columnar nanocrystallines and the thickness of the films was obviously dependent on He/Ar ratio.Nanoindentation measurements revealed that the FeCrNi films fabricated through DC/RF plasma magnetron sputtering methods exhibited similar hardness values at each He/Ar ratio,while the dispersion of Y_2O_3 apparently increased the hardness of the films.Elastic recoil detection(ERD)showed that DC/RF magnetron sputtered FeCrNi films contained similar He amounts(~17 at.%).Compared with the minimal change of He level with depth in DC-sputtered films,the He amount decreases gradually in depth in the RF-sputtered films.The Y_2O_3-doped FeCrNi films were shown to exhibit much smaller amounts of He owing to the lower backscattering possibility of Y_2O_3 and the inhibition effect of nano-sized Y_2O_3 particles on the He element.     

Tungsten erosion in the baffle and outboard regions of the ITER-like ASDEX Upgrade divertor

Similar to the design of the next-step device ITER, ASDEX Upgrade is equipped with vertical divertor targets with adjacent baffles extending towards the main chamber. In ITER, it is intended to employ tungsten as a plasma-facing material in this baffle area. Tungsten-coated graphite tiles were installed in the divertor baffle and the outboard side regions of ASDEX Upgrade for a full experimental campaign. The erosion behavior of tungsten was investigated by scanning electron microscopy and by measuring the thickness of the tungsten coatings before and after exposure. The coatings had an initial thickness of approximately 450 nm. Two distinct erosion mechanisms were observed: in the outer baffle region a reduction of the coatings’ thickness up to 100 nm was determined after about 6300 s of plasma discharge. On the roof baffle and on the inner baffle modules, no clear reduction of the film thickness was found. In the tracks of arcs, however, the tungsten coatings were completely removed. This represents an erosion of 5-10% of the tungsten-coated surface area in this region.     

Enhancement of hydrogen isotope retention capacity for the impurity deposited tungsten by long-term plasma exposure in LHD

The stress relieved tungsten samples were placed at three positions, PI (sputtering erosion dominated area), DP (deposition dominated area) and HL (Higher heat load area) during 15th plasma experiment campaign in Large Helical Device (LHD) at National Institute for Fusion Science (NIFS), Japan and were exposed to ～ 6700 shots of hydrogen plasma in a 15th long-term experiment campaign in LHD. Thereafter, the additional deuterium ion implantation to these tungsten samples was performed to evaluate the change of hydrogen isotope retention capacity in the samples by long-term plasma exposure. It was found that the carbon-dominant mixed-material layer with more than 100 nm thickness was formed on a wide area of the tungsten surface. The thicker mixed-material layer was formed on the DP sample, where the deuterium retention was about 21 times as high as that for pure W. The major desorption temperature of deuterium was shifted toward higher temperature side, which was comparable to the trapping characteristic of carbon or irradiation damages.     

Behavior of carbon readsorbed on tungsten during low energy Ar ion irradiation at elevated temperatures

A study of the behavior of carbon sputtered and readsorbed after scattering collisions with particles of surrounding gas on the tungsten surface affected by Ar ion irradiation with the flux equal to 2 × 10¹⁶ cm⁻² s⁻¹ extracted from plasma under 300 V negative bias voltage in the temperature range 370-870 K was performed. The dependence of the W sample weight change on the working gas pressure in the range 0.1-10 Pa was registered and the information was deduced about prevailing sputtering-redeposition processes. The depth profiles of carbon at the tungsten surface were measured. We found that carbon distribution profiles in tungsten depend on the C redeposition rate for fixed ion irradiation parameters. Three regimes have been distinguished: (i) at working gas pressure equal to 5 Pa and more, the C redeposition rate prevails the sample surface erosion rate and the W surface is covered by continuous amorphous carbon film (the C film growth regime), (ii) at working gas pressure equal to about 1 Pa, the C redepostion rate is approximately equal to the erosion rate and the W surface is partially covered by redeposited carbon, and (iii) at working gas pressure less than 0.2 Pa, the erosion rate prevails the C redeposition rate (the W surface erosion regime). In the regime of balanced redeposition and erosion deep C penetration depth into nanocrystalline W was registered.It is suggested that under simultaneous C adsorption and ion irradiation at elevated temperature C adatoms are driven from the W surface into grain boundaries and into the bulk by the difference in chemical potentials between the activated W surface and grain boundaries. As the W surface is covered by amorphous C film, the grain boundaries are blocked and the efficiency of carbon transport decreases.     

Estimation of plasma parameters in the process of micro-scale powder plastic and characteristics of its products

The temperature and density of plasma jets were estimated with a Boltzmann plot and Stark broadening of Ar I (696.54 nm) lines by optical emission spectroscopy (OES) in the process of plasma plastic, and the morphology and microstructure of tungsten (W) powders were investigated by scanning electron microscope (SEM) and x-ray Diffraction (XRD), respectively. The results show that the assumption of local thermodynamic equilibrium (LTE) was invalid at the end of the plasma jets, and earlier than this after the injection of tungsten powder. The temperature and electron density of the plasma jets were up to about T=6797 K with Q_c=50 slpm and n_e=1.05×10¹⁶ cm⁻³ with Q_s=115 slpm at Z=60 mm, respectively, and both dropped rapidly with the injected tungsten powders of 20 μm. After the plasma plastic process, the spherical tungsten powders were prepared and there were some satellite particles on the surface of the spherical products. The tungsten powders were both composed of a single equilibrium α-W phase with a body centered cubic (bbc) crystal structure before and after plasma treatment.     

