Potential Plasma-facing Materials Research Articles

Potential plasma facing materials: A case study on CeB6 particulate reinforced W–1Ni matrix composites produced via different sintering techniques

Tungsten (W) is the most promising material for future plasma-facing materials. Its alloys or composites reinforced with various compounds have been studied to improve the irradiation resistance of W. This study aimed to investigate the effect of boron-containing reinforcement in the W matrix against He+ ion irradiation. Cerium hexaboride (CeB6) particulates, which have high neutron shielding properties, were incorporated into the W matrix pre-alloyed by 1 wt% Ni (W1Ni). CeB6 powders were home-made and prepared from CeO2/Mg/B2O3 powder blends via mechanochemical synthesis and purification steps. 1, 5, and 10 wt% CeB6 powders were added to pre-alloyed W1Ni by mechanical alloying, and then they were consolidated by using pressureless sintering (PS, 1400 °C, 1 h) and spark plasma sintering (SPS, 1410 °C, 1 min) techniques. CeB6 particle-reinforced W1Ni composites contained different amounts of reinforcements were prepared by two different sintering methods and were compared with respect to their compositional, microstructural, and microhardness properties and wear and irradiation behaviors. Based on the results, increasing the CeB6 reinforcement amount in the composite triggered the formation of the W2B phase, especially in the W1Ni–10CeB6 composite after both sintering methods. The mechanical and irradiation properties were enhanced more by increasing the CeB6 amount in the case of using the SPS method. When compared to other sintered samples, the SPS'ed W1Ni–10CeB6 composite has the lowest specific wear rate of ∼4 × 10−7 mm3/Nm and the maximum hardness value of ∼21 GPa. According to surface deformation, the W1Ni–5CeB6 composite exhibited comparatively higher resistance to He+ ion irradiation.

Initial study on thermal stability of cold spray tantalum coating irradiated with deuterium for fusion applications

Removal of neutral hydrogen atoms in the plasma edge reduces the number of charge exchange events and thus, the net energy losses in the plasma, significantly improving performance of fusion devices. Effective control of the residual pressure of hydrogen isotopes (HIs) in the plasma edge may be achieved by utilizing a hydrogen absorbing first wall interface capable of withstanding the harsh fusion environment. In this study, we have investigated tantalum (Ta) coating deposited by cold spray technology on 316L stainless steel substrate as a potential plasma-facing material surface. High fluence low energy deuterium plasma irradiation experiments and subsequent thermal annealing cycles associated with thermal desorption spectrometry (TDS) demonstrated superior structural stability of the Ta coating. TDS experiments revealed the outgassing of deuterium (as measure of its retention) for cold spray Ta coatings to be three times higher than bulk Ta and two orders of magnitude greater than bulk polycrystalline W. X-ray photoelectron spectroscopy revealed evolution of oxidation states upon deuterium irradiation and a partial recovery of the metallic signature of Ta after the thermal treatment at 1100 K.

MoB2/MoB ceramic particulate reinforced W-1 wt % Ni matrix composites as potential plasma facing materials

This study reports on the synthesis and characterization of MoB2/MoB ceramic particulate reinforced W-1 wt % Ni matrix (W1Ni) composites for their potential usage as plasma facing materials. A powder metallurgical route consisting of mechanical alloying (MA), cold pressing (CP), cold isostatic pressing (CIP) and pressureless sintering (PS) was applied for the production of MoB2/MoB particulate reinforced W1Ni composites. Different percentages of MoB2/MoB (x = 1, 2, 5 and 10 wt %) were incorporated into W1Ni pre-alloy. MA was performed in a planetary ball mill for different times as 24, 48 and 72 h. After compaction using CP and CIP, green bodies were pressureless sintered at 1400 °C for 1 h under Ar/H2 gas flow. According to the results of XRD characterization, and lattice strain, crystallite size and particle size analyses performed on the powders, MA time was determined as 72 h with an average lattice strain of 2.37% and crystallite size of 7.40 nm. Sintered samples characterized in terms of compositional, microstructural, wear, density and microhardness properties exhibited that W1Ni-10 wt % MoB2/MoB composite has the highest hardness value of ∼6.74 GPa and the lowest wear volume loss of 2.08 × 10-4 mm3, as compared to those of other samples. Sintered samples exposed to 20 eV He+ ion irradiation with a flux of 1.102 × 1021 ions/(m2s), irradiation fluence of 1.32 × 1024 ions/m2 for 20 min had a degraded wave-shaped structure (wavy erosion layers) on their surfaces. However, W1Ni-2 wt % MoB2/MoB composite showed relatively more resistant to ion irradiation attributed to its higher densification rate.

A prospect of using ternary W-5 wt%V-5 wt%Ta alloy manufactured by mechanical alloying and spark plasma sintering as plasma-facing material

For ensuring W employed as the armor of the divertor component in International Thermonuclear Experimental Reactor, a ternary W alloy W-5 wt%V-5 wt%Ta was manufactured by mechanical alloying and spark plasma sintering. (V, Ta)-enriched fine grains and W-enriched grains were observed in W-5 wt%V-5 wt%Ta with a single-phase body centered cubic structure. W-5 wt%V-5 wt%Ta has an ultrafine grain size, and higher relative density, microhardness, and strength than the pure W. The strenghening mechanisms were estimated, and the results indicate that grain refinement, densification, and solid solution play the main, secondary and small strengthening roles, respectively. W-5 wt%V-5 wt%Ta has a good irradiation resistance. This good irradiation resistance has a relationship with the ultrafine grains as well as higher compressive and tensile strength of W-5 wt%V-5 wt%Ta. The damage of irradiation was quantitatively analyzed by the 3D laser scanning confocal microscope. This research work suggests that W-5 wt%V-5 wt%Ta can be applied as a potential plasma-facing material.

Cracks and blisters formed in nanocrystalline tungsten films by helium implantation

Tungsten is regarded as a potential plasma facing material (PFM) in nuclear fusion reactors owing to its high melting point, good thermal conductivity, and low sputtering yield. However, various types of defects such as interstitial atoms, vacancies, dislocation rings, cavities, helium bubbles, and “fuzz” occur in W under helium irradiation. In this study, we obtained α- and β-phase W films with nanochannels and used a helium ion microscope to conduct helium ion implantation experiments on the films. The implantation dose and energy were 1.5 × 1022 ions/m2, and 30 keV, respectively. There was no helium bubble formation in the β-W film until the completion of the implantation. The uniformly distributed columnar crystals growing perpendicular to the substrate constitute remarkable nanochannels. These effectively inhibit the aggregation of helium atoms. For α-W films prepared at high temperatures, the escape efficiency of the implanted helium atoms was reduced owing to the chaotic and reduced number of nanochannels, and helium bubbles are eventually formed. Cross-sectional images revealed that helium atoms tended to gather near the grain boundaries and gradually followed the nanochannels formed by the grain boundaries. The expansion eventually formed numerous longitudinal fissures. The results show that grain boundaries play a decisive role in the radiation resistance of nanostructured materials. Furthermore, the preparation of regular nanochannels that reach the sample surface is an effective method to improve the helium escape capability.

Microstructure and texture evolution in warm-rolled fine-grained tungsten

Fine-grained tungsten was prepared by sintering nano‑tungsten powder and its properties was further improved by warm rolling in the paper. And the microstructure and texture evolution of fine-grained tungsten under different rolling reduction was mainly investigated. Detailed microstructure characterizations under different rolling reduction stages showed a large number of sub-grain structures appeared in the fine-grained rolling tungsten grains, which increases the grain boundary area. Quantitative texture analysis showed that the rolled fine-grained tungsten had a typical BCC texture, the overall texture strength was enhanced after 50% RR and the texture intensity of grains larger than the average grain size was stronger than that of the whole grains, especially in the center of {111} pole figure. Warm rolling of fine-grained tungsten strongly affects the mechanical properties of tungsten, such insights in warm-rolled tungsten are valuable for an understanding of their performance as potential plasma-facing materials in future fusion reactors.