Sterilization of Turmeric by Atmospheric Pressure Dielectric Barrier Discharge Plasma

In this study atmospheric pressure dielectric barrier discharge （DBD） plasma has been employed for sterilizing dry turmeric powders. A 6 kV, 6 kHz frequency generator was used to generate plasma with Ar, Ar/Ou, He, and He/O2 gases between the 5 mm gap of two quartz covered electrodes. The complete sterilization time of samples due to plasma treatment was measured. The most important contaminant of turmeric is bacillus subtilis. The results show that the shortest sterilization time of 15 rain is achieved by exposing the samples to Ar/O2 plasma. Survival curves of samples are exponential functions of time and the addition of oxygen to plasma leads to a significant increase of the absolute value of time constant of the curves. Magnitudes of protein and DNA in treated samples were increased to a similar value for all samples. Taste, color, and solubility of samples were not changed after the plasma treatment.     

Evaluation of energy and particle impact on the plasma facing components in DEMO

We analyze the first wall blanket W/EUROFER configuration for DEMO under steady-state normal operation and off-normal conditions, such as vertical displacement events (VDE) and runaway electrons (RE). The main issue is to find the optimal thickness of the W armor which will prevent tungsten surface from evaporation and melting and, on the other hand, will keep EUROFER below the critical thermal stresses. Under steady-state operation heat transfer into the coolant must remain below the critical heat flux (CHF) to avoid the possible severe degradation of the coolant heat removal capability. From the plasma side it is particularly demanding to keep the bulk plasma contamination during the reactor long operational discharges below the fatal level. The possible damage of the FW materials due to the plasma sputtering erosion is estimated. The minimum thickness of the tungsten amour about 3 mm for W/EUROFER sandwich structure will keep the maximum EUROFER temperature below the critical limit for EUROFER steel under steady-state operation and ITER like cooling conditions.     

Ion beam studies of H and He in metals

The use of ion beams to study hydrogen and helium in metals is demonstrated. The ³He (d,p)⁴He nuclear reaction previously has been used together with ion channeling to determine the lattice locations of ion-implanted D and ³He in tungsten. Preliminary results applying these techniques to helium bubble and blister formation in tungsten are also presented and show that changes attributed to helium bubble formation are observed in tungsten at a He fluence as low as 6 × 10¹⁶ He/cm². The retention of ion-implanted deuterium in W, Au, and Pd surfaces is shown to be greatly enhanced by prior He ion-induced lattice damage. The amount of the damage trapping is also found to depend on whether the metal is in single crystal or polycrystalline form.     

低能氦等离子体低温条件下钨材料的表面形貌

为研究氦等离子体在钨表面造成的表面纳米结构,利用荷兰基础能源研究所Pilot-PSI直线等离子体发生装置在673 K温度下,对钨材料进行了低能（40 eV）高束流强度（4×10²³ m^-2•s^-1）氦等离子体辐照。实验结果表明,辐照后钨材料表面出现了多种不同形态的纳米结构,表面纳米结构和晶粒的表面法向之间存在明显关联。在表面法向为［111］的晶粒表面出现三角形的纳米结构,在［110］取向的晶粒表面出现条带状的纳米结构,而在［001］取向的晶粒表面没有明显的结构出现。晶粒表面的纳米结构尺寸在50 nm左右,高度起伏在5 nm以下。另外,氦等离子体辐照会造成晶界处的高度差,在25 nm左右。分析推测氦等离子体辐照造成的晶粒表面和晶界的形貌可能是由近表面的气泡所导致的。     

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.     

Effects of Low Energy and High Flux Helium／Hydrogen Plasma Irradiation on Tungsten as Plasma Facing Material

The High-Z material tungsten (W) has been considered as a plasma facing material in the divertor region of ITER (International Thermonuclear Experimental Reactor). In ITER, the divertor is expected to operate under high particle fluxes (> 1023 m-2s-1) from the plasma as well as from intrinsic impurities with a very low energy (< 200 eV). During the past dacade, the effects of plasma irradiation on tungsten have been studied extensively as functions of the ion energy, fluence and surface temperature in the burning plasma conditions. In this paper, recent results concerning blister and bubble formations on the tungsten surface under low energy (< 100 eV) and high flux (> 1021 m-2s-1) He/H plasma irradiation are reviewed to gain a better understanding of the performance of tungsten as a plasma facing material under the burning plasma conditions.     