Study of the normal spectral emissivity of tungsten between 170 and 500 °C by a single-wavelength infrared thermometer

Tungsten is considered as the potential plasma facing material in future fusion reactors due to its high melting point, low sputtering, and low fuel retention, and is already used as the baseline material for divertors in some Tokamak devices. The emissivity of tungsten is low and varies with the target temperature, surface state, and wavelength, etc. Therefore, it is indispensable to obtain the emissivity characteristics of tungsten for accurate temperature measurement in the Tokamak devices. In this work, the emissivity measurement system using a single-wavelength infrared thermometer with the same operating conditions as Tokamak devices is constructed and the normal spectral emissivity of tungsten between 170 and 500 °C is measured. The measurement results indicate that the normal spectral emissivity of tungsten ranges from 0.0880 to 0.1291 between 170 and 500 °C with the wavelength of 10.2 μm, and the emissivity increases with the increase of temperature. Meanwhile, the uncertainty of emissivity measurement results is estimated, which ranges from 0.0064 to 0.0125. Besides tungsten, the proposed emissivity method can also be applied to measure the emissivity of other first wall materials such as molybdenum, etc. This work provides a useful reference for accurate temperature measurement of the first wall in nuclear fusion devices.

Theoretical Predictions of the Structural and Mechanical Properties of Tungsten-Rare Earth Element Alloys.

Tungsten (W) is considered as the potential plasma facing material of the divertor and the first wall material in fusion. To further improve the ductility of W, the structural and mechanical properties of W–M (M = rare earth element Y, La, Ce and Lu) alloys are systematically investigated by first-principles calculations. Our results reveal that all the W1−xMx (x = 0.0625, 0.125, 0.1875, 0.25) alloys can form binary solid solution at the atomic level, and the alloys keep bcc lattice structures until the concentration of M increases to a certain value. Although the moduli of the alloys are reduced compared to that of pure W metal, the characteristic B/G ratio and Poisson’s ratio significantly increase, implying all the four rare earth elements can efficiently improve the ductility of W metal. Considering both factors of mechanical strength and ductility, La and Ce are better alloying elements than Y and Lu.

Study of the corrosion characteristics of 304 and 316L stainless steel in the static liquid lithium

• Corrosion behavior of 316L SS and 304 SS exposed to liquid Li at 600 K for 1320 h under Argon (Ar) atmosphere were investigated. • The corrosion characteristics of 316L SS and 304 SS was compared. • 316L SS was found to be more corrosion resistant in high-temperature liquid lithium than 304 SS. Liquid lithium (Li) is considered as the tritium breeder as well as a potential blanket coolant and/or plasma-facing material (PFM) in fusion devices, while steels are widely used as an important structural material in the fusion reactors. Investigation of corrosion behaviors of steels exposed to liquid Li is one of the keys to simultaneous apply Li and steels as potential PFM or blanket materials in the fusion devices. Hence, in this study, the corrosion behavior of 304 stainless steel (304 SS) and 316L stainless steel (316L SS) in static liquid Li, at 600 K, for 1320 h, under Argon (Ar) atmosphere is analyzed. After exposed to liquid Li, the mass loss of 304 SS is about 3.6 times more than that of 316L SS. And the corrosion depth rate of 316L SS and 304 SS is 0.71 μ m / a and 2.6 μ m / a , respectively. Both the materials have shown a non-uniform corrosion behavior. The Signs of grain boundary attack, pitting corrosion, and corrosion products formation, due to the selective dissolution of elements and chemical reactions are also observed. The percentage of corrosion product formation (i.e., carbides) on the surface of 304 SS, is higher than 316L SS. The hardness of two SS materials has a slightly increased after liquid Li exposure.

First-principles investigation of W[sbnd]V and W[sbnd]Mo alloys as potential plasma facing materials (PFMs) for nuclear application

Density-functional theory (DFT) based first-principles calculations were used to investigate the crystal structure, binding energy, phase stability and elastic properties of body-centered cubic (BCC) W-based binary solid solutions. The effect of alloying up to 50 atomic percent (at.%) concentration range is determined from the virtual-crystal approximation (VCA) approach. Resulting BCC solid solutions are assessed in comparison to the ideal Vegard's law. Solubility of the alloying elements is characterized by the negative enthalpy of mixing. The values of elastic constants computed for the ground state structures are used to assess the effect of alloying on the ductility and hardness. Based on current results, it seems key to strike a tricky balance between moderate Pugh's modulus ratio (B/G) between bulk and shear moduli (B,G) and elastic anisotropy (A) such that high hardness is not completely compromised at the expense of ductility. The predicted property trends are in agreement with existing theoretical and experimental data, which is indicative that the VCA approach is reliable in development of such alloys.

Corrosion of 304 stainless steel in static liquid lithium under high vacuum

• Corrosion behaviors of 304SS exposed to liquid Li at 700 K for 1320 h under high vacuum were investigated. • Corrosion of 304 SS in liquid Li at 700 K was compared to at 600 K and 640 K. • Cu of sealing gasket was deposited on the wall of test vessel. To study the compatibility especially the corrosion characteristic of structural material stainless steel (SS) with liquid lithium (Li) is an important factor for the simultaneous application of liquid Li and SS as a blanket material or potential plasma facing material (PFM) in future fusion devices. The corrosion behaviors of 304SS exposed to liquid Li at 700 K for 1320 hr under high vacuum were investigated. After exposure to liquid Li, 304SS produces non-uniform corrosion which includes corrosion products and grain boundary corrosion. Compared with the corrosion behaviors at 600 K and 640 K, temperature plays an important role in the corrosion behaviors of 304SS in static liquid Li. The weight loss, corrosion depth and corrosion products of 304SS specimens increase with the corrosion temperature. In general, the compatibility between 304SS and static liquid Li is relatively good under the experimental conditions. And besides, it is unexpectedly found that the compatibility of copper with Li is not good.

LIBS analysis of samples from the COMPASS vacuum chamber after liquid metal experiments – Li campaign

• Quantitative elemental composition of Li-based coating on screws by CF-LIBS. • Li, B, C, Fe, Na, K, Mg, Ca, and Si have been observed in the LIBS spectra. • The electron temperature has been evaluated using multi-elemental SB plots. • The electron density was obtained from the Stark broadening of the H α line. • Evaporated/redeposited Li in the COMPASS vacuum vessel information is provided. To overcome the erosion of the plasma-facing materials (PFMs) inside a fusion reactor, liquid metals such as lithium or tin are being explored as potential PFMs in fusion reactors due to their several advantages such as self-repair or high-power exhaust capability. Lithium can reduce the oxygen and carbon impurity contamination by chemically binding these species to the wall. Thus, experiments of power handling performance with a special liquid metal (Li) divertor module based on the capillary porous system technology were performed in the tokamak COMPASS. The elemental analysis of 14 screw samples located around the vacuum vessel provides information about the migration of evaporated/redeposited liquid elements. The main goal of this work is to characterize and quantify screws containing Li-based coating by LIBS by measuring the quantitative elemental composition. The studied samples were Li-coated Ni-Cr-based screws. The laser ablation was performed in air at atmospheric pressure using a 5 ns pulse Nd:YAG laser at 1064 nm. Suitable Li I, Ca I-II, and Fe I spectral lines were employed for the evaluation of the electron temperature of the plasma using multi-elemental Saha-Boltzmann plots. The electron density was obtained from the Stark broadening of the H α spectral line. The elemental content of each sample was then obtained by calibration-free LIBS (CF-LIBS).

Evolution of microstructure and texture of moderately warm-rolled pure tungsten during annealing at 1300 °C

The mechanical behavior, microstructure and texture evolution were investigated during isothermal annealing at 1300 °C of pure tungsten moderately warm-rolled to 67% thickness reduction. The degradation of the mechanical properties is characterized by hardness testing. The microstructure and texture evolution during heat treatment were characterized by Electron Backscatter Diffraction. During annealing of the moderately warm-rolled tungsten, recrystallization occurred first, quickly followed by relatively slow grain growth. The recrystallized volume fractions determined from hardness measurements and microstructural characterization were essentially the same. The evolution of the grain sizes during recrystallization was analyzed independently for deformed and recrystallized grains. Quantitative texture analysis showed that the overall texture strength is enhanced after recrystallization. As recrystallization strongly affects the mechanical properties of tungsten, such insights in the annealing behavior of warm-rolled tungsten plates are valuable for an understanding of their performance as potential plasma-facing materials in future fusion reactors.