Theoretical studies of macroscopic erosion mechanisms of melt layers developed on plasma facing components

During plasma instabilities in tokamak devices, metallic plasma facing components (PFC) undergo surface vaporization and melting. Macroscopic losses of melt layers are of a serious concern to the lifetime of PFC, the damage of nearby components, and potential core plasma contamination. A normal or inclined plasma stream flowing at the melt layer surface of PFC at very high velocities (∼10⁵ m/s) can induce Kelvin-Helmholtz (K-H) instabilities. We present an extensive linear stability theory and capillary droplet ejection model adapted to the problem of melt layer erosion and splashing. Based on this linear analysis, the stability criterion is established accounting the influence of the thicknesses of both plasma stream and melt layer. The growth rate of the most unstable wave is investigated with respect to different parameters such as plasma density and velocity, material properties, and melt layer thickness. A capillary droplet ejection model is then developed and used to analytically estimate the erosion rate of the melt layer for tungsten and aluminum targets. The present work brings a detailed understanding of the onset of K-H instabilities developed in melt layers due to plasma stream impact and builds a theoretical basis to estimate a macroscopic erosion rate, material losses and lifetime for PFC.     

Experimental results of the erosion of molybdenum due to blistering by He+ bombardment

We report some preliminary measurement of the erosion rate of plasma dumps when bombarded with 100 keV He⁺ ions at high power density ( 1 MW/m²). The expected erosion rates, based on measurements of He blistering that were made at lower power density ( 0.3 MW/m²), indicate a potentially serious problem for fusion reactors. Our tests use a reactorlike power density and produce He blisters at a rate that is slower than predicted by 2 to 4 orders of magnitude, depending on the temperature of the molybdenum target.     

Fuel hydrogen retention of tungsten and the reduction by inert gas glow discharges

Polycrystalline tungsten was exposed to deuterium glow discharge followed by He, Ne or Ar glow discharge. The amount of retained deuterium in the tungsten was measured using residual gas analysis. The amount of desorbed deuterium during the inert gas glow discharge was also measured. The amount of retained deuterium was 2–3 times larger compared with a case of stainless steel. The ratios of desorbed amount of deuterium by He, Ne and Ar glow discharges were 4.6, 3.1 and 2.9%, respectively. These values were one order of magnitude smaller compared with the case of stainless steel. The inert gas glow discharge is not suitable to reduce the fuel hydrogen retention for tungsten walls. However, the wall baking with a temperature higher than 700 K is suitable to reduce the fuel hydrogen retention. It is also shown that the use of deuterium glow discharge is effective to reduce the in-vessel tritium inventory in fusion reactors through the hydrogen isotope exchange.     

Removing of Mixed Coatings by Plasma Discharges

Next generation tokamaks offer the possibility of highly efficient energy generation from the fusion reaction of hydrogen isotopes. In tokamak operation, the core plasma interaction with the wall materials could produce tiles erosion. Redeposition of the eroded materials (C–W–Be) leads to an increase in the allowable tritium load if the coatings are not periodically removed. Amongst removal methods, plasma based techniques employing Ar, H₂ gas have been investigated. Plasma cleaning has been carried out on hydrogenated carbon and carbon–tungsten coatings. It has been shown that at a RF power density of 1.3 W/cm² (pressure of 1 Pa), the plasma cleaning was effective in removing the coatings. Details of further work in this research activity will be presented.     

Measurement of the depth distribution of light impurities in first-wall materials: He in Nb

We have measured the concentrations and depth profiles of implanted helium in niobium by a method demonstrated previously with hydrogen and lithium in copper. The three targets, bombarded at room temperature with 10 keV He⁺ at doses of 0.01, 0.16 and 0.98 C/cm⁺, were respectively: unblistered; covered with circular blisters; and marked with “microrelief”, without blisters. The corresponding doses retained in the metal were 0.0076, 0.039 and 0.052 C/cm² (i.e.≈3 × 10²² He atoms/cm³) with a 10% normalization uncertainty. The profile shapes did not change much: in particular we did not observe, as the dose increased, an accumulation near the surface, which is receding by erosion (sputtering, blistering). These results show that a mechanism of helium loss starts operating at a dose ?0.16 C/cm², i.e. before the bursting of blisters (if they burst at all), and it is most effective near the surface.     

Analysis of an Ar plasma jet in a dielectric barrier discharge conjugated with a microsecond pulse

In this work, an Ar plasma jet generated by an AC-microsecond-pulse-driven dielectric barrier discharge reactor, which had two ring-shaped electrodes isolated from the ambient atmosphere by transformer oil, was investigated. By special design of the oil insulation, a chemically active Ar plasma jet along with a safe and stable plasma process as well as low emission of CO and NO_x were successfully achieved. The results indicated that applied voltage and frequency were basic factors influencing the jet temperature, discharge power, and jet length, which increased significantly with the two operating parameters. Meanwhile, gas velocity affected the jet temperature in a reverse direction. In comparison with a He plasma jet, the Ar plasma jet had relatively low jet temperature under the same level of the input parameters, being preferable for bio-applications. The Ar plasma jet has been tested to interact with human skin within 5 min without the perception of burnt skin and electrical shock.