Microstructure and mechanical properties of Tungsten and Tungsten-Tantalum thin film deposited RAFM steel

ABSTRACTTungsten (W) for its high irradiation behaviour is being studied globally as a potential plasma-facing material. However, thermo-mechanical fatigue caused by high temperature in the fusion process poses a great challenge for the materials engineers. Tungsten, known for its high stiffness and sputtering threshold, has a typical Ductile to Brittle Transition Temperature (DBTT) at 600K. Due to the DBTT of W, stresses raised by thermal fluctuations generate cracks and failure of material occurs. To overcome this challenge, nanostructured composite tungsten–tantalum (W–Ta) thin film has been deposited on Reduced Activation Ferritic Martensitic (RAFM) steel by RF magnetron sputtering process. The secondary phase of Ta in the W matrix has been found to reduce the surface hardness at room temperature. Effects of RF power, substrate temperature and Ta percentage in W matrix have been considered in this study. Various thin-film properties viz adhesion, microstructure and hardness have been studied and analysed in this report.

Site preference and diffusion behaviors of H influenced by the implanted-He in 3C-β SiC

SiC materials are potential plasma facing materials in fusion reactors. In this study, site preference and diffusion behaviors of H in the pure 3C-β SiC and in the He-implanted 3C-β SiC are investigated, on the basis of the first-principles calculations. We find that the most stable sites for H in the pure 3C-β SiC is the anti-bond site of SiC at the side of C (ABc), while it becomes the bond-center (BC) site of SiC in the He-implanted 3C-β SiC. Analysis on the electronic structures reveals that such change is attributed to the reduction of hybridization of CSi bonds induced by the implanted He. Moreover, the presence of He strongly affects the vibration features of the system in the high frequency region, causing a blue shift of 3.10 meV for CH stretch mode with H at ABc site and a red shift of 20.46 meV for that at BC site, with respect to that in the pure system. In the pure 3C-β SiC, H is diffusive with an energy cost of about 0.50 eV, preferring to rotate around the C atom in a SiC tetrahedron with an energy barrier of just about 0.10 eV. In contrast, in the He-implanted 3C-β SiC, the energy barrier for H migration goes up to be about 0.95 eV, indicating that the implanted-He blocks the diffusive H to some extent. Our calculations also show that the influence of He on H diffusion works in a short range, just covering the nearest neighbors.

Studies of He irradiated tungsten using transmission electron microscope

Tungsten (W) as one of the potential plasma-facing materials (PFMS) in the nuclear fusion reactors would suffer severe irradiation which can result in deterioration of properties. Understanding the irradiation effects on PFMS is thus critical to improve their performances. Here, the He irradiation effects is studied by focusing on microstructures of the W surfaces. As the irradiation fluence increases, surface morphologies of W become rough with abundant nano-structures, namely nanotendrils. Transmission electron microscope (TEM) images show these nanotendrils contain lots of He voids, which reveals good deformation behavior albeit brittle.

Development of the flowing liquid lithium limiter for EAST tokamak

Lithium coating technology and liquid lithium limiter have been applied on HT-7 tokamak and many significant results were obtained. The study of exploring lithium as potential plasma facing material is being carried out on EAST tokamak. A flowing liquid lithium (FLiLi) limiter has been successfully tested in the EAST in 2014, and the in-vessel electromagnetic (EM) pump was validated to make liquid lithium circulate from the bottom collector to the distributor of the limiter, which realized the FLiLi operated under steady-state and also drastically reduced the total amount of lithium used in the experiment. To adapt to the plasma shape during discharge, the limiter can be moved along the guide rail into vacuum vessel to meet various plasma scenarios by the driving system. During the operation, the lithium can be easily controlled by adjusting the current of EM pump and the <0.1mm thickness of lithium film and a mass flow rate ∼2cm3/s were achieved. However, some problems were also exposed, which require improvement for the next generation of limiter. The current design, manufacturing and the latest results of the new EAST FLiLi limiter were introduced in this paper.

Effect of high-flux, low-energy He+ ion irradiation on Ta as a plasma-facing material.

The goal of this work is to assess Ta as a potential plasma-facing material for future fusion reactors in terms of its response to high-flux, low-energy He+ ion irradiation. Ta samples were irradiated with 100 eV He+ ions at various fluences up to 3.5 × 1025 ions m−2 while simultaneously heated at constant temperatures in the range 823–1223 K. SEM studies show that irradiated Ta surfaces undergo significant morphology changes that have a strong dependence on both ion fluence and sample temperature. Optical reflectivity complements SEM and demonstrates a vertical growth of surface structures with increasing fluence. Ex situ XPS and XRD both show significant oxidation of the irradiated Ta surfaces, giving further qualitative information on the extent of surface modification. Overall, these irradiation-induced structures on Ta are similar to early-stage “fuzz” structures observed in W. However, Ta exhibits a higher fluence threshold for structure formation. While Ta may have less desirable bulk properties (e.g., thermal conductivity) when compared to W, its higher resilience to He+ ion-induced surface modification suggests that surface thermal and mechanical properties may not degrade as quickly in extreme fusion environments; this quality may be a redeeming factor for Ta as a plasma-facing material.

Study of the corrosion behaviors of 304 austenite stainless steel specimens exposed to static liquid lithium at 600 K

Investigation of corrosion behavior of stainless steel served as one kind of structure materials exposed to liquid lithium (Li) is one of the keys to apply liquid Li as potential plasma facing materials (PFM) or blanket coolant in the fusion device. Corrosion experiments of 304 austenite stainless steel (304 SS) were carried out in static liquid Li at 600 K and up to1584 h at high vacuum with pressure less than 4 × 10−4 Pa. After exposure to liquid Li, it was found that the weight of 304 SS slightly decreased with weight loss rate of 5.7 × 10−4 g/m2/h and surface hardness increased by about 50 HV. Lots of spinel-like grains and holes were observed on the surface of specimens measured by SEM. By further EDS, XRD and metallographic analyzing, it was confirmed that the main compositions of spinel-like grains were M23C6 carbides, and 304 SS produced a non-uniform corrosion behavior by preferential grain boundary attack, possibly due to the easy formation of M23C6 carbides and/or formation of Li compound at grain boundaries.

Positron annihilation lifetime measurement and X-ray analysis on 120 MeV Au+7 irradiated polycrystalline tungsten

In order to simulate radiation damages in tungsten, potential plasma facing materials in future fusion reactors, surrogate approach of heavy ion irradiation on polycrystalline tungsten is employed. Tungsten specimen is irradiated with gold heavy ions of energy 120 MeV at different fluences. Positron annihilation lifetime measurements are carried out on pristine and ion beam irradiated tungsten specimens. The variation in positron annihilation lifetime in ion irradiated specimens confirms evolution of vacancy clusters under heavy ion irradiation. The pristine and irradiated tungsten specimens have also been characterized for their microstructural, structural, electrical, thermal, and mechanical properties. X-ray diffractograms of irradiated tungsten specimens show structural integrity of polycrystalline tungsten even after irradiation. Nevertheless, the increase in microstrain, electrical resistivity and microhardness on irradiation indicates creation of lattice damages inside polycrystalline tungsten specimen. On the other hand, the thermal diffusivity has not change much on heavy ion irradiation. The induction of damages in metallic tungsten is mainly attributed to high electronic energy loss, which is 40 keV/nm in present case as obtained from SRIM program. Although, concomitant effect of nuclear losses on damage creation cannot be ignored. It is believed that the energy received by the electronic system is being transferred to the atomic system by electron-phonon coupling. Eventually, elastic nuclear collisions and the transfer of energy from electronic to atomic system via inelastic collision is leading to significant defect generation in tungsten lattice